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FISPACT-II User Manual - Culham Centre for Fusion Energy

1. 19 The abnormal error returns from LSODES 168 20 Low energy group boundaries a ao o e e a e a 21 High energy 55 MeV group structures a 22 Energy group boundaries for LANL 66 aoaaa aaa a 23 Energy group boundaries for CCFE 162 24 Energy group boundaries for LLNL 616 180 25 Energy group boundaries for CCFE 709 183 26 Non TENDL evaluations in TENDL 2013 198 CCFE Page 13 of 200 CCFE R 11 11 Issue 6 LIST OF TABLES FISPACT II User Manual This page has been left intentionally blank CCFE Page 14 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual 1 Introduction This document contains guidance and reference material for users of FISPACT II FISPACT II is a completely new inventory code designed to be a functional replacement for FisPACT 2007 1 This new code is written in object style Fortran 95 al 6 7 and has extended physical models and improved numerical algorithms compared to the old code Users familiar with the old code will be able for most cases to use the new code with their existing control input files Some new keywords have been added to deal with the new capabilities and some of the old keywords have become obsolete The differences between FISPACT 11 and FISPACT 2007 that are relevant to the user are listed in Section 3 below This version of the user manual refers to the
2. 2 190 B 5 2 Deuteron eaf d fis and eaf_d_asscfy 190 B 5 3 Proton eaf p fis and eaf passcfy o 193 B 6 Radiological Data s 193 B 6 1 Biological hazard index eaf haz 193 B 6 2 Legal transport index eafa2 s B 6 3 Clearance index eaf clear o o a B 7 Absorption Data eafabs o 20000004 e C TENDL Library Data 195 C 1 Cross section Data 196 C 2 Fission Yield Data o e len 1983 C 3 Variance and Covariance a 198 CA Probability Fables cw car Re REIR ee E 199 7 5 Decay Data 22 ge ag Reo noe y ue a e a ERAS ROBUR Ge C 6 Radiological Data es D ENDF B VIT 1 Library Data E JENDL 4 0 Library Data Y N HH o O kel ko F JEFF 3 2 Library Data CCFE Page 12 of 200 LIST OF FIGURES CCFE R 11 11 Issue 6 FISPACT I User Manual List of Figures 1 Files used in the cross section collapse run example 2 Files used in the decay and fission data condense run example 3 The total activity graph produced by the inventory run 4 Files used in the inventory run example 38 5 Graphical output produced using the gnuplot visualisation package 6 Directed graph representation of reactions and decays 134 7 Projection operator S E th Geek E Bape Me Be Dyed as ae Gs oes Ge A 136 8 Decay secondaries ll ss 9 Pat
3. P 35E 2 61E 02 08 10E 05 20E 02 81E413 3 47E 04 2 9 44E 03 28 39E 04 02E 00 15E 12 8 22E 02 68 4 32E 04 3 59E 03 Cooling 1 1 Cooling gt 2 85E 03 94E413 8 16E 04 2 Cooling 1 Mass of material input 1 0000E 00 kg Total irradiation time 7 889400E407 s Total fluence 3 374304E422 n cm2 Mean flux 4 277010E 14 n cm2 s Number of on times il ispact run time 0 37894 secs The final section of the output file contains QA information that displays a list of all the external files used during the run and run timestamps The numbers after the unit names are the internal unit numbers input 5 inventory i graph 10 inventory gra a2data 11 EAF2010data eaf a2 20100 collapxi 12 collapsed cross section data arrayx 13 condensed decay and fission data hazards 14 EAF2010data eaf haz 20100 gnuplot 15 inventory plt ind nuc 18 EAF2010data eaf index 20100 output 38 inventory out absorp 39 EAF2007data eaf_abs_20070 runlog 48 inventory log Run timestamp 17 11 57 25 July 2011 Current time 17 11 58 25 July 2011 7 2 The Inventory Run runlog File The runlog file contains the run monitoring and error logging data from a FISPACT II run The first part gives the name of the log file the run timestamp the files file name the fileroot and a list of the file mappings specified in the files file LOG FILE inventory log 1
4. The notation used in the sub subsection headings defining the keywords is as follows 1 keywords are displayed in bold font 2 arguments are in lower case italics 3 default values of arguments assumed by the program if the keyword is not used are displayed in curly brackets 4 arguments that are present only for certain values of earlier arguments are dis played in angle brackets lt gt Some keywords may appear in more than one section of the input file In these cases the full descriptions of the keywords are placed in the first subsection below where the keywords are permitted Summary descriptions are repeated in other subsections where the keywords are permitted and any context specific details are noted CCFE Page 46 of 200 5 1 Library Data Preparation CCFE R 11 11 Issue 6 FISPACT 1 User Manual 5 1 Library Data Preparation The first section of the input file deals with the input and processing of library data and with initial output settings It is terminated with the keyword FISPACT 5 1 1 CLOBBER In order to prevent accidental loss of data the default action of FISPACT II is to terminate with a fatal error if output files of the same names as specified in the current run already exist in the present working directory This keyword allows existing output files to be overwritten without any error messages from the program 5 1 2 COVARIANCE If this keyword is present a collapse run will c
5. The density of the material specified by MASS or FUEL is 8 96 g cm 5 2 7 DOSE ndose 1 dist 0 Dose rates are calculated for a semi infinite slab of the material This is the default if the keyword is not used or if ndose 1 but if ndose 2 then the calculations are done for a point source of 1 g of material at a distance of dist metres dist is not used for the semi infinite slab as the contact dose rate is always assumed The minimum distance is 0 3 m if a smaller value is specified then dist is set to 0 3 m and a message to this effect is printed An example of the use of this keyword is DOSE 2 1 0 In this case the dose due to a point source 1g of the irradiated material at a distance of 1 m is calculated The DOSE keyword must not appear more than once in an input file CCFE Page 60 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual 5 2 8 EAFVERSION neafv 7 This keyword may also appear in the library data preparation section of the input file see Section 5 1 3 EAFVERSION must be used before the keywords GETXS or GETDECAY to which it refers as it determines which input streams from the files file are used to read the nuclear data 5 2 9 END Title This keyword terminates the input of data for a particular run It is the final keyword that is read from the input file and the remainder of the file is ignored The text used in Title is arbitrary and must be preceded
6. 239 2 6303E 4 417 7 2444E 0 595 1 9953E 3 62 1 9600E 7 240 2 5119E 4 418 6 9183E 0 596 1 9055E 3 63 1 9400E 7 241 2 3988E 4 419 6 6069E 0 597 1 8197E 3 64 1 9200E 7 242 2 2909E 4 420 6 3096E 0 598 1 7378E 3 65 1 9000E 7 243 2 1878E 4 421 6 0256E 0 599 1 6596E 3 66 1 8800E 7 244 2 0893E 4 422 5 7544E 0 600 1 5849E 3 67 1 8600E 7 245 1 9953E 4 423 5 4954E 0 601 1 5136E 3 68 1 8400E 7 246 1 9055E 4 424 5 2481E 0 602 1 4454E 3 69 1 8200E 7 247 1 8197E 4 425 5 0119E 0 603 1 3804E 3 70 1 8000E 7 248 1 7378E 4 426 4 7863E 0 604 1 3183E 3 71 1 7800E 7 249 1 6596E 4 427 4 5709E 0 605 1 2589E 3 72 1 7600E 7 250 1 5849E 4 428 4 3652E 0 606 1 2023E 3 73 1 7400E 7 251 1 5136E 4 429 4 1687E 0 607 1 1482E 3 74 1 7200E 7 252 1 4454E 4 430 3 9811E 0 608 1 0965E 3 75 1 7000E 7 253 1 3804E 4 431 3 8019E 0 609 1 0471E 3 T6 1 6800E 7 254 1 3183E 4 432 3 6308E 0 610 1 0000E 3 77 1 6600E 7 255 1 2589E4 4 433 3 4674E 0 611 9 5499E 4 78 1 6400E 7 256 1 2023E 4 434 3 3113E4 0 612 9 1201E 4 79 1 6200E 7 257 1 1482E 4 435 3 1623E 0 613 8 7096E 4 80 1 6000E 7 258 1 0965E 4 436 3 0200E 0 614 8 3176E 4 81 1 5800E 7 259 1 0471E 4 437 2 8840E 0 615 7 9433E 4 82 1 5600E 7 260 1 0000E 4 438 2 7542E 0 616 7 5858E 4 83 1 5400E 7 261 9 5499E 3 439 2 6303E 0 617 7 2444E 4 84 1 5200E 7 262 9 1201E 3
7. 5 3 10 NOSTABLE This keyword may also be used in the initial conditions section of the input file see Section 5 2 33 5 3 11 NOT1 This keyword may also be used in the initial conditions section of the input file see Section 5 2 34 5 3 12 NOT2 This keyword may also be used in the initial conditions section of the input file see Section 5 2 35 CCFE Page 90 of 200 5 3 Inventory Calculation Phase CCFE R 11 11 Issue 6 FisPACT II User Manual 5 3 13 NOT3 This keyword may also be used in the initial conditions section of the input file see Section 5 2 36 5 3 14 NOTA This keyword may also be used in the initial conditions section of the input file see Section 5 2 37 5 3 15 OVER ja This keyword may also be used in the initial conditions section of the input file see Section 5 2 38 5 3 16 PARTITION npart sym n rpart n n 1 npart This keyword allows the material to be split or partitioned into two streams during an irradiation or cooling The part that continues to be considered by the code consists of all elements not specified npart elements are specified and the fractions zpart n of the specified elements sym n The stream containing the remainder is lost and cannot be followed any further by the code Typical uses of this keyword might be to model recycling of irradiated material or the loss by diffusion of tritium from a material In the first case PARTITION would be used after irradiation
8. 571 6 0256E 3 38 5 4000E 7 216 7 5858E 4 394 2 0893E 1 572 5 7544E 3 39 5 2000E 7 217 7 2444E 4 395 1 9953E 1 573 5 4954E 3 40 5 0000E 7 218 6 9183E 4 396 1 9055E 1 574 5 2481E 3 41 4 8000E 7 219 6 6069E 4 397 1 8197E 1 575 5 0119E 3 42 4 6000E 7 220 6 3096E 4 398 1 7378E 1 576 4 7863E 3 43 4 4000E 7 221 6 0256E 4 399 1 6596E 1 577 4 5709E 3 44 4 2000E 7 222 5 7544E 4 400 1 5849E 1 578 4 3652E 3 45 4 0000E 7 223 5 4954E 4 401 1 5136E 1 579 4 1687E 3 46 3 8000E 7 224 5 2481E 4 4 402 1 4454E4 1 580 3 9811E 3 47 3 6000E4 7 225 5 0119E 4 403 1 3804E4 1 581 3 8019E 3 48 3 4000E 7 226 4 7863E 4 404 1 3183E 1 582 3 6308E 3 49 3 2000E 7 227 4 5709E 4 405 1 2589E4 1 583 3 4674E 3 50 3 0000E 7 228 4 3652E 4 406 1 2023E 1 584 3 3113E 3 51 2 9000E 7 229 4 1687E 4 407 1 1482E 1 585 3 1623E 3 52 2 8000E 7 230 3 9811E 4 408 1 0965E 1 586 3 0200E 3 53 2 7000E 7 231 3 8019E 4 409 1 0471E 1 587 2 8840E 3 54 2 6000E 7 232 3 6308E 4 410 1 0000E 1 588 2 7542E 3 55 2 5000E 7 233 3 4674E 4 411 9 5499E 0 589 2 6303E 3 56 2 4000E 7 234 3 3113E 4 412 9 1201E 0 590 2 5119E 3 57 2 3000E 7 235 3 1623E 4 413 8 7096E 0 591 2 3988E 3 58 2 2000E 7 236 3 0200E 4 414 8 3176E 0 592 2 2909E 3 59 2 1000E 7 237 2 8840E 4 415 7 9433E 0 593 2 1878bE 3 60 2 0000E 7 238 2 7542E 4 416 7 5858E 0 594 2 0893E 3 61 1 9800E 7
9. 700 1 5849E 5 167 T 2A44E4 5 345 1 9953E 2 523 5 4954E 2 701 1 5136E 5 168 6 9183E 5 346 1 9055E 2 524 5 2481E 2 702 1 4454E 5 169 6 6069E 5 347 1 8197E 2 525 5 0119E 2 703 1 3804E 5 170 06 3096E 5 348 1 7378E4 2 526 4 7863E 2 704 1 3183E 5 171 6 0256E 5 349 1 6596E 2 527 4 5709E 2 705 1 2589E 5 172 5 7544E4 5 350 1 5849E4 2 528 4 3652E 2 706 1 2023E 5 173 5 4954E4 5 351 1 5136E4 2 529 4 1687E 2 707 1 1482E 5 174 5 2481E4 5 352 1 4454E 2 530 3 9811E 2 708 1 0965E 5 175 5 0119E 5 353 1 3804E 2 531 3 8019E 2 709 1 0471E 5 176 4 7863E 5 354 1 3183E 2 532 3 6308E 2 177 4 5709E 5 355 1 2589E 2 533 3 4674E 2 178 4 3652E 5 356 1 2023E 2 534 3 3113E 2 B 1 1 Weighting spectra Different micro flux weighting spectra are used depending upon which group structure is required and for which application the calculation needs to be performed The weighting spectra are usually generated at a temperature of 294K however higher temperatures 574 K and 824 K have also been prepared The weighting spectra used to generate fission relevant libraries in the WIMS XMAS and TRIPOLI group format from EAF point wise data are as follows CCFE Page 186 of 200 B 2 Cross section Data CCFE R 11 11 Issue 6 FisPACT II User Manual Energy range Micro flux weighting spectrum 1 0 x 107 0 2eV Maxwellian T 0 0253 eV
10. CCFE Page 111 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual test65 shows three irradiation steps each with a different time interval flux ampli tude and neutron spectrum as indicated by the rateeq number In examples where the flux amplitudes vary but the same collapsed cross sections are used throughout e g test18 the rateeq number remains unchanged PATHWAY ANALYSIS IRRADIATION PHASE number of steps irradiation time 2 63952E 06 secs step start end delta t flux rateeq number sec sec sec n cm 2 s number 2 0 00000E 00 5 27818E 01 5 27818E 01 2 59032E 14 3 5 27818E 01 5 27818E 05 5 27817E 05 2 64634E 14 4 5 27818E 05 2 63952E 06 2 11170E 06 2 66930E 14 path floor 5 00000E 01 of target inventory loop floor 1 00000E 00 of path inventory max depth 10 maximum number of edges between source and target Pathways are given in order of decreasing dominance of target nuclide as ordered in the dominant nuclide tables above Pathways are retained if they contribute more than the path floor percentage of the number of target atoms given by the full rate equation solution for the time interval Loops are retained in a pathway if they contribute more than the loop floor percentage of the number of target atoms created along the pathway The max depth is the maximum number or parent daughter pairs edges that are considered in a path Pathways are analysed between the nuclides of the ini
11. H n npt 184 2 4 2 H 3H n ndt 185 2 5 2 H H n nph 186 3 4 2 1H He n ndh 187 3 5 2 H He n nth 188 3 6 2 H He n nto 189 3 7 2 H He n 2n2p 190 2 3 214 1H n ph 19 3 3 2 H He n dh 192 3 4 2 H He n ha 193 4 6 2 He He n 4n2p 194 2 5 2 lH tH n 4n2a 195 4 11 2 He He n 4npa 196 3 8 2 H He n 3p 197 3 2 S H 1H tH n n3p 198 3 3 31H 1H HH n 3n2pa 199 4 8 3 H H He n 5n2p 200 2 6 2 H H A 4 1 Other reactions gas heat and damage The neutron induced cross section set of Table 12 has been extended and complemented by a further set of diagnostic reactions of technological importance in the design and assessment of nuclear power systems These are listed in Table 13 For MT 201 207 the z denotes any projectile y n d p and X is a positive integer There may be other products from the reaction but these are not displayed Data for these reactions are included in the new TENDL 2013 libraries and are summarised in the printlib cross section output tables The NJOY modules GASPR and HEATR can be used on a properly filled evalua tion to generate gas production reactions heat production cross sections and radiation damage energy production Heating is described by the Kerma Kinetic Energy Re lease in Materials coefficient and the damage caused by i
12. He n 2a 108 4 7 2 4He He n 3a 109 6 11 3 He tHe He n 2p 111 2 1 2 H 1H n pa 112 3 4 2 H_ He n t2a 113 5 10 3 H He He n dla 114 5 9 3 H He He n pd 115 2 2 2 H H n pt 116 2 3 2 H H n da 117 3 lt 2 H He n 5n 152 0 4 0 n n 153 0 5 0 n 2nt 154 1 4 1 9H n ta 155 3 6 2 H He n 4np 156 1 4 1 H n 3nd 157 1 4 1 H n ndo 158 3 6 2 H He n 2npa 159 3 6 2 lH He n 7n 160 0 6 0 n 8n 161 0 7 0 n Sup 162 1 5 1 TH n 6np 163 1 6 1 H n 7np 164 1 7 1 1H n 4na 165 2 7 1 4He n dna 166 2 8 1 He n 6na 167 2 9 1 He n 7na 168 2 10 1 He n 4nd 169 1 5 I H n 5nd 170 1 6 1 H n 6nd 171 1 7 1 H n 3nt 172 1 5 1 H n 4nt 173 1 6 11H continued on next page CCFE Page 141 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual continued from previous page Projectile Products MT AZ AA NSEC Secondaries n 5nt 174 1 7 11H n 6nt 175 1 8 1 H n 2nh 176 2 4 1 He n 3nh 177 2 5 1 He n 4nh 178 2 6 1 He n 3n2p 179 2 4 2 H tH n 3n2a 180 4 10 2 4He He n 3npa 181 3 7 2 1H He n dt 182 2 4 2 H H n npd 183 2 3 2 H
13. This keyword causes the gamma ray spectrum in MeV s in the 24 energy group format or 22 group format if the GROUP parameter is 1 to be written to an external file TAB4 In addition a second column showing the number of gammas per group is also given in TAB4 Note that the stream number id is now ignored Both NOTA and TAB4 may be used several times during a run to restrict and restore the output as required 5 2 57 TIME t When used in the initial conditions section of the input file this keyword sets the first time interval for the inventory calculation terminates the initial conditions section and triggers the processing of any keyword actions that may have been queued The time interval is specified in seconds by default but the value of the time may be followed by one of the following keywords SECS MINS HOURS DAYS or YEARS so that time units other than seconds may be used 5 2 58 TOLERANCE itol atol 101 rtol 2 x 107 This keyword is used to set absolute atol and relative rtol tolerances that are passed to the LSODES solver to control the convergence of the solution If tol 0 CCFE Page 82 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual the tolerances are applied to the main inventory calculation and if tol 1 they are applied to pathways calculations The keyword may be used twice to adjust both pairs of tolerances If the keyword is used twice or more for a given it
14. the cross section data reading condensing and storing the decay data and fission data and storing the regulatory radiological data potential biological hazards clearance data and legal transport data FISPACT II constructs effective cross sections by collapsing the energy dependent cross sections in the EAF or ENDF libraries i e taking the weighted average over energy of the cross section weighted by the irradiating projectile flux Equation on page 134 where the projectiles may be neutrons protons deuterons alpha particles or gamma rays The collapse process differs slightly depending on the data libraries used EAF The cross section data in the EAF 2010 library come in 9 different energy group structures and the most appropriate choice depends on the application and the energy spectrum of the irradiating flux see Appendices B 1 and B 2 Provided with the EAF library are a number of sample projectile flux spectra for specific CCFE Page 17 of 200 CCFE R 11 11 Issue 6 2 WHAT FISPACT II DOES FISPACT II User Manual applications in energy groups that match the cross section energy groups see Appendix B 3 FisPACT II can also take user defined spectra specified in arbi trary groups and convert them into a suitable form to match the cross section energy groups Uncertainties in the cross sections are available only as variances in fewer up to five energy groups and are assumed to be uncorrelated T
15. 2 20 9 0690E 2 187 8 4820E4 2 188 7 4850E 2 144 7 4852E 189 7 0790E 2 190 6 7730E 2 60 6 7729E 2 191 6 3100E 2 192 5 8300E 2 145 5 8295E 2 71 5 8294E42 193 5 1450E 2 194 4 5400E 2 146 4 5400E 2 72 4 5399E 2 61 4 5400E 2 195 3 9810E 2 2 70 7 4851E4 2 59 7 4852E 2 62 3 7170E4 2 21 3 6730E 2 T T 196 3 5360E 2 147 3 5358E4 2 73 3 535 7E42 197 3 0430E4 2 63 3 0433E4 2 198 2 7540E 2 148 2 7536E 4 2 74 2 7536E 2 199 2 4300E4 2 200 2 1450E 2 149 2 1445E4 2 75 2 1445E42 201 2 0400E4 2 64 2 0400E 2 202 1 7780E 2 203 1 6700E 2 150 1 6702E 2 76 1 6701E4 2 204 1 5850E 2 65 1 4863E 2 22 1 4870E 2 205 1 3670E4 2 66 1 3674E 2 206 1 3010E 2 151 1 3007E 2 77 1 8007E4 2 207 1 1220E4 2 continued on next page CCFE Page 174 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FisPACT II User Manual continued from previous page TRIPOLI 315 VITAMIN J 175 GAMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 208 1 0130E 2 152 1 0130E4 2 78 1 0130E 2 209 9 1660E 1 67 9 1661E 1 210 8 5280E 1 211 7 8890E 1 153 7 8893E T ra 79 7 8892E4 T mM 68 7 5674E 1 23 7 5500E 1 212 7 0790E 1 213 6 7900E 1 69 6 7904E 1 214 6 3100E 1 215 6 1440E 1 154 6 1442E4 1 80 6 1441E4 1 216 5 5590E 1 70 5 5595E 1 Tl 5 1578E 1 217 5 0120E 1 72 4 8252E 1 2
16. 39 1 I i I 1 Lui iui I 1 107 Cirt 9109 Cit Si 95 935 Srt 95 95 10 Cig Neutron energy eV b Paluel LWR 10 T om T sE 3 Y nia 10 i d i i Y sE ce s 3 TD Neutron flux per lethargy 1 r 1 y 1 r L y L 5 i0 5 99 5 93 5 94 5 99 5 99 5 Neutron energy eV d Superph nix fast reactor 10 T sem _ E uet E i 107 r SE gt E E 3 9 107 eed Bo os E 8 x 3 a 8 b 10 B B sE 3 3 E 10 4 SE 5 10 i 1 a L 10 2 4 6 8 i09 2 4 e 8 108 2 4 6 8 io 2 Neutron energy eV f Californium 252 fission neutron spectra CCFE Page 191 of 2 CCFE R 11 11 Issue 6 FISPACT II User Manual B EAF LIBRARY DATA Neutron flux per lethargy T 10 M 108 1 I 1 I 1 I 1 i0 9 91 9 99 Siol C 5 99 5 9 9 55 10 10 Neutron energy eV a Bigten experiment 10 T mmo T BUE 10 Neutron flux per lethargy 8 10 5 10 8 ki 1 L L i L L ip 5193 5 99 19 9499 F210 Fp 10 10 10 Neutron energy eV c ENEA Frascati D T 1015 T T T T sE 1014 Neutron flux per lethargy 3 109 t Dii fi L 1 1 i to f4g9 poh Su O apt Said Neutron energy ev e IFMIF D Li Figure 11 CCFE Page 192 of 2 Neutron flux per lethargy Neutron flux per lethargy Neutron flux per lethargy 1 Y 1 r L Y 1 1 L 25 1 5190 5 9 5 92 5 93 aot 5 99 510988 Sao Neutron energy eV b JAEA Fusio
17. 440 2 5119E 0 618 6 9183E 4 85 1 5000E 7 263 8 7096E 3 441 2 3988E 0 619 6 6069E 4 86 1 4800E 7 264 8 3176E 3 442 2 2909E 0 620 6 3096E 4 87 1 4600E 7 265 7 9433E 3 443 2 1878E 0 621 6 0256E 4 continued on next page CCFE Page 184 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FISPACT II User Manual continued from previous page CCFE 709 group structure erp energy eV grp energy eV grp energy eV grp energy eV 88 1 4400E 7 266 7 5858E 3 444 2 0893E 0 622 5 7544E 4 89 1 4200E 7 267 7 2444E 3 445 1 9953E 0 623 5 4954E 4 90 1 4000E 7 268 6 9183E 3 446 1 9055E 0 624 5 2481E 4 91 1 3800E 7 269 6 6069E 3 447 1 8197E 0 625 5 0119E 4 92 1 3600E 7 270 6 3096E 3 448 1 7378E 0 626 4 7863E 4 93 1 3400E4 7 271 6 0256E 3 449 1 6596E 0 627 4 5709E 4 94 1 3200E 7 272 5 7544E 3 450 1 5849E 0 628 4 3652E 4 95 1 3000E 7 273 5 4954E 3 451 1 5136E4 0 629 4 1687E 4 96 1 2800E 7 274 5 2481E 3 452 1 4454E 0 630 3 9811E 4 97 1 2600E 7 275 5 0119E 3 453 1 3804E 0 631 3 8019E 4 98 1 2400E 7 276 4 7863E 3 454 1 3183E 0 632 3 6308E 4 99 1 2200E 7 277 4 5709E 3 455 1 2589E 0 633 3 4674E 4 100 1 2000E 7 278 4 3652E 4 3 456 1 2023E 0 634 3 3113E 4 101 1 1800E 7 279 4 1687E 3 457 1 1482E 0 635 3 162
18. 507 1 5849E 3 46 2 6303E 6 200 2 1878E 3 354 1 8197E 0 508 1 5136E 3 47 2 5119E 6 201 2 0893E 3 355 1 7378E 0 509 1 4454E 3 48 2 3988E 6 202 1 9953E 3 356 1 6596E 0 510 1 3804E 3 49 2 2909E4 6 203 1 9055E 3 357 1 5849E 0 511 1 3183E 3 50 2 1878E 6 204 1 8197E 3 358 1 5136E 0 512 1 2589E 3 51 2 0893E 6 205 1 7378E 3 359 1 4454E4 0 513 1 2023E 3 52 1 9953E 4 6 206 1 6596E 3 360 1 3804E 0 514 1 1482E 3 53 1 9055E 6 207 1 5849E 3 361 1 3183E 0 515 1 0965E 3 54 1 8197E 6 208 1 5136E 3 362 1 2589E 0 516 1 0471E 3 55 1 7378E 6 209 1 4454E 3 363 1 2023E 0 517 1 0000E 3 56 1 6596E 6 210 1 3804E 3 364 1 1482E 0 518 9 5499E 4 57 1 5849E 6 211 1 3183E 3 365 1 0965E 0 519 9 1201E 4 58 1 5136E 6 212 1 2589E 3 366 1 0471E 0 520 8 7096E 4 59 1 4454E 6 213 1 2023E 3 367 1 0000E 0 521 8 3176E 4 60 1 3804E 6 214 1 1482E 3 368 9 5499E 1 522 7 9433E 4 61 1 3183E 6 215 1 0965E 3 369 9 1201E 1 523 7 5858E 4 62 1 2589E 6 216 1 0471E 3 370 8 7096E 1 524 7 2444E 4 63 1 2023E 6 217 1 0000E 3 371 8 3176E 1 525 6 9183E 4 64 1 1482E 6 218 9 5499E 2 372 7 9433E 1 526 6 6069E 4 65 1 0965E 6 219 9 1201E 2 373 7 5858E 1 527 6 3096E 4 66 1 0471E 6 220 8 7096E 2 374 7 2444E 1 528 6 0256E 4 67 1 0000E 6 221 8 3176E 2 375 6 9183E 1 529 5 7544E 4 68 9 5499E 5 222 7 9433E 2 376 6 6069E 1 530 5 4954E 4 69 9 12
19. Bateman The Solution of a system of Differential Equations Occurring in the Theory of Radio active Transformations Proc Camb Phil Soc 16 423 1910 http en wikipedia org wiki adjacency matrix http en wikipedia org wiki directed graph J Ch Sublet and P Ribon A Probability Table Based Cross Section Processing System CALENDF 2001 J Nuc Sci Tech Supplement 2 856 859 August 2002 LI Bondarenko editor Group Constants for Nuclear Reactor Calculations Con sultant Bureau New York 1964 G I Bell and S Glasstone Nuclear Reactor Theory Van Nostrand Reinhold New York 1970 V Gopalakrishnan and S Ganesan Self Shielding and Energy Dependence of Dilution Cross Section in the Resolved Resonance Region Ann Nucl Energy 25 11 839 857 1998 N P Baumann Resonance integrals and self shielding factors for detector foils Technical Report DP 817 du Pont de Nemours Savannah River Laboratory Jan uary 1963 A J Koning and D Rochman Modern Nuclear Data Evaluation with the TALYS Code System Nuclear Data Sheets 113 2 2841 2934 Dec 2102 D Rochman A J Koning J Kopecky J C Sublet P Ribon and M Moxon From average parameters to statistical resolved resonances Ann Nuc Sci 51 60 68 2013 F H Fr hner Evaluation and analysis of nuclear resonance data Technical report Nucear Energy Agency 2000 JEFF Report 18 R G Jaeger editor Engineering Compendium on Radiation Shielding Spring
20. DEC 99 00X 100 99 EEF121M GP END END OF RUN The NOHEADER keyword suppresses the heading information in the output SPEK causes approximate y spectra to be generated for nuclides in the decay library that have no spectral data GETDECAY 1 causes the decay data to be read from the EAF library files connected to decay by the files file Note that when a library file has the extension 001 then the program will search the library directory for files with the same root and extensions 002 003 etc and add them to the input queue Keyword FISPACT marks the end of the library processing section of the input and END marks the end of input To do the run to condense the decay data type fispact condense and you should get at the terminal window a message of the form condense cpu time 0 639 secs No errors warnings and the program will have generated ascii output files condense log condense out and binary file condensed decay and fission data Figure 2 shows the input and output files used in the condense run example where the file unit names are mapped to real files according to the specifications in the files file Note that a summary of the files used and their mapping is written to the output and log files for all runs to provide a quality record 4 4 Library Summary Printing The library summary print example print lib uses the binary files containing cross section and decay data generated by the collapse and condense r
21. Sc V V Mn Mn Mn Mn Mn Fe Co 37 38 39 44 43 44m 45m 45m 47 46 49 ST 51 57 59 60 54 55 mt 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 old sigma barns 96858E 03 93047E 04 26196E 03 17911E 03 13298E 03 72141E 04 10641E 03 44849E 04 62702E 02 07775E 02 70205E 03 84027E 02 84192E 03 71788E 03 37615E 03 04988E 03 84688E 03 97518E 03 58414E 03 new sigma barns 87209E 03 05455E 04 71234E 03 36584E 03 04557E 03 84122E 04 57961E 03 33646E 04 10540E 02 33420E 03 86722E 03 28603E 02 66382E 03 56122E 03 27235E 03 28837E 03 62945E 03 19275E 03 06991E 03 68 self shielding factor 63 81 83 89 82 75 58 63 67 05 78 85 82 EA 85 87 88 84 88 5 06 11 15 06 27 65 62 16 94 91 57 78 77 04 41 23 27 78 CCFE Page 125 of CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT 1 User Manual This page has been left intentionally blank CCFE Page 126 of 200 REFERENCES CCFE R 11 11 Issue 6 FISPACT I User Manual References 1 12 13 R A Forrest FISPACT 2007 User Manual Technical Report UKAEA FUS 534 EURATOM UKAEA Fusion Association March 2007 J W Eastwood and J G Morgan Fortran 95 Conventions for FISPACT Techni cal Report CEM 081203 WI 1 Issue 2 Culham Electromagnetics Ltd F
22. This gives the user the flexibility to manipulate the dilutions but in general one should specify the same mixture for the inventory run as is used for the collapse run If subsequent collapses are requested by GETXS keywords then additional SSFFUEL or SSFMASS keywords will be needed for them To illustrate the usage we consider the following input file for a cross section collapse calculation using probability table data GETXS 1 616 PROBTAB 1 1 SSFCHOOSE 4 0 W182 W183 W184 W186 SSFFUEL 4 W182 1 34834187E 22 W183 7 27597094E 21 W184 1 55899050E 22 CCFE Page 121 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual W186 1 44654079E 22 FISPACT COLLAPSE EAF 616 FLT with PT for W END END OF RUN The first keyword specifies collapse of the 616 energy group EAF cross section data the EAF group structure for which probability table data are presently available The second activates the reading of probability table data to compute self shielding factors and the use of partial cross sections to compute self shielding factors with the infinite dilution values in the EAF library being replaced rather than scaled The SSFCHOOSE keyword specifies 4 entries on the list and suppresses detailed printing 0 The subsequent line or lines list the elements or nuclides In this case they are four isotopes of tungsten The SSFFUEL keyword specifies the mixture In this case all the nuclides
23. crossec EAF2007data eaf_n_gxs_069_fis_20070 crossunc EAF2007data eaf_un_20070 output collapxo COLLAPX 03 The inventory run using these collapsed cross sections uses GETXS in the input file to replace the collapsed cross sections as required lt lt physical data from condensed library gt gt GETXS 0 lt lt get cross section from first COLLAPX 01 file in files gt gt GETDECAY 0 lt lt get decay data from ARRAYX gt gt FISPACT lt lt first part using COLLAPX 01 gt gt TIME 6 109E 06 DAYS SPECTRUM lt lt second part using COLLAPX 02 gt gt GETXS 0 FLUX 2 64634E 14 TIME 6 108994E0 DAYS SPECTRUM lt lt third part using COLLAPX 03 gt gt GETXS 0 FLUX 2 66930E 14 TIME 2 44410E 01 DAYS CCFE Page 89 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual ATOMS where now the files file contains the queue of collapsed cross section files with one collapxi for each GETXS 0 in the input file collapsed cross sections queue collapxi COLLAPX 01 collapxi COLLAPX 02 collapxi COLLAPX 03 Complete examples of this usage are given in the test directory fispQA2010 Tst burn 5 3 8 GRPCONVERT nestre ndstrc The primary use of this keyword is in the library data preparation section of the input file see Section Its use in the inventory section is in conjunction with GETXS 5 8 9 LOGLEVEL level 2 See Section on page 51 for more information
24. g 2 9 Ed 1e 13 F 5n 4 1e 12 L L L 1 L L 1e 06 1e 05 0 0001 0 001 0 01 0 1 1 10 Time after irradiation years file name inventory gra run timestamp 10 16 28 19 February 2014 Figure 3 The total activity graph produced by the inventory run Figure 4 shows the input and output files used in the inventory run example where the file unit names are mapped to real files according to the specifications in the files CCFE Page 37 of 200 CCFE R 11 11 Issue 6 4 GETTING STARTED FISPACT II User Manual file Figure 4 The files used by FISPACT II in the inventory run example An explanation of the contents of the output file may be found in Section 7 1 4 6 ENDF format Library Data The ENDF format of library data are provided in directories containing separate files for each nuclide These directories are made available to FISPACT II by including the stream names prob tab xs_endf dk endf and fy_endf in the files file as listed in Table FISPACT II is directed to use these new libraries by including the EAFVERSION keyword with argument 8 near the head of the input file before the GETXS keyword for example EAFVERSION 8 GETXS 1 162 then the remainder of the input file is unchanged The number of energy groups must be consistent with the library data currently 162 and 709 energy group libraries are provided in ENDF format together with a translation of the EAF 616 group data into ENDF format
25. mass of source kg r distance from source m u E energy attenuation coefficient of air m Both Equations and are approximations suitable for FISPACT II calculations but it is noted that they may not be adequate for specific health physics problems A 10 3 Approximate gamma spectrum Wherever possible decay data from JEFF 3 1 files have been used to construct the decay data library decay see Appendix B 4 used with FISPACT II Intensity in a spectrum energy group is computed from the sum of intensities of discrete spectral lines lying in the energy group However for 254 unstable nuclides the file contains only the average y energy no data for the y spectrum are available Without the y spectrum FISPACT II is unable to calculate the y dose rate contribution for these nuclides In order to check if any of these nuclides are likely to contribute significantly to the total dose rate the following method is used to calculate an approximate spectrum see SPEK keyword on page 54 The maximum y energies Em for decays assumed in the method are given in Table 17 The intensity in the i th group J is given by aly e 2 1 Mi LE 32 62 CCFE Page 154 of 200 A 10 Gamma Radiation CCFE R 11 11 Issue 6 FisPACT II User Manual Table 17 Maximum y energies for various decay modes Decay mode Em 8 2 B Bt 5 MeV a 0 Isomeric transition y where a 14 arbitrary constant
26. 0 2eV 0 82085 MeV 1 E 0 82085 MeV Emax Maxwellian fission spectrum T 1 3539 MeV It is important not to have any fusion peak in order not to bias the high threshold reactions such as n Xn One may also keep in mind that the fission spectrum has a tail that extends well above 10 MeV The weighting spectra used to generate fusion relevant libraries in the VITAMIN J GAM II and TRIPOLI group format from EAF point wise data are as follows Energy range Micro flux weighting spectrum 1 0 x 107 0 414eV Maxwellian T 0 0253 eV 0 414 eV 12 52 MeV 1 E 12 52 MeV 15 68 MeV Velocity exponential fusion peak Ey 14 07 MeV kT 0 025 MeV 15 68 MeV 19 64 MeV 1 E A flat weighting spectrum is used to generate multi purpose libraries from EAF point wise data in the various group formats and in these cases the finer the structure the better Such libraries could be used to model cases where the neutron field is not similar to one described above for example from accelerator beam target interactions e g IFMIF or experimental devices Such libraries also allow group wise data to be plotted without weighting It is the user s responsibility to select the appropriate group wise library depending on the type of activation transmutation calculations that will be made The micro flux weighting process can have a significant impact on the cross sections B 2 Cross section Data This section gives a brief summa
27. 0 or 1 It is not used by FISPACT II but is retained for backwards compatibility with FIsPACT 2007 If output from more pathways is required then increase nmaz and decrease pmin The ZERO or RESULT keyword initiates the calculation of the routes over all the time intervals before its occurrence An example of the use of this keyword is ROUTES Ti46 Sc44 5 1E14 0 RESULT 1 Sc44 1 00621E15 The output for a run using these commands gave ROUTES ANALYSIS FOR IRRADIATION PHASE L 7 88940E 07 secs no of steps irradiation time flux 4 27701E 14 n cm 2 s path floor 9 93829E 00 of target inventory loop floor 1 00000E 00 of path inventory max depth 5 maximum number of edges between source and target Source Nuclides Ti 46 Target Nuclides Sc 44 CCFE Page 78 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual Target nuclide Sc 44 43 147 of inventory given by 3 paths path 1 9 768 Ti 46 R Sc 45 R Sc 44 S 98 16 n np 00 00 n 2n 1 84 n d path 2 2 392 Ti 46 R Sc 45 R Sc 44m b Sc 44 S 98 16 n np 00 00 n 2n 100 00 IT 1 84 n d 0 00 n n path 3 0 987 Ti 46 R Sc 45m d Sc 45 R Sc 44 S 96 62 n np 00 00 IT 100 00 n 2n 3 38 n d GENERIC ROUTES FOR IRRADIATION PHASE Target nuclide Sc 44 43 147 of inventory given by 1 path path 1 43 147 Ti 46 R Sc 4
28. 00 0 0000E 00 0 0000E 00 5763E 21 4 2780E 03 7123E 02 0 0000E 00 0000E 00 0 0000E 00 0 0000E 00 0 0000E 00 2358E 09 5 3732E 15 6928E 13 1 7214E 08 0204E 13 1 8729E 08 1 11025E 13 0 0000E 00 4013E 09 3 9875E 15 2675E 13 3 5813E 08 1229E 13 9 9559E 08 5 9016E 13 0 0000E 00 7 1 6 Gamma spectrum In this section the total powers MeV s from a 8 and y radiations and the total number of spontaneous fission neutrons are listed followed by two columns giving the y spectrum MeV s per group and number of gammas per group cm s71 in either a 24 or 22 group form depending on the parameter used for GROUP GAMMA SPECTRUM AND ENERGIES SECOND NEUTRONS PER SECOND ARISING FROM SPONTANEOUS FISSION 00000E 00 Calculated density g cc POWER FROM ALPHA PARTICLES MeV per Second 37839E 02 POWER FROM BETA PARTICLES MeV per Second 19112 13 TOTAL GAMMA POWER FROM ACTIVATION MeV per Second 02831E 14 Total gammas per cc per second 51308E411 GAMMA RAY POWER FROM ACTIVATION DECAY MeV s 05782E 09 Gammas per group per cc per second 5 50050E 09 88216E 08 69668E 07 93236E 05 28837E405 71468E 02 27528E 01 10646E 12 40223E410 36863E 10 30143E 08 00000E 00 00000E400 00000E 00 00000E400 00000E 00 00000E400 The total dose rate is then given in one of two forms depending on the DOSE param eter these two outputs are for contact dose from a semi infinite slab of the material PLANE SOURCE see Equation 58 o
29. 1 460 1 3804E 2 614 1 1482E 5 153 1 9055E 4 307 1 5849E 1 461 1 8183E 2 615 1 0965E 5 154 1 8197E4 4 308 1 5136E 1 462 1 2589E 2 616 1 0471E 5 617 1 0000E 5 Table 25 Energy group boundaries for the CCFE 709 group structure CCFE 709 group structure erp energy eV grp energy eV grp energy eV grp energy eV 1 1 0000E 9 179 4 1687E 5 357 1 1482E 2 535 3 1623E 2 2 9 6000E 8 180 3 9811E 5 358 1 0965E 2 536 3 0200E 2 3 9 2000E 8 181 3 8019E 5 359 1 0471E 2 537 2 8840E 2 4 8 8000E 8 182 3 6308E 5 360 1 0000E 2 538 2 7542E 2 5 8 4000E 8 183 3 4674E 5 361 9 5499E 1 539 2 6303E 2 6 8 0000E 8 184 3 3113E 5 362 9 1201E 1 540 2 5119E 2 7 7 6000E 8 185 3 1623E 5 363 8 7096E 1 541 2 3988E 2 8 7 2000E 8 186 3 0200E 5 364 8 3176E 1 542 2 2909E 2 9 6 8000E 8 187 2 8840E 5 365 7 9433E 1 543 2 1878E 2 10 6 4000E 8 188 2 7542E 5 366 7 5858E 1 544 2 0893E 2 11 6 0000E 8 189 2 6303E 5 367 7 2444E 1 545 1 9953E 2 12 5 6000E 8 190 2 5119E4 5 368 6 9183E 1 546 1 9055E 2 13 5 2000E 8 191 2 3988E 5 369 6 6069E 1 547 1 8197E 2 14 4 8000E 8 192 2 2909E4 5 370 6 3096E 1 548 1 7378E 2 15 4 4000E 8 193 2 1878E 5 371 6 0256E 1 549 1 6596E 2 16 4 0000E 8 194 2 0893E4 5 372 5 7544E 1 550 1 5849E 2 17 3 6000E 8 195 1 9953E 5 373 5 4954E 1 551 1 5136E 2 18 3 2000E 8 196 1 905
30. 10096 CCFE Page 70 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual 5 2 28 MCSAMPLE distrib 1 nsamples 10 lb 3 0 ub 3 0 This keyword is used in conjunction with the SENSITIVITY keyword to change the sampling used in the Monte Carlo calculation In the Monte Carlo calculation values of cross sections are randomly selected from a distribution with a mean and standard deviation given by the value and uncertainty specified The first argument distrib is an integer that specifies the distribution to be used log normal normal uniform log uniform gt QwQ0wNR nsamples is the number of Monte Carlo samples per parent daughter pair specified by SENSITIVITY lb and ub give the cutoffs for the log normal and normal distri butions For the log normal distribution these define the range in multiples of the standard deviation from the logarithm of the mean at which the logarithm of the sam ple is accepted For the normal distribution they define the range in multiplies of the standard deviation from the mean at which the sample is accepted lb and ub are not used for the uniform and log uniform distributions 5 2 29 MCSEED dim seed i i 1 dim This keyword is provided to allow repeatable selections of pseudo random numbers to be made by specifying the random number seed FISPACT II uses the intrinsic Fortran 95 random number generator and the dimension of the seed depends on the compiler used di
31. 129 9 1188E4 3 60 9 1187E4 156 8 2510E 3 157 7 4660E 3 47 7 4659E 3 158 7 1020E 3 130 7 1017E 3 61 7 1016E 3 159 6 2670E 3 3 46 9 1188E43 15 9 1180E 3 continued on next page CCFE Page 173 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT II User Manual continued from previous page TRIPOLI 315 VITAMIN J 175 GAMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 160 5 5310E 3 131 5 5308E4 3 62 5 5308E 3 48 5 5308E4 3 16 5 5300E 3 161 5 0040E 3 49 5 0045E 3 162 4 6430E 3 163 4 3070E 3 132 4 3074E 4 164 3 9810E 3 165 3 7070E 3 133 3 7074E 3 166 3 5480E 3 50 3 5266E4 3 17 3 5190E4 3 167 3 3550E 3 134 3 3546E4 3 64 3 3546E43 51 3 3546E 3 168 3 1620E 3 169 3 0350E 3 135 3 0354E4 3 170 2 8180E 3 171 2 7470E 3 136 2 7465E 3 172 2 6610E 3 173 2 6130E 3 137 2 6126E4 3 65 2 6125E43 174 2 4850E 3 138 2 4852E 3 175 2 3710E 3 176 2 2490E 3 139 2 2487E4 3 52 2 2487E 3 18 2 2390E 3 177 2 1130E43 178 2 0350E 3 140 2 0347E 3 66 2 0347E 3 53 2 0347E 3 179 1 7960E 3 180 1 5850E 3 141 1 5846E 3 67 1 5846E43 181 1 5070E 3 54 1 5073E 3 55 1 4338E 3 19 1 4250E 3 3 63 4 3074E 3 T T T 182 1 3640E4 3 183 1 2340E 3 142 1 2341E 3 68 1 2341E 3 56 1 2341E 3 184 1 1170E 3 185 1 0100E 4 3 57 1 0104E4 3 186 9 6110E4 2 143 9 6112E4 2 69 9 6110E 2 58 9 1424E
32. 2 Let ing and ig be the largest i for which E gt E and E gt E ft respectively then inf 1 bh Y oF 50 i 1 ife l oy gt gt OB 51 i in Y AE 52 iip 9 r rt di 53 and the collapsed fission yield is given by Y nYs fYz HY 0 54 In terms of the description of the reaction network as a directed graph each fission reaction gives rise to many edges in the graph connecting the fissionable parent nuclide to all of its possible fission products The effective reaction cross section needed to calculate the flow along each edge of the graph is simply the fission cross section multiplied by the appropriate fission yield To make up for the lack of data on fission for many actinides a surrogate daughter algorithm is used This is in addition to the use of associated fission yield data The surrogate daughter algorithm replaces fission product daughters not known to the program with similar nuclides that are known The algorithm works as follows e If the daughter fission product is in the list of nuclides known to the program then assign the fission yield to that daughter e If the daughter is not listed then assign its yield to the first nuclide encountered in the list of nuclides with the same A and the same or larger Z e If neither of the above cases is satisfied then assign the yield to the sink nuclide CCFE Page 151 of 200 CCFE R 11 11 Issue 6 A THE MODEL FIS
33. 22 eV sec 97704E 22 eV sec 00000E 00 eV sec 55007E422 eV sec 65720E 22 eV sec 05826E 20 eV sec 70779E 22 eV sec 78414E 02 kW cm 27880E 02 kW cm 00000E 00 kW cm 95549E 02 kW cm 13687E 02 kW cm 38127E 04 kW cm 20068E 02 kW cm KERMA RATE n Kktot KERMA RATE n Kphot KERMA RATE n Kfiss 61588E 00 kW kg KERMA RATE n Kinel 97409E 00 kW kg 00000E 00 kW kg 48348E 00 kW kg 06384E 00 kW kg 10422E 02 kW kg 14488E 00 kW kg KERMA RATE n Knone KERMA RATE n Kel KERMA RATE n Ktot ow wd wo wy ow ow C Ur O1 IB O iS dr noone ow ow o 0 Q t9 O 1 00 ow wd ow ow Ww ow SDIPODO w w w CO www GAS RATE n Xa GAS RATE n Xh GAS RATE n Xt 47206E 13 atoms per sec 50099E 08 atoms per sec 20388E 09 atoms per sec 16080E 13 atoms per sec 15739E 14 atoms per sec 92906E 06 appm sec 39191bE 11 appm sec 97105E 10 appm sec 07644E 06 appm sec 85527E 05 appm sec GAS RATE n Xd GAS RATE n Xp ow wd ow ow way ow d on on WENED The displacements per atom DPA for a single element is given by 26 Eq 90 For mixtures of elements with different lattice displacement energies the total displace ments rate Dict may be estimated using the ratio of the mean total available energy to the mean displacement energy Nn Dior eab gt Nidi 2Ea 1 i l where is the flux amplitude in cm s N is the number of atoms of nuclide i and d is the collap
34. 3096E 6 181 5 2481E4 3 335 4 3652E 0 489 3 6308E 3 28 6 0256E 6 182 5 0119E 3 336 4 1687E 0 490 3 4674E 3 29 5 7544E4 6 183 4 7863E4 3 337 3 9811E 0 491 3 3113E 3 30 5 4954E46 184 4 5709E 3 338 3 8019E 0 492 3 1623E 3 31 5 2481E46 185 4 3652E 3 339 3 6308E 0 493 3 0200E 3 32 5 0119E46 186 4 1687E 3 340 3 4674E 0 494 2 8840E 3 33 4 7863E 6 187 3 9811E43 341 3 3113E 0 495 2 7542E 3 34 4 5709E 6 188 3 8019E 3 342 3 1623E 0 496 2 6303E 3 35 4 3652E 6 189 3 6308E 3 343 3 0200E 0 497 2 5119E 3 36 4 1687E 6 190 3 4674E 3 344 2 8840E 0 498 2 3988E 3 37 3 9811E46 191 3 3113E4 3 345 2 7542E 0 499 2 2909E 3 38 3 8019E 6 192 3 1623E 3 346 2 6303E 0 500 2 1878E 3 39 3 6308E 6 193 3 0200E 3 347 2 5119E 0 501 2 0893E 3 40 3 4674E 6 194 2 8840E 3 348 2 3988E 0 502 1 9953E 3 continued on next page CCFE Page 180 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FISPACT II User Manual continued from previous page LLNL 616 group structure erp energy eV grp energy eV grp energy eV grp energy eV 41 3 3113E4 6 195 2 7542E 3 349 2 2909E 0 503 1 9055E 3 42 3 1623E4 6 196 2 6303E 3 350 2 1878E 0 504 1 8197E 3 43 3 0200E4 6 197 2 5119E 3 351 2 0893E 0 505 1 7378E 3 44 2 8840E 6 198 2 3988E 3 352 1 9953E 0 506 1 6596E 3 45 2 7542E 6 199 2 2909E 3 353 1 9055E 0
35. 4 0870E 4 39 3 7270E 1 56 4 2000E 2 6 6 0650E 6 23 2 5540E 4 40 2 2600E 1 57 3 5000E 2 continued on next page CCFE Page 178 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FisPACT II User Manual continued from previous page LANL 66 group structure grp energy eV grp energy eV grp energy eV grp energy eV 7 49658E 6 24 1 9890E 4 41 1 3710E 1 58 3 0000E 2 8 3 6788E 6 25 1 5030E 4 42 8 3150E 0 59 2 5000E 2 9 2 8650E 6 26 9 1190E 3 43 5 0430E 0 60 2 0000E 2 10 2 2313E 6 27 5 5310E 3 44 3 0590E 0 61 1 5000E 2 11 1 7377E 6 28 3 3550E 3 45 1 8550E 0 62 1 0000E 2 12 1 3534E46 29 2 8400E43 46 1 1250E 0 63 5 0000E 3 13 1 1080E4 6 30 2 4040E43 47 6 8300E 1 64 2 0000E 3 14 8 2085E 5 31 2 0350E 3 48 4 1400E 1 65 1 0000E 3 15 6 3928E 5 32 1 2340E 3 49 2 5100E 1 66 5 0000E 4 16 4 9790E45 33 7 4850E42 50 1 5200E 1 67 1 0000E 5 17 3 8870E4 5 34 4 5400E 2 51 1 0000E 1 Table 23 Energy group boundaries for the CCFE 162 group structure CCFE 162 group structure grp energy eV grp energy eV grp energy eV grp energy eV 1 1 0000E 9 42 2 0000E 7 83 3 8750E 6 123 5 7500E 4 5 2 9 6000E 8 43 1 9000E 7 84 3 7500E 6 124 5 5000E 5 3 9 2000EX8 44 1 8000E 7 85 3 6250E 6 125 5 2500E 5 4 8 8000E 8 45 1 7000E 7 86 3 5000E 6 126 5 0000E
36. 5 5 84000E 8 46 1 6000E 7 87 3 3750E 6 127 4 7500E 5 6 8 0000E 8 47 1 5000E 7 88 3 2500E 6 128 4 5000E 5 7 7 6000E 8 48 1 4000E 7 89 3 1250E 6 129 4 2500E 5 8 7 2000EX8 49 1 3000E 7 90 3 0000E 6 130 4 0000E 5 9 6 8000E 8 50 1 2000E 7 91 2 8750E X 6 131 3 7500E 4 5 10 6 4000E 8 51 1 1000E 7 92 2 7500E46 132 3 5000E4 5 11 6 0000E 8 52 1 0000E 7 93 2 6250E 6 133 3 2500E4 5 12 5 6000E 8 53 9 8000E 6 94 2 5000E 6 134 3 0000E4 5 13 5 2000E 8 54 9 6000E 6 95 2 3750E 6 135 2 8000E 5 14 4 8000E 8 55 9 4000E 6 96 2 2500E 6 136 2 6000E 5 15 4 4000E 8 56 9 2000E 6 97 2 1250E 6 137 2 4000E 5 16 4 0000E 8 57 9 0000E 6 98 2 0000E 6 138 2 2000E4 5 17 3 6000E 8 58 8 8000E 6 99 1 8750E4 6 139 2 0000E 5 18 3 2000E 8 59 8 6000E 6 100 1 7500E 6 140 1 8000E4 5 19 2 8000E 8 60 8 4000E 6 101 1 6250E 6 141 1 6000E4 5 20 2 4000E 8 61 8 2000E 6 102 1 5000E4 6 142 1 4000E4 5 21 2 0000E 8 62 8 0000E 6 103 1 3750E 6 143 1 2000E4 5 22 1 8000E 8 63 7 8000E 6 104 1 2500E 6 144 1 0000E4 5 23 1 6000E 8 64 7 6000E 6 105 1 1250E 6 145 9 5000E4 4 24 1 4000E 8 65 7 4000E 6 106 1 0000E 6 146 9 0000E 4 25 1 2000E 8 66 7 2000E 6 107 9 7500E4 5 147 8 5000E 4 26 1 0000E 8 67 7 0000E 6 108 9 5000E4 5 148 8 0000E 4 27 9 0000E 7 68 6 8000E 6 109 9 2500E4 5 149 7 5000E 4 28 8 0000E4 7 69 6 6000E 6 110 9 0000E 5 150 7 0000
37. 5709E 5 144 2 0893E4 6 322 5 7544E 2 500 1 5849E 1 678 4 3652E 5 145 1 9953E 6 323 5 4954E 2 501 1 5136E 1 679 4 1687E 5 146 1 9055E 6 324 5 2481E 2 502 1 4454E 1 680 3 9811E 5 147 L8197E46 325 5 0119E 2 503 1 3804E 1 681 3 8019E 5 148 1 7378E 6 326 4 7863E 2 504 1 3183E 1 682 3 6308E 5 149 1 6596E4 6 327 4 5709E 2 505 1 2589E 1 683 3 4674E 5 150 1 5849E4 6 328 4 3652E 2 506 1 2023E 1 684 3 3113E 5 151 1 5136E46 329 4 1687E 2 507 1 1482E 1 685 3 1623E 5 152 1 4454E46 330 3 9811E 2 508 1 0965E 1 686 3 0200E 5 153 1 3804E 6 331 3 8019E 2 509 1 0471E 1 687 2 8840E 5 154 1 3183E 6 332 3 6308E 2 510 1 0000E 1 688 2 7542E 5 155 1 2589E46 333 3 4674E 2 511 9 5499E 2 689 2 6303E 5 156 1 2028E46 334 3 3113E 2 512 9 1201E 2 690 J2 5119E 5 157 1 1482E 6 335 3 1623E42 513 8 7096E 2 691 2 3988E 5 158 1 0965E 6 336 3 0200E 2 514 8 3176E 2 692 2 2909E 5 159 1 0471E 6 337 2 8840E 2 515 7 9433E 2 693 2 1878bE 5 160 1 0000E 6 338 2 7542E 2 516 7 5858E 2 694 2 0893E 5 161 9 5499E4 5 339 2 6303E 2 517 7 2444E 2 695 1 9953E 5 162 9 1201E 5 340 2 5119E 2 518 6 9183E 2 696 1 9055E 5 163 8 7096E4 5 341 2 3988E 2 519 6 6069E 2 697 1 8197E 5 164 8 3176E 5 342 2 29009E 2 520 6 3096E 2 698 1 7378E 5 165 7 9433E4 5 343 2 1878E 2 521 6 0256E 2 699 1 6596E 5 166 7 5858E 5 344 2 0893E 2 522 5 7544E 2
38. 6 24 1 4957E4 6 56 1 4230E4 6 57 1 4227E 6 57 1 3530E 6 58 1 3534E 6 25 1 3533E 6 18 1 3534E46 5 1 3530E 6 continued on next page CCFE Page 171 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT II User Manual continued from previous page TRIPOLI 315 VITAMIN J 175 GAMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 58 1 2870E 6 59 1 2874E 6 59 1 2250E 6 60 1 2246E 6 26 1 2245E 6 19 1 2246E 6 60 1 1650E 6 61 1 1648E 6 61 1 1080E 6 62 1 1080E 6 27 1 1080E 6 20 1 1080E 6 62 1 0030E 6 63 1 0026E 6 28 1 0026E 6 21 1 0026E 6 63 9 6160E 5 64 9 6167E4 5 64 9 0720E 5 65 9 0718E4 5 29 9 0716E 5 22 9 0718E 5 65 8 6290E 5 66 8 6294E4 5 66 8 2090E 5 67 8 2085E4 5 30 8 2084E4 5 23 8 2085E4 5 6 8 2100E 5 67 T 8080E4 5 68 7 8082E 5 68 7 4270E 5 69 7 4274E 5 31 7 4272E 5 69 7 0650E 5 70 7 0651E 5 70 6 7210E 5 71 6 7206E4 5 32 6 7204E4 5 T1 6 3930E 5 72 6 3928E4 5 72 6 0810E 5 73 6 0810E4 5 33 6 0809E 5 24 6 0810E 5 73 5 7840E 5 74 5 7844E4 5 74 5 5020E 5 75 5 5023E4 5 34 5 5022E 5 25 5 5023E 5 7 5 0000E 5 75 5 2340E 5 76 5 2340E4 5 77 4 9787E 5 35 4 9786E 5 26 4 9787E 5 76 4 5050E 5 78 4 5049E 5 36 4 5048E4 5 27 4 5049E 5 77 4 0760E 5 79 4 0762E 5 37 4 0762E45 28 4 0762E 5 78 3 8770E 5 80 3 8774E 5 79 3 6880E 5
39. 600 849 Various proton production reactions 851 859 Lumped reaction covariances 875 891 Various double neutron productions CCFE Page 143 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual Dpa displacements per atom The resulting dataset can be defined in terms of an MT number and may be read into FISPACT II and used in subsidiary calculations during inventory runs to quantify the damage to materials caused by neutron irradiation See the end of Section 7 1 4 for a description of the output of kerma dpa and appm rates A 4 2 Ignored reactions The new ENDF style libraries of cross section data may contain MT values not included in Tables and 13 Data for the MT numbers shown in Table 14 are silently ignored Data for any other MT encountered cause warning messages to be issued A 4 3 Self shielding of resonant channels using probability tables The probability tables keyword in FISPACT II see Section allows probability table data generated by CALENDF to be used to model dilution effects in the computation of the collapsed effective cross sections CALENDF provides data in five sets of macro partial cross sections The CALENDF set MT numbers cal mt are defined in Table 15 The sum of these macro partial cross sections gives the total cross section in each energy group over the resonance regions covered Table 15 CALENDF MT number cal mt description MT in set 2 elastic s
40. 6071E 6 20 8 1870E 6 21 8 1873E 6 T 8 1872E 6 7 8 1873E 6 21 7 7880E 6 22 7 7880E 6 22 7 4080E 6 23 7 4082E 6 8 7 4081E 6 23 7 0470E 6 24 7 0469E 6 24 6 7030E 6 25 6 7032E 6 9 6 7031E 6 8 6 7032E 6 25 6 5920E 6 26 6 5924E4 6 26 6 3760E 6 27 6 3763E 6 27 6 0650E 6 28 6 0653E4 6 10 6 0652E 6 9 6 0653E 6 2 6 0660E 6 28 5 7690E 6 29 5 7695E4 6 29 5 4880E 6 30 5 4881E4 6 11 5 4880E4 6 10 5 4881E 6 30 5 2200E 6 31 5 2205E4 6 31 4 9660E4 6 32 4 9659E 6 12 4 9658E4 6 32 4 7240E 6 33 4 7237E 6 33 4 4930E 6 34 4 4933E4 6 13 4 4932E 6 11 4 4933E4 6 34 4 0660E 6 35 4 0657E 6 14 4 0656E 6 35 3 6790E 6 36 3 6788E 6 15 3 6787E 6 12 3 6788E 6 3 3 6790E 6 36 3 3290E 6 37 3 3287E 6 16 3 3287E 6 37 3 1660E4 6 38 3 1664E 6 38 3 0120E 6 39 3 0119E 6 17 3 0119E46 13 3 0119E 6 39 2 8650E 6 40 2 8651E4 6 40 2 7250E4 6 41 2 7253E 6 18 2 7253E 6 41 2 5920E 6 42 2 5924E4 6 42 2 4660E 6 43 2 4660E4 6 19 2 4659E4 6 14 2 4660E 6 43 2 3850E 6 44 2 3851E4 6 44 2 3650E 6 45 2 3653E4 6 45 2 3460E 6 46 2 345 7E4 6 46 2 3070E 6 47 2 3069E4 6 4T 2 2310E 6 48 2 2313E 6 20 2 2313E 6 15 2 2313E 6 4 2 2310E 6 48 2 1220E 6 49 2 1225E4 6 49 2 0190E 6 50 2 0190E4 6 21 2 0189E 6 16 2 0190E 6 50 1 9210E 6 51 1 9205E 6 51 1 8270E 6 52 1 8268E 6 22 1 8268E 6 52 1 7380E 6 53 1 7377E 6 53 1 6530E4 6 54 1 6530E 6 23 1 6530E 6 17 1 6530E 6 54 1 5720E 6 55 1 5724E 6 55 1 4960E 6 56 1 4957E
41. 8 52E 11 Bq Error is Total Heat Production is 3 60059E 02 3 09E 04 kW Error i Total Gamma Dose Rate is 5 63098E 04 04E 02 Sv hr Error i Total Ingestion Dose is 1 38528E 05 17E 03 Sv Error i Total Inhalation Dose is 06441E 05 40E 03 Sv Error i 44E 00 Total Gamma Heat Prod is 24955E 02 90E 04 kW Error i 92kE 01 Total Beta Heat Prod is 51040E 03 07E 05 kW Error i T4E 01 Nuclide Atoms E Atoms Activity E Activity Heat E Heat Dose Rate E Dose Rate Ingest E Ingest 9 90164E 18 8 60E 16 4 366E 13 3 79E 11 196E 02 2 17E 04 4 136E 04 3 59E 02 7 422E 04 6 45E 02 2 57782E 20 6 68E 18 2 468E 13 6 40E 11 389E 03 2 18E 04 1 363E 04 3 53E 02 3 702E 04 9 60E 02 8 09046E 18 1 21E 17 1 937E 13 2 89E 11 408E 04 1 26E 05 1 827E 02 2 73E 00 1 046E 04 1 56E 02 8 82538E 13 2 33E 12 5 968E 11 1 57E 10 611E 04 1 22E 05 5 452E 02 1 44E 01 1 552E 03 4 09E 01 Note that uncertainties that drop to zero are usually indicate that important pathways are being ignored The SORTDOMINANT LOOKAHEAD and PATHRESET keywords can be used to deal with this problem c f Section 5 2 59 7 1 11 Pathways Pathways analysis is initiated by the UNCERTAINTY keyword and is performed over all steps preceding the ZERO keyword the irradiation phase The initial path ways output summarises the steps over which the pathways calculations are performed and the criteria used in pruning the tree search for paths The example below from
42. 8 Compressed ENDF Library Files The TENDL nuclear data libraries are large they contain many gigabytes of data If a sequence of FISPACT II runs uses the same incident particle spectrum then the time for a single collapse run may be spread over many runs by using a preliminary collapse run c f Section 4 2 However if the sequence of runs uses different flux spectra in each run then the computational time for the collapses becomes significant particularly if data are accessed across a network To speed up calculations in these cases three capabilities have been added to FISPACT II The first described in this section is to preprocess the ENDF libraries and store only those data by FISPACT II in a single compressed binary file The second is to store the cross sections versus energy in FISPACT II and then perform a number of collapses without re reading the ENDF data The third approach to speedup is to use a reduced master index as described in the next section A separate executable compress xs endf is used to convert the ASCII ENDF libraries into a compressed binary file It has up to five arguments in the following order 1 the fileroot name used to construct binary output and log file names default compress xs endf 2 the projectile a letter that denotes the projectile used for the reaction data Valid values are n p d a g default n 3 the bin size number of energy bins in cross section data For the present TE
43. Ao 0 is the fractional uncertainty in the cross section for reaction r De is the set of edges on pathway p where the parent nuclide is long lived or where the parent is short lived and the daughter nuclide is the target nuclide of the path A short lived nuclide is one whose half life is less than the time interval of the irradiation pulse sequences Ate Te is the fractional uncertainty in the half life of the parent nuclide on the edge The total reaction rate for the edge is the sum of the reaction rates for the parent daughter nuclides on the edge zy R 78 rESr Let there be J time intervals in the irradiation phase and let the time of interval j 1 J be At and the flux amplitude be In addition assume that there are T different collapsed cross sections with cross section i 1 7 being the value used for pulses j J Ji41 J 1 and Jr44 J then D Ji 1 oe 6ASyT 79 i l j Ji where J T gt AE 80 In the case of fission reactions o is replaced by f fp where f is the fission yield for the reaction product corresponding to the daughter nuclide on the edge A 14 Method of Solution of Rate Equations The rate equations and subsets of the rate equations used for pathway calculations are all specific examples of first order systems of odes with the general form Oh Rilly 81 with initial conditions yi t 0 y o given 82 CCFE Page 162 of 200 A 14 Method of Solution of
44. Issue 3 Culham Elec tromagnetics Ltd September 2011 J W Eastwood and J G Morgan The FISPACT Phase 2 CVS Repository Tech nical Report CEM 081203 WI 2 Issue 1 Culham Electromagnetics Ltd February 2009 E Martinho LF Gongalves and J Salgado Universal curve of epithermal neu tron resonance self shielding factors in foils wires and spheres Appl Radiation Isotopes 58 3 371 375 March 2003 E Martinho J Salgado and LF Goncalves Universal curve of the thermal neutron self shielding factor in foils wires spheres and cylinders J Radioanal Nucl Chem 261 3 637 643 September 2004 J Salgado E Martinho and I F Gon alves The calculation of neutron self shielding factors of a group of isolated resonances J Radioanal Nucl Chem 260 2 317 320 May 2004 J Ch Sublet and M Gilbert Decay heat validation FISPACT II amp TENDL 2012 2011 and EAF 2010 nuclear data libraries Technical Report CCFE R 13 20 CCFE 2013 R A Forrest The European Activation System EAF 2007 biological clearance and transport libraries Technical Report UKAEA FUS 538 UKAEA 2007 R E MacFarlane and A C Kahler Methods for Processing ENDF B VII with NJOY Nuclear Data Sheets 111 12 2739 2890 December 2010 CCFE Page 128 of 200 REFERENCES CCFE R 11 11 Issue 6 FISPACT I User Manual 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 H
45. MF 2 resolved resonance range data CCFE Page 123 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual 2 The SSFGEOMETRY keyword must be added to the input file to trigger the use of this self shielding model 3 The target shape and size must be specified with arguments to SSFGEOME TRY 4 The mixture of nuclides whose resonances are to be included in the calculation of the self shielding factors must be specified using either the SSFMASS of SSFFUEL keywords Note that the values specified for these may be different from the MASS or FUEL mixtures specified in the inventory run This gives the user the flexibility to manipulate the self shielding factors but in general one should specify the same mixture for the inventory run as is used for the collapse run If subsequent collapses are requested by GETXS keywords then additional SSFFUEL or SSFMASS keywords will be needed for them To illustrate the usage we consider the following input file for a cross section collapse calculation using probability table data EAFVERSION 8 GETXS 1 709 SSFGEOMETRY 1 0 8 SSFFUEL 4 W182 1 34834187E 22 W183 7 27597094E 21 W184 1 55899050E 22 W186 1 44654079E 22 FISPACT COLLAPSE tal2011 n gxs 709 with universal curve SSF for a foil END END OF RUN The first keyword specifies that an ENDF format cross section library is to be read from the directory indicated by xs_endf in the files file and the
46. MONITOR with argument 1 indicates how far the run has gone and in the case of failure helps identify where things have gone wrong Once problems are ironed out then it is recommended that these keywords are removed from the input file CLOBBER allows the accidental overwriting of output files and MONITOR duplicates information that can be obtained from the runlog file in production runs Executing the command fispact testi42a generates the following output ERROR in INPUT file Unexpected eof encountered while reading the input file Check for missing closing gt gt on comment string beginning on line 33 test142a FATAL ERROR run terminated for details see runlog file test142a log Inspection of line 33 indeed reveals that the closing gt gt of the comment is missing and this caused the remainder of the input file to be treated as a comment Adding this missing comment closure saving the input file as test142b i and rerunning fispact then generates the error messages ERROR in INPUT file l CCFE Page 40 of 200 4 7 Developing New input Files CCFE R 11 11 Issue 6 FISPACT 1 User Manual keyword FUEL has too many arguments Line number 16 token Ti60 Skipping to next keyword ERROR in INPUT file keyword DOSE has too many arguments Line number 27 token 1 0 Skipping to next keyword ERROR in INPUT file Abbreviated keyword MIN matches 2 keywords Detected on line number 36 Skipping to next k
47. OVER now accepts an alternative name of ADCROSS for the subordinate keyword ERROR and a new subordinate keyword ADLAM for the input of a new error factor for the half life PATHRESET is a new keyword that allows pathways calculations to be repeated at times after the initial pathways calculation at the end of the irradiation phase PROBTABLE is new keyword that causes the probability table data to be read and used for the nuclides specified by the SSFCHOOSE keyword SAVELINES is a new keyword that causes spectral lines to be read from the decay file and stored It is needed if the new PRINTLIB 5 option to print decay lines is used CCFE Page 24 of 200 3 8 Keyword Changes CCFE R 11 11 Issue 6 FISPACT II User Manual SEQNUMBER has no effect apart from generating a warning message SEQUENTIAL has no effect apart from generating a warning message SORTDOMINANT is a new keyword to control the uncertainty calculations and their display in the output file SSFCHOOSE is a new keyword used to specify the nuclides for which the self shielding factors are computed SSFDILUTION is a new keyword to provide expert control of dilution values used in applying the self shielding corrections SSFFUEL is a new keyword used to specify directly the mixture of nuclides to be used in the self shielding calculations SSFGEOMETRY is a new keyword that allows thin and thick target geometry information to be input for use in conjunction with th
48. Qross SectlODS omo o BA T E a oe s 119 7 3 4 Bremsstrahlung candidates 120 7 3 5 Projectile spectrum 4324 304 o 5446 ds ee awe 120 7 3 6 Decay spectral lines 2e 120 7 4 Probability Table Collapse Run 22252239 sae dee ees 121 7 5 Universal Curve Self Shielding Collapse Run 123 References CCFE Page 10 of 200 CONTENTS CCFE R 11 11 Issue 6 FISPACT I User Manual APPENDICES 133 A The Model 133 A The Rate Equations e sprccirnrsss dr AA oA 133 A 2 Data Collapse ses 135 A3 Decay Modes zs dosa ak Boake alot RUE oW EA A 3BA Heating i4 29 9x e MPO a ERU uo 3E b eR a Hoe A 3 2 Gamma spectrum 2 4 ee ee ee 139 A 3 8 Neutron yield 22222 139 A 4 Neutron Activation 2 2 2 2l ls sss sss 139 A 4 1 Other reactions gas heat and damage 142 AAD Ignored reactions 4 4 ao A RR RE e eng A 4 3 Selfshielding using probability tables A 4 4 Selfshielding using the universal curve model Ab FUSSION else Sl o koh San ba ee ROR RO Foy Bow ow 150 A 5 1 BAF data ses ss lors 150 A 5 2 ENDF data o sapara rs 152 A 6 Gamma Activation lll les 152 A 7 Proton Activation 2 llle sss 152 A 8 Deuteron Activation 2 2l llle 152 A 9 Alpha Activation 2 2 e 153 A 10 Gamma Radiation len 153 A 10 1 Contact y dose Tate len 153 A 10 2 Point source y do
49. Rate Equations CCFE R 11 11 Issue 6 FisPACT II User Manual for 1 lt i j N These initial value problems can be solved to give y t using numerical methods A 14 1 Properties of the equations The rate equations are linear because at the level of approximation used in FISPACT II the reaction rates and decay constants are independent of the current inventory being determined purely by intrinsic physical properties of the nuclides and also the imposed external projectile flux is assumed not to be modified by the presence of the target material Each reaction or decay typically produces a single principal daughter nuclide and a few secondary products although fission reactions are an exception Even with fissions included less than 396 of the matrix elements of the system matrix A in Equation are non zero Without fissions this proportion drops to about 0 8 This sparsity is very significant for numerical approaches to the solution In principle Equation can be solved in closed form for each time interval during which A is constant Introduce a matrix S to define a similarity transformation which diagonalises A and rewrite Equation as d SN dt SAS SN 83 where SAS diag ux p2 Pin 84 and u are the eigenvalues of A Note that the probability of A having repeated eigenvalues is vanishingly small since the elements of A are derived from physical data Hence A is diagonalisable the matrix S exists
50. Tst 709pt directory illustrates the combination of the CALENDF probabil ity table data with the TENDL 2013 709 group cross section data for neutron irradi ation to add self shielding corrections to the collapsed cross sections The Tst 709uc directory illustrates the use of the alternative universal sigmoid curve approximation for self shielding corrections The Tst 709 directory contains some sample inventory runs using different 709 group library data Further tests using the different ENDF libraries are in the Tst 7091ib directory The Tst 709cern directory contains exam ples where the LOOKAHEAD and PATHRESET keywords are used to capture pathways for late time dominant nuclides The Tst 709mc directory contains examples of the use of the Monte Carlo sensitivity calculation Tst binxs illustrates the use of compressed binary files derived from the ENDF data libraries to speed up calcula tions Tst 709 ns gives examples of the validation calculations for decay heat 24 and Tst pulse contains examples of validation tests using pulse irradiation of actinides 7 Interpretation of Output All FisSPACT II runs have two main output files output containing the physical results of the calculation and runlog containing error reporting and logging information 7 1 The Inventory Run output File The layout of output has been designed to follow closely that of FISPACT 2007 Unless stated otherwise the following excerpts are taken from the inventory ru
51. Version 2 20 release of FISPACT II There have been three major versions of the code Version 0 was a direct functional replacement for FISPACT 2007 It differed in that it used improved algorithms and was written in object Fortran 95 The data encapsulation together with full dynamic memory allocation provided a robust and flexible platform for the new capabilities introduced in the later versions to be built on It used the same user input files and was designed to use the European Activation File EAF 2007 8 9 and EAF 2010 13 data sets for cross section decay fission and radiological quantities and was extensively validated and cross checked against FISPACT 2007 Discrepancies in the results from the two codes have been shown to arise from the increased number of reactions and improved numerical methods employed in the new code Version 1 began the process of extending the activation transmutation prediction capability whilst maintaining the validation heritage of FISPACT 2007 New pathways and new monte carlo sensitivity capabilities were introduced to extend both pathways and sensitivity calculations to multi pulse irradiation cases The reading and processing of CALENDF probability table data was introduced in the calculation of cross section collapse to include self shielding effects in the inventory calculations Version 2 The major change introduced in Release 2 0 of FISPACT II was the addition of the reading and processing o
52. a statement of the number of individual pathways combined to create the generic pathway Target nuclide Sc 46 97 564 of inventory given by 3 paths path 1 87 893 Ti 46 R Sc 46 S This generic pathway is the sum of 2 pathways path 2 9 124 Ti 47 R Sc 46 S This generic pathway is the sum of 2 pathways 7 1 13 Run summary At the end of a run tables are printed containing the total values for each time in terval The intervals are listed as Irradiation Phase or Cooling Phase in the most appropriate unit sec min days and cumulatively in years Six columns present Ac tivity Bq Dose rate Sv h Heat output kW Ingestion dose Sv Inhalation dose Sv and Tritium activity Bq For all except the latter the estimated uncertainty is also given If the SPLIT keyword is used with parameter 1 then a second summary table con taining Beta Heat KW Gamma Heat kW Mean Beta Energy MeV and Mean CCFE Page 113 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual Gamma Energy MeV is printed For all quantities the estimated uncertainty is also given Time Cumulative Activity Dose rate Heat output Ingestion dose step Years Bq Sv h kW Sv Irradn 25 25E414 E 63E 04 9 60E 02 9 1 39E 05 Cooling Cooling Cooling 90E 06 14E 14 E 61E 04 3 57E 02 98 38E 05 16E 04 09E 14 E 49E 04 9 3 45E 02
53. all settings that are to apply to the reading of new cross sections must be declared before the use of GETXS 5 2 18 GRAPH numg grshow guncrt nopt i i 1 numg This keyword specifies what information is stored in the file graph for subsequent post processing The number of graphs required numg is input and for each graph an option number nopt i is read Allowable values for the options are 1 Total Activity 2 Total y dose rate 3 Total heat output 4 Ingestion dose 5 Inhalation dose The parameter grshow allows slightly different versions of the data file to be con structed If grshow 0 then an output suitable for PC post processing is obtained if grshow 1 then the output might be more suitable for other platforms If grshow 2 then a gra file is written in a form suitable for gnuplot and a plt file contain ing gnuplot commands to plot the graphs is also written For example issuing the command gnuplot test81 plt CCFE Page 65 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual 5 CONTROL FILE KEYWORDS will create the file test81 gra ps from test81 gra An example of an activity output graph produced by this command on a Linux workstation is shown in Fig IRRADIATION OF TI EEF 175 FW 1 0 MW M2 1e 15 z 5l y E T EJ ae Si DEAR TET i Activity Bq kg 4 Uncertainty 4 T i value t half for nuclide 1 ll Jr in 1e 14 A e Asc 1 2 9 457 a gt 10 13 q gt 2 Msc 1
54. and cooling and would model the loss of volatile elements during re fabrication In the second case the irradiation might be split into several intervals and PARTITION used in each interval to model the loss of tritium An example of the use of this keyword is PARTITION 2 Ar 0 01 K 0 20 In this case all elements except argon and potassium remain unmodified all argon isotopes are reduced by a factor of 100 and all potassium isotopes are reduced to a fifth of their values before the keyword was used CCFE Page 91 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual 5 3 17 PATHRESET showpathways For inventory calculations with long cooling times the dominant nuclides at late times may not be significant at the end of the irradiation phase and this leads to poor estimates for the uncertainties One remedy for this is to use the LOOKAHEAD keyword In some instances particularly where there are actinides in the source mate rial the look ahead approach may lead to excessively large numbers of target nuclides in the pathways calculations at the end of the irradiation phase and this may cause slow calculations and in some cases exhaustion of available heap storage The PATHRESET keyword provides an alternative means of including late time dominant nuclides Its inclusion after the ZERO keyword leads to the pathways calculation being repeated at the cooling step immediately before its occurrence It can
55. and generation rates of gas atoms New stiff ode solver the solution of the inventory equations is now based on the LSODES package There is no equilibrium approximation and the time depen dence of actinide inventories is treated in full New pathways analysis graph theory based tree searching methods are now used to identify significant pathways removing the restrictions of the previous meth ods All loops and paths are automatically included if their contribution is above the user specified thresholds and searches can be made to arbitrary depths Pathways analysis works for single and multi pulse irradiation phases and chang ing cross sections Information on all reactions between a given parent and daughter is available and is displayed by pathways output Covariance data Reaction cross section data for different reactions can be read from the TENDL 2013 files and used to produce collapsed covariances and correlations New sensitivity analysis the local derivative sensitivity analysis calculation imple mented in FISPACT 2007 that was only applicable for single irradiation pulses has been replaced by a Monte Carlo sensitivity calculation that works for single and multi pulse irradiation phases and changing cross sections New reduced nuclide set runs with subsets of the nuclides in the EAF or ENDF libraries can be undertaken Encapsulated data FISPACT II is written using an object based modular code de sign including built in er
56. and the additional complexity of the Jordan normal form is not required Then the solution is SN t diag exp uit exp uot exp ust SN 0 85 for t 0 T a time interval during which A is constant The extension to piecewise constant A can be found in the obvious way by evaluating S and the eigenvalues for each interval and concatenating intervals with a rotation of the solution vector after the endpoint of each interval Unfortunately this analysis does not represent a practical method of solution because the process of calculating the eigenvalues is numerically intensive destroys the sparsity structure of the matrix and is subject to the possibility of extreme ill conditioning However the eigenvalues are useful in identifying the properties of the matrix A as now described CCFE Page 163 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual Table 18 The largest decay rates in the EAF library Range of A s71 Nuclides 21 lt logio A lt 22 Li UN F 20 lt logy A lt 21 Be 19 lt logy A lt 20 F 18 lt logio A lt 19 none 17 lt logio lt 18 B 16 lt logio lt 17 none 15 lt logio A lt 16 Be 10 lt logy lt 15 none 9xloggA 10 13Be 7 logjgA lt 9 none 6 logigA lt 7 212P 946 Aq Ra A study was performed by extracting the matrix A from FISPACT II and using it in a testbed program which did
57. are not applied The new treatment of fission yield is described below in Subsection A 5 1 EAF data Projectile induced fission yield data are available in three projectile energy ranges thermal under 200 keV fast between 200 keV and 5 MeV high over 5 MeV The boundary energies are E 5 MeV and Er 200 keV It is assumed that there is a maximum of one fission yield fraction in each of these energy ranges for a given projectile parent and daughter fragment The algorithm for infilling unknown values is e If yields for the thermal fast and high energy projectiles Y Y and Y are known then these are used CCFE Page 150 of 200 A 5 Fission CCFE R 11 11 Issue 6 FISPACT I User Manual e If only one value Y is known then set Y Y Y Y e If only Y and Y are known set Yp Y Y 2 and if values for Y or Yp are unknown then set them to Y A single fission yield factor for use in the inventory equations is obtained by collapsing the available data in a manner similar to that used for cross sections Equation 10 Fluxes in the thermal fast and high energy groups are found by summing fluxes in the narrower groups used for the cross sections A simple nearest grid point algorithm for this is as follows Let be the flux in the cross section energy group that lies between energies E and E 1 data are in decreasing energy order and 1 lt i lt imar then the group energy is E E Ej 4
58. be changed to assess their effects CCFE Page 20 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual A reduced list of nuclides can be generated from the pathways information and full inventory calculations can be undertaken on the reduced set of nuclides and reactions to check whether all important reactions and decays are included Uncertainty estimates can be made by combining the pathways information with un certainty data for cross sections and decay rates Sensitivity calculations provide a complementary method of identifying important re actions providing uncertainty estimates and for quantifying how the uncertainty in the final amount of a nuclide depends on the uncertainty of specific reactions 3 Differences from FISPACT 2007 Many keywords in the FISPACT 2007 control input file have been retained in FISPACT II to provide a substantial degree of backwards compatibility In many cases the new code will run with existing control input files Some new keywords have been added to deal with the new capabilities of FISPACT II and some of the old keywords have become obsolete Where a keyword no longer works as before the new code will issue a warning or fatal error message 3 1 New Features The new and extended features of FISPACT II are Additional projectiles five projectiles may now be used n p d a and y Additional reactions 90 reaction types are now recognised Additional nuclides and elements Elements
59. by the An example of the use of this keyword is END END of Fe run 5 2 10 ERROR nerror parent i daughter i ermat i i 1 nerror This keyword inputs the number nerror of reactions and the identifiers of the parent and daughter of each reaction and optionally the fractional error of the reaction cross section In versions of FISPACT prior to 3 0 the user had to input a value of the fractional error but this is now available from the EAF or ENDF uncertainty files If data from the uncertainty file are to be used then ermat must be set to 1 If the keyword is absent then all ermat values default to 1 Note that if no uncertainty data exist in the library then the fractional error must be input using 1 will cause an error message to be printed This keyword should only be used following the keyword SENSITIVITY to give the error in the number of atoms of a nuclide due to the specified reactions for routine calculations the uncertainty calculations are automatically performed by a simplified method Parent daughter pairs listed must also appear in the SENSITIVITY list An example of the use of this keyword is CCFE Page 61 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT 1 User Manual ERROR 2 Li7 Li8 1 0 Be9 He6 1 0 Line 2 specifies that the reaction Li n y Li is to be considered Line 3 specifies that the reaction Be n a He is to be considered The uncertainty for both reacti
60. calculation The target nuclides included in pathways calculation are by default selected by merging the dominant nuclide lists at the end of the irradiation phase The number of nuclides included in the merged list is controlled by the topxx argument of the SORTDOMINANT keyword see page B0 The number of nuclides selected by the default toprr and the pruning of the path ways search tree caused by the default path_floor loop floor and maz depth values usually lead to a quick and accurate pathways and uncertainty calculation However even the pruned tree search is subject to combinatorial growth and so in some cases computational times may become excessive or the available heap storage may become exhausted Balanced against this is the need to keep sufficient pathways to ensure that important reaction and decay chains are identified and included If excessive time for pathways calculations is encountered then try using larger path floor and loop floor CCFE Page 84 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual values and smaller maz_depth and topxx values to get faster calculations and then use different values to assess convergence Inventory calculations which have long cooling times pose a particular problem in that dominant nuclide in late cooling times may be insignificant at the end of the irradi ation phase A symptom of this problem are uncertainties that drop to zero at late times bec
61. cross section data forms Together FISPACT 11 and TENDL s nuclear data forms make up the European Activation System II EASY II that is a complete package tailored for all application needs nuclear fission and fusion nuclear fuel cycle accelerator physics isotope production material characterisation storage and life cycle earth exploration astrophysics homeland security and more The fol lowing data libraries are required CCFE Page 195 of 200 CCFE R 11 11 Issue 6 C TENDL LIBRARY DATA FISPACT II User Manual 1 Cross section data for neutron proton deuteron alpha and gamma induced reactions 2 Fission yields data for neutron proton deuteron alpha and gamma induced reactions 3 Variance covariance data for neutron induced reactions 4 Probability tables data for neutron induced reactions in the resonance energy ranges 5 Decay data 6 Radiological data e Biological hazard data e Legal transport data e Clearance data To streamline simplify and control any feature of all the nuclear data assimilation processes the code development philosophy has been to follow in all aspects as much as possible the format described in the ENDF 6 format manual 16 Three processing codes are used in sequence and in parallel to produce process check and compare the nuclear data forms NJOY12 021 PREPRO 2013 and CALENDF 2010 All the processing steps cannot be handled by only one or even two of those unique proce
62. cross section for nuclide p in energy group g and for MT value y belonging to the macro partial group x is given by LIB 0 y dp 0 2 dp A 37 and for the total scaling factor g HB got are y dy c z oo y en 38 M yes TF y J Xof t oo CCFE Page 146 of 200 A 4 Neutron Activation CCFE R 11 11 Issue 6 FisPACT II User Manual The initial values of the dilutions are given by Equations and and the iterative refinements where CALENDF probability table data are available are given by Q SOl Y fo q g d q g 39 q 1 so g tot dep y 7 ap g a 40 p The set of nuclides for which the self shielding correction is calculated is specified by the SSFCHOOSE keyword The set of nuclides included in the mixture for computing the dilution cross section is specified by either the SSFMASS or SSFFUEL keyword Nuclides included in the SSFCHOOSE keyword list that are not included in the in mixture will cause a fatal error message to be issued by the program The values of dilution given by Equation Or may be overridden using the SSFDILUTION keyword Section 5 1 19 Tables of the energies cross sections dilutions and self shielding factors are printed for each of the nuclides to which the self shielding correction is applied i The final diagnostic table gives the collapsed cross sections with o p y and with out a7 P p y the self shielding correction Also printed is th
63. de a Ae URGE Ge ERE S 5 231 NOCOMP 0 200 0 ee 72 5232 NOSORT 4 3532 aoa ke el doe De poa E we Be a 5 2 30 NOS TABLB ss sii Da ah Ree a Swe de a 5 2 94 NOTI veo asda e ee ON We Ge ee ja res we eeu 5 2 39 NOU gorse e unm ae EXER EU RIP ERE ER ob ee ee dels d QUA 5 2 30 NOT oda uerum heo da e X CR eS dd paar NOTA 2 6 5 sx mos xb uu the a don a dinde xu ed la Rom 5 2 OO ONER AR tns obo in in rer ge eee RES dE LEE fem un CCFE Page 8 of 200 CONTENTS CCFE R 11 11 Issue 6 FISPACT II User Manual 5 3 5 2 99 PATH 2 we ante cg Rath ecu ws eae eo de deut ew es 5 2 40 PATHRESET sess 76 padl PRINTLID 2 uu ua ee ae ohm RUE a C a ae o 77 5 242 PROBTABLE e 5 2488 ROUTES coo o lt lt lt lt o RE S s 78 5 2 44 SENSITIVITY we eee a p eee Be pande qe 3 79 5 245 SORTDOMINANT sers 5 2 46 SPECTRUM 2 2 2 22 2 222522255 DAG SPLIT uuu ec aoe Be eS ke Xem Ew he eR a 4604s 5 248 SSFCHOOSE 2 0 a e 5 249 SSFDILUTION les 5 2 00 SS ERU EI 0 2 0040 RR UR a oe BB RSEN S 5 2 51 SSFGEOMETRY 2 5 2 52 DOFMASS i i uo pai GON ER BLE RS ew R X 3 PR EURO EK 52 53 TABI s 2 msn M Rue kem XD E e He SDN E a d 6o D204 TABQ x 244 drm goce a oh e Rede ee P d D5 TABS s gees dee Sars citi Gs eds gh pened id id iaa 5 2 50 ABA uu ick ao Boe we eod Reb a ee ek ee doe 5 2 94 TIME uta eer cec XE a PR Be oe Se 82 5 2 58 TOLER
64. end of the run This summary table contains separate information on the heat production by beta and gamma radiation at each time interval and is output after the existing summary table By default this new summary table is not printed but it can be displayed if split is set to 1 Note that if the new summary table is required then the keyword HAZARDS must be used to ensure that uncertainties are correctly printed CCFE Page 80 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual 5 2 48 SSFCHOOSE ncho 0 nprint 0 sym j j 1 ncho The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 18 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 2 49 SSFDILUTION nnuc nucname j num j grp i j dilution 1 j i 1 num j j 1 nnuc The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 19 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 2 50 SSFFUEL ni is j atoms j j 1 n1 The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 20 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 2 51 SSFGEOMETRY type length1 lt length2 gt The primary use of this keyword is in the library data preparation section of the input f
65. floor 0 005 All pathways contributing more than the path floor fraction of the inventory of the final target nuclide are retained loop_floor 0 01 All loops that increase the inventory contribution of the path they are on by a fraction greater than the oop floor are retained CCFE Page 83 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT 1 User Manual max_depth 10 is the maximum number of links in a path from a source nuclide to a target nuclide iuncer following all the other parameters allows values 0 1 2 or 3 to be input again so that after resetting the default values an actual calculation with the new values can be done Note any use of the keyword UNCERTAINTY to change the default settings for the pathways calculation must precede the first occurrence of the keywords ATOMS or SPECTRUM Note that if no uncertainty data exist in the cross section library then the valid values of iuncer are only 1 0 or 3 Omitting the keyword will ensure that only inventory calculations are carried out and should be done if a fast scoping run is required Examples of the use of this keyword are This will ensure that in addition to the inventory calculations the pathways to form the dominant nuclides and the uncertainty estimates are output This is the standard use of the keyword for a full investigation of activation UNCERT 1 0 001 0 005 10 2 This resets the default values and then carries out a full
66. gives the adjacency ma trix and its components give weighting factors for constructing the digraphs CCFE Page 134 of 200 A 2 Data Collapse CCFE R 11 11 Issue 6 FisPACT II User Manual used for pathways analysis The ind nuc file in the EAF 2010 library contains 2233 distinct nuclides so A has a size of 2234 x 2234 including the sink nuclide However there are only approximately 120000 non zero elements in A If actinides are not rel evant to a calculation the fission reactions can be omitted and about 42 000 elements of A remain non zero This number drops to less than 5000 during cooling periods when only decays are required These properties of the system matrix are relevant to the method of solution described in Section A 14 below A 2 Data Collapse The reaction data input to FISPACT II are the projectile flux spectrum cross sections induced fission yields and covariances tabulated in energy groups where in general the cross section data are tabulated at much smaller energy intervals than the fission yield or covariance data These data are collapsed using flux spectrum weighting into energy independent values for use in the inventory calculations Consider the collapsed cross section X and its uncertainty A that are used in FISPACT II The input data for X are cross sections X and the projectile flux p in energy groups i 1 N is the flux em s in energy range E to E 1 and we use it t
67. input and output files for a run of the code The input file has the name lt fileroot gt i and the output files have names fileroot ext where the list extensions is given in Table Thus for example if the input file is called example i then the run CCFE Page 26 of 200 4 1 Introduction CCFE R 11 11 Issue 6 FISPACT II User Manual Table 1 Filename extensions for user input and output files extension unit name description 1 input The run control file out output The main output file log runlog The logging and error output file gra graph The graphical data output file plt gnuplot The gnuplot plot command file sens sens The raw sensitivity data output file tabl tabl Number of atoms and grams of each nuclide table tab2 tab2 Activity and dose rate of each nuclide table tab3 tab3 Ingestion and inhalation dose of each nuclide table tab4 tab4 Gamma spectrum table Table 2 Mapping of internal unit names to external EAF library files unit unit EAF library file name number absorp 39 Element gamma absorption data ind nuc 18 Index of materials included in run ind nuco 49 Output file for reduced index of materials from pathways crossec 19 Cross section library crossunc T Cross section fractional uncertainties library decay 16 Decay data fissyld 9 Fission yield data asscfy 8 Links between fissionable nuclides and fission yields a2data 11 A2 transport data clear 40 Clea
68. inventory run are specified 5 3 21 SPECTRUM This keyword is an alternative to ATOMS It suppresses the inventory output so that only the y spectrum and total values are printed for the time interval It is useful if summary information is required for many time intervals but the details of the individual nuclide contributions are not needed This keyword may also be used in the initial conditions section of the input file see Section 5 2 46 5 3 22 SSFCHOOSE ncho 0 nprint 0 sym j j 1 ncho The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 18 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 8 28 SSFDILUTION nnuc nucname j num j grp i j dilution 1 j i 1 num j j 1 nnuc The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 19 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 3 24 SSFFUEL ni is j atoms j j 1 n1 The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 20 In this section it will apply to the actions of the next occurrence of the GETXS keyword CCFE Page 94 of 200 5 3 Inventory Calculation Phase CCFE R 11 11 Issue 6 FisPACT II User Manual 5 8 25 SSFGEOMETRY type length1 lt length2 gt The primary use of this keyword is in the
69. is taken completely from the JEFF 3 1 1 39 fission yield library and FISPACT II reads the file in ENDF 6 format with no pre processing Only 19 of the 102 nuclides in eaf n xs which have fission cross sections have any fission yield data in JEFF 3 1 1 at relevant energies For the remainder a neighbouring fission yield is used For the EAF 2010 library the file eaf_n_asscfy_20100 connected to the stream asscfy contains these associations B 5 2 Deuteron eaf d fis and eaf d asscfy eaf d fisis taken completely from the UKFY 4 0 fission yield library 47 and FISPACT II reads the file in ENDF 6 format with no pre processing Only 19 of the 90 nuclides CCFE Page 190 of 200 B 5 Fission Yield Data CCFE R 11 11 Issue 6 FisPACT II User Manual Neutron flux per lethargy 3 11 18 1 I 1 I 1 lib vu m iv 107 F104 9109 5 9 CiP Si 5 o 95 95 5 99 9 9 Neutron energy eV a EEF study 10 T T T T 5 10 5 10 Neutron flux per lethargy 3 107 5 8 19 1 I il Imm 1 1 fi I L 107 9 9 5 09 5 9 95 9 5 935 5 9 5 95 108 5107 Neutron energy eV res c Ph nix fast reactor 10 T sE 3 10 sE 4 1012 gt sE 3 a E m 3 1019 pue oe 3 sE 7 3 3 a 10 7 L H sE 3 4 16 8 10 E sE 3 3 2 15 10 sE 3 101 sE 3 109 1 i 905 2 a 4 507898 2 a 4 567897 2 3 Neutron energy eV e NIF ignited Figure 10 Sample Neutron flux per lethargy RE 3
70. mom dod hx Rx E boues 5 2 2 JUTEWQO ua seule woo Ehe EO Rp a de ode dou 5 2 3 BREMSSTRAHLUNG ss 5 2 4 CLEAR sesa aoe tta s modo RR X ok m ad RO RO deo 525 OUEPDADB uen bois Ki es ee PIENO HS eS d 5 26 DENSITY i8 dae a gu eR CR es ee A e Re b DOSE sarraa Ronde A a GREC Reve S 5 2 8 EAFVERSION esee es D 2 9 END scott o Bax web xo Pe UE od E d 5 2 10 ERROR restaranga eem Rkch Ee ce Ae de PSR ek a 5 2 11 FISCHOOSE 42s 5 2 12 EISYIBELD thou e a ee aie ee ee 3 PIENE RE Nm a 62 5 2 13 FELUS 3 i a ER dig snc eene ley bcp oad Ee 5 2 14 FUEL ves roca e a REG a Xe Rm p EROR EDAD GR CR XOU E 64 5 2 15 EULDXS x wae eee xu ox ar Ex wu E Romo d o4 64 5 2 16 GENERIC amp 3m degoR Ce ke e BR Y dedu RR eS Did GETAT sergen Enp Re mo aho edo de e ad 65 5 2 18 GRAPH 2 uou do Ep REGIE GRE Er x 5 2 19 GROUP ius Eo oe ki eR A al uM Aen d 67 5 2 20 GRPCONVERT Secos ra rp e Poe Be a dd 67 5 2 2T HALE o cotairean ei k ene BEE a dee mee Ed e 67 5 2 22 HAZARDS gt 42440084 eae be EO oe OS RR e Eo 52 23 INDEXPA EH 45 cuicos oho Paese urbes dem ds d ern 3 5 2 24 TRON i xm woo secos RR RR REOR OR OR Rom ED Eod 5 2 25 bOGEBEBNMVIELE atu Re ta Wo 5 2 20 EOOKAHEAD i33 6G ee A Rs pU RUE ER a 69 52 24 MASS D ui 408 dox oc ae due herum oed deu e uude ek dd 70 5 2 28 MCSAMPLE 4s 5 2 29 MCSEED i 4 a a EUR dr EGER EU o Xu RR Roy e 5 2 30 MIND ius a anuo ne Ree
71. ni Ei Em Gamma doses for approximate spectra are found using the intensity from Equation 62 to find the emission rate Equation 60 and then using this rate in Equation 58 or as appropriate A 10 4 Bremsstrahlung corrections The contribution of high energy P particle bremsstrahlung to the total y dose rate can be significant in cases where the y emission is small FISPACT II uses a similar approach to Jarvis 40 who considers y emission from a mono energetic electron The energy distribution of y rays emitted by a mono energetic electron in a matrix of charge Z is given by 63 PN aZ 252 dE 0 lt E Eo E gt Eo where dN number of 4 rays with energy E keV Eo energy of electron keV a 2 76 x 1079 keV 1 Integrating Equation over the energy bins give the number of y rays associated with that bin There are three cases aZ Eglog E j1 E Eij1 Ej Eo 2 Eia N i aZ Eolog Eo E Eo Fi Ej gt Eo gt E 64 0 Eo E where E and E are the lower and upper energy bounds of group i The intensity for group 1 is given by I N i Ei Ej41 2 65 The bremsstrahlung corrections to gamma doses are found using the intensity from this equation to find the emission rate Equation 60 and then using this rate in Equation or as appropriate CCFE Page 155 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual The above discussion is valid only for mono e
72. number of tests reveals that LSODES still performs well on a small non stiff reaction network typical of a pathway calculation The automatic error control see Section below ensures that the results of Gear s method remain accurate although LSODES may well be more efficient using fewer internal timesteps if the Adams method had been used instead LSODES performs the solution of stiff systems of equations without recourse to the equilibrium approximation that was used in earlier versions of FISPACT All nuclides are followed dynamically and those with a rapid transient response automatically ad just to have near equilibrium inventories A 14 3 The interface to the solver Some of the complexity of the interface to LSODES arises because of the limitations in Fortran 77 concerning fixed size arrays which must be defined at compile time These limitations can be overcome with the dynamic memory allocation features now available in Fortran 95 The present development provides a Fortran 95 wrapper for the old Fortran 77 code with a simplified interface and automatic circumvention of LSODES error reports caused only by inadequately sized workspace arrays Most of the details concerning the operation of LSODES can be encapsulated in a way that is consistent with the object oriented approach adopted for the present develop ment However it is necessary to provide a description of the sparsity structure of the system matrix A in a specific manner tail
73. of J dependent variables 155 gel sg 0245 In the present context the independent variables are cross sections and their uncertainties or decay constants and their uncertainties The dependent variables are the numbers of atoms of nuclides j or some related radiological quantity The implementation of this scheme uses the SENSITIVITY keyword to initialise the collecting of data within the main inventory calculation The keyword ZERO causes the series of S runs with different independent variables to be undertaken to compute process and output the set IE The default distribution is taken to be log normal but other options are possible See keyword MCSAMPLE on page 71 CCFE Page 156 of 200 A 11 Monte Carlo Sensitivity Estimation CCFE R 11 11 Issue 6 FisPACT II User Manual Any sequence of irradiation pulses changes in cross section etc that are possible with FISPACT II can be used in the sensitivity calculations The code performs the base calculation with full output then repeats S times the sequence of steps with different sets X The results of the base calculation are not included in the sensitivity calculation Sensitivity calculations provide both uncertainty and sensitivity output Summary uncertainty output of means X and Y and standard deviations A X and AY are sent to the output file a L Xi GX 67 s 1 AX X sp x 68 7 qs OD Rn 68 12 RCM 69 s 1 po a AY STA 70 s 1 Diffe
74. of magnitude and can reduce the computation time of larger Monte Carlo sensitivity calculations by several orders of magnitude The reduced index can be created by hand editing the full nuclide index to retain only the light gas nuclides and those in the region of Z A space around the target materials being studied Alternatively a reduced index can be automatically generated from a full calculation with pathways analysis by including the INDEXPATH keyword in the input file 5 Control File Keywords A run of FISPACT II is controlled by a sequence of commands given in a user supplied input file as illustrated in the previous section Each command is introduced by a keyword which may be followed by integer real or character string parameters Some commands require further data to be supplied in records of the file following the key word Some commands are followed by subordinate keywords which cannot be used independently of their parent keyword The keywords belong to one of two classes distinguished by their effect on the calcula tion Some keywords provide settings such as logical flags and numerical values while others cause FISPACT II to perform actions Depending on the context the effect of an action keyword may be immediate or its action may be added to a queue and its execution deferred The input file is divided into three sections 1 library data preparation reading and processing the physical and regulatory data suppli
75. perform the eigenvalue calculation by employing the library routine GEEVX from the LAPACK library A cooling step provides information on the decays in isolation whereas an irradiation step with unit flux provides the sum of the decay rate matrix and the cross section matrix The eigenvalue analysis of the decay rate matrix highlights the presence in the EAF library of a few nuclides with very rapid decays as listed in Table These very large decay rates ensure that all practical FISPACT II calculations with the full inventory for many applications are always stiff This remark applies as much to laser fusion applications with nanosecond irradiation pulses as it does to magnetic confinement applications with irradiation times of years However when subsets of the nuclides are used in pathways calculations the reduced set of equations may not be stiff A 14 2 The choice of solver It can be seen from the previous section that the key characteristics of the system of inventory equations are that they are linear stiff and sparse A web search reveals several suitable solvers but it appears that only one can be ob tained with built in efficient handling of sparse systems This is the package LSODE 17 written at Lawrence Livermore Laboratory The variant of LSODE usually used in FISPACT II is the double precision version with efficient handling of sparse Jacobian matrices called DLSODES although on some platforms the single precision v
76. second keyword specifies collapse of the 709 energy group cross section data The SSFGEOMETRY keyword activates the universal sigmoid curve self shielding approximation and indicates that a foil target 8mm thick is to be irradiated The SSFFUEL keyword specifies the mixture of nuclides whose resonances are to be used to calculate the self shielding factors The output from a run using this dataset has the labelling and heading information for this self shielding approximation followed by the list of the parent nuclides that provide resonances for the calculation of the self shielding factors SIGMOID CURVE SELF SHIELDING CHANGES TO CROSS SECTIONS Target geometry set by the SSFGEOMETRY keyword foil with thickness 8 00000E 01 cm CCFE Page 124 of 200 7 5 Universal Curve Self Shielding Collapse Run CCFE R 11 11 Issue 6 FIsPACT 1 User Manual Target mass and inventory numbers of atoms refer to unit foil area The self shielding factors are calculated from the resonances of the materials specified with the SSFFUEL or SSFMASS keywords Material Mixture List Nuclide W 182 W 183 W 184 W 186 A full list of collapsed cross sections can be obtained using PRINTLIB The collapse run simply summarises the reactions whose cross sections are changed significantly by self shielding reduced to less than 9096 of their infinitely dilute values The table for Atoms percent 26 534 14 319 30 680 28 467 this example starts as f
77. step run add pathstep run output inventory run pathways uncertainty TIME 1 HOURS ATOMS run cooling step run add pathstep run output inventory run pathways uncertainty TIME 1 DAYS ATOMS run cooling step run add pathstep run output inventory run pathways uncertainty TIME 7 DAYS ATOMS run cooling step run add pathstep run output inventory run pathways uncertainty TIME 1 YEARS ATOMS run cooling step run add pathstep run output inventory run pathways uncertainty END END run output summary run closedown deallocate and closedown The QA information on files used that was written to the output file is also written for cross reference to the runlog file followed by a cpu timing summary of the major program components CCFE Page 116 of 200 7 2 The Inventory Run runlog File CCFE R 11 11 Issue 6 FISPACT II User Manual Log fispact run time 0 37094 secs Log rateeq init flux 0 0000 secs Log rateeq irrad step 0 10199 secs Log rateeq cool step 0 77989E 01 secs Log output inventory step 0 13996E 01 secs Log pathways step 0 80988E 01 secs Log sensitivity step 0 0000 secs Error Summary total number of errors warnings number of serious errors warnings 16 03 10 22 May 2013 END OF LOG FILE In runs where errors are flagged output of the following form taken from test10 is displayed 00001 Warning output m output inventory 1 gt 20 of dose from nuclides with no spectral data Th
78. that allows fine tuning of the dilution factors is provided using the SSFDILUTION keyword See Appendix A 4 3 for a more detailed explanation of the alternative calculations per formed and Section T 4 for an illustration of including self shielding in the computation of the effective collapsed cross section data At present the probability table data are available only for the 616 and 709 energy group structures for neutrons Attempts to use this keyword with other projectiles or cross section datasets with other than the 616 or 709 energy group structure will cause FISPACT II to terminate with a fatal error The TENDL 2011 TENDL 2012 and TENDL 2013 709 group data for neutron pro jectiles do contain elastic scattering cross section data and so both choices of multzs should give very similar results if these are used in conjunction with probability table data 5 1 15 PROJECTILE nproj 1 This keyword defines the incoming particle for the activation calculations This key word must be used if a library other than a neutron activation one is used At present cross section uncertainty data are known only for neutron induced reac tions so if nproj is not 1 then the NOERROR keyword must also be used CCFE Page 53 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual For a gamma library nproj should be set to 5 for a deuteron library nproj should be set to 2 for a proton library nproj shoul
79. the other basic quantities required by an inventory code are data on the decay properties such as half life of all the nuclides considered These data are available in the various evaluated decay data libraries FISPACT II is able to read the data directly in ENDF 6 format it requires no pre processing to be done although file debugging has always been found necessary eaf dec 20100 is based primarily on the JEFF 3 1 1 and JEFF 2 2 radioactive decay data libraries with additional data from the most recent UK evaluations However not all of the 2233 nuclides that are needed are included in such sources For these nuclides data are taken from sources such as Brown and Firestone 46 and ENDF 6 format files have been constructed Reference documents the eaf dec 20100 library Care has been taken to ensure that eaf xs and eaf dec are compatible All nuclides including isomeric states that can be formed from the various reactions in eaf_xs are included so long as their half lives are greater than 1 second Some nuclides with shorter half lives are included where it is felt that they are of particular importance Short lived 1 s isomers which return to the ground state by an isomeric transition usually have no impact on activation calculations and most of these have been ignored B 5 Fission Yield Data B 5 1 Neutron eaf n fis and eaf n asscfy FISPACT II requires fission yield data if actinides are included in the input materi als eaf n fis
80. up to Z 111 are recognised and the new libraries contain 3873 nuclides ENDF format data libraries Capability to read and process TENDL 2011 TENDL 2012 and TENDL 2013 cross section uncertainty decay and fission yield li braries Self shielding Probability table data generated by CALENDF can now be used in conjunction with the 616 energy group EAF and the 709 energy group TENDL 2013 cross section data for neutron induced reactions to model dilution effects in the computation of collapsed cross sections c f Appendix A 4 3 Alternatively The universal sigmoid curve method for approximating self shielding can be em ployed using the MF 2 data in the TENDL 2013 files c f Appendix A 4 4 Additional decay types a total of 24 decay types are now recognised 7 single decay and 17 multiple particle decay modes CCFE Page 21 of 200 CCFE R 11 11 Issue 6 3 DIFFERENCES FROM FISPACT 2007 FisPACT Il User Manual Input file syntax checking Checking of the correctness of the input file and de tailed error reporting has been added to aid the development and testing of new input control files see Section 4 7 Additional PRINTLIB output it is now possible to specify the output of photon and material spectral lines Kerma dpa and gas appm Additional cross section data in the T ENDL 2013 cross section data files are read and processed to permit the output of derived rates of kinetic energy release displacements per atom
81. user interface of FISPACT II differs from that of FISPACT 2007 primarily in the use of command line arguments in running the program and the availability of more user friendly mnemonics and comments in the files file Users of FISPACT 2007 will need to make very few changes in the way in which they work to transfer to the new code and will benefit from the improved physical and numerical models the new code offers This section provides a simple step by step guide to running FISPACT II for those unfamiliar with the old code The new code follows the sequence used by its predecessors It has four stages to the prediction of inventories 1 process the library data e collapse cross section data e condense decay and fission data e print summary of library data 2 set initial conditions 3 run irradiation heating phases 4 run cooling phases All of these stages can be undertaken in a single run of FISPACT II but the library data processing produces intermediate binary files that can be reused for many inventory calculations However in this introduction we shall separate the parts of the library processing item 1 from the inventory calculation items 2 4 To run FIspAct II go to the directory in which the run input files are and type fispact lt fileroot gt or fispact lt fileroot gt lt files gt where it is assumed that FISPACT II has been installed and is on the user s path fileroot is the name root of all the user
82. 0 000E 00 U 233 FIS YIELD 3 289E 03 9 337E 04 1 409E 02 0 000E 00 2 079E 01 2 153E 03 CCFE Page 118 of 200 7 3 The Printlib Run output File CCFE R 11 11 Issue 6 FISPACT II User Manual Cm244 FIS YIELD 3 828E 03 1 170E 03 1 910E 02 0 000E 00 2 420E 01 0 000E 00 Cm245 FIS YIELD 3 607E 03 1 102E 03 1 850E 02 0 000E 00 2 280E 01 0 000E 00 Fraction of neutrons 200 keV 0 4120 fraction of neutrons gt 200 keV and 5 7 3 2 Branching ratios The second block gives percentage branching ratios for each decay mode of the ra dionuclides The parent and daughter nuclides are given with a code representing the decay between them These codes are summarised in Table I0 on page PERCENTAGE BRANCHING TIOS 000E402 Li 000E402 Li 5 p 000E 02 050E 01 Be 950E 01 Be 6 pp 000E 02 000E 02 B 000E402 b 700E 01 000E 02 Be 000E 02 bta 000E 02 842E 01 Be 580E 00 b 972E 01 000E 02 000E 00 b n 360E 01 000E 01 Be 300E 01 bta i 700E 01 000E 02 N 000E 02 b 000E 02 190E 01 N 160E 01 b n 840E 01 7 3 3 Cross sections The third section gives the effective cross section obtained by collapsing with the neu tron spectrum followed by the percentage error obtained by collapsing the cross section uncertainties Note that if there are no uncertainty data in the library then the keyword NOERROR switches the output in this section to include only the cross section The parent and daughter nucl
83. 000E 1 43 9 1000E 1 281 8 7640E 1 171 8 7643E 1 97 8 7641E 1 282 8 6000E 1 131 8 6000E 1 132 8 5000E 1 44 8 5000E 1 283 7 9000E 1 133 7 9000E 1 134 7 8000E 1 45 7 8000E 1 284 7 0500E 1 135 7 0500E 1 285 6 8260E 1 172 6 8256E 1 98 6 8255E 1 286 6 2500E 1 136 6 2500E 1 46 6 2500E 1 287 5 4000E 1 137 5 4000E 1 288 5 3160E 1 173 5 3158E 1 99 5 3157E 1 138 5 0000E 1 47 5 0000E 1 continued on next page CCFE Page 176 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FisPACT II User Manual continued from previous page TRIPOLI 315 VITAMIN J 175 GAMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 289 4 8500E 1 139 4 8500E 1 290 4 3300E 1 140 4 3300E 1 291 4 1400E 1 174 4 1399E 1 100 4 1399E 1 141 4 0000E 1 48 4 0000E 1 292 3 9100E 1 142 3 9100E 1 293 3 5200E 1 143 3 5000E 1 49 3 5000E 1 144 3 2000E 1 50 3 2000E 1 294 3 1450E 1 145 3 1450E 1 146 3 0000E 1 51 3 0000E 1 295 2 8250E 1 147 2 8000E 1 52 2 8000E 1 296 2 4800E 1 148 2 4800E 1 53 2 5000E 1 297 2 2000E 1 149 2 2000E 1 54 2 2000E 1 298 1 8900E 1 150 1 8900E 1 151 1 8000E 1 55 1 8000E 1 299 1 6000E 1 152 1 6000E 1 153 1 4000E 1 56 1 4000E 1 300 1 3400E 1 154 1 3400E 1 301 1 1500E 1 155 1 1500E 1 302 1 0000E 1 175 1 0000E 1 156 1 0000E 1 57 1 0000E 1 303 9 5000E 2 157 9 5000E 2 158 8 000
84. 00E 7 10 4 6000E 7 11 4 5000E 7 11 4 5000E 7 12 4 4000E 7 12 4 4000E 7 13 4 3000E 7 13 4 3000E 7 14 4 2000E 7 14 4 2000E 7 15 4 1000E 7 15 4 1000E 7 16 4 0000E 7 16 4 0000E 7 17 3 9000E 7 17 3 9000E 7 18 3 8000E 7 18 3 8000E 7 19 3 7000E 7 19 3 7000E 7 20 3 6000E 7 20 3 6000E 7 21 3 5000E 7 21 3 5000E 7 22 3 4000E 7 22 3 4000E 7 23 3 3000E 7 23 3 3000E 7 24 3 2000E 7 24 3 2000E 7 25 3 1000E 7 25 3 1000E 7 26 3 0000E 7 26 3 0000E 7 27 2 9000E 7 27 2 9000E 7 28 2 8000E 7 28 2 8000E 7 29 2 7000E 7 29 2 7000E 7 30 2 6000E 7 30 2 6000E 7 31 2 5000E 7 31 2 5000E 7 32 2 4000E 7 32 2 4000E 7 33 2 3000E 7 33 2 3000E 7 34 2 2000E 7 34 2 2000E 7 35 2 1000E 7 35 2 1000E 7 36 2 0000E 7 36 2 0000E 7 37 1 9640E 7 1 1 9640E 7 37 1 9640E 7 1 1 9640E 7 38 1 7330E 7 2 1 7330E 7 38 1 7330E 7 2 1 7330E 7 n 4 36 sd n I n 36 ee n 338 1 0000E 1 302 1 0000E 1 211 1 0000E 1 175 1 0000E 1 n 36 T n pd 351 1 1000E 4 315 1 1000E 4 352 1 0000E 5 316 1 0000E 5 212 1 0000E 5 176 1 0000E 5 Table 22 Energy group boundaries for the LANL 66 group structure LANL 66 group structure erp energy eV grp energy eV grp energy eV grp energy eV 1 2 5000E 7 18 3 0200E45 35 2 7540E42 52 8 0000E 2 2 2 0000E 7 19 1 8320E4 5 36 1 6700E 2 53 6 7000E 2 3 1 6905E 7 20 L1110E 5 37 1 0130E 2 54 5 8000E 2 4 1 4918E 7 21 6 7380E4 4 38 6 1440E 1 55 5 0000E 2 5 1 0000E 7 22
85. 01E45 223 7 5858E 2 377 6 3096E 1 531 5 2481E 4 70 8 70906E 5 224 7 2444E4 2 378 6 0256E 1 532 5 0119E 4 71 8 3176E45 225 6 9183E 2 379 5 7544E 1 533 4 7863bE 4 72 7 9433E 5 226 6 6069E 2 380 5 4954E 1 534 4 5709E 4 73 7 5858E 5 227 6 3096E 2 381 5 2481E 1 535 4 3652E 4 74 7 2444E 5 228 6 0256E 2 382 5 0119E 1 536 4 1687E 4 75 6 9183E4 5 229 5 7544E 2 383 4 7863E 1 537 3 9811E 4 T6 6 6069E 5 230 5 4954E 2 384 4 5709E 1 538 3 8019E 4 77 6 30906E45 231 5 2481E 2 385 4 3652E 1 539 3 6308E 4 78 6 0256E 5 232 5 0119E 2 386 4 1687E 1 540 3 4674E 4 79 5 7544E4 5 233 4 7863E 2 387 3 9811E 1 541 3 3113E 4 80 5 4954E4 5 234 4 5709E 2 388 3 8019E 1 542 3 1623E 4 81 5 2481E4 5 235 4 3652E 2 389 3 6308E 1 543 3 0200E 4 82 5 0119E4 5 236 4 1687E 2 390 3 4674E 1 544 2 8840E 4 83 4 7863E 5 237 3 9811E 2 391 3 3113E 1 545 2 7542bE 4 84 4 5709E 5 238 3 8019E 2 392 3 1623E 1 546 2 6303E 4 85 4 3652E 5 239 3 6308E 2 393 3 0200E 1 547 2 5119E 4 86 4 1687E 5 240 3 4674E 2 394 2 8840E 1 548 2 3988E 4 87 3 9811E 5 241 3 3113E 2 395 2 7542E 1 549 2 2909E 4 88 3 8019E 5 242 3 1623E 2 396 2 6303E 1 550 2 1878E 4 89 3 6308E 5 243 3 0200E 2 397 2 5119E 1 551 2 0893E 4 90 3 4674E 5 244 2 8840E 2 398 2 3988E 1 552 1 9953E 4 91 3 3113E4 5 245 2 7542E4 2 399 2 2909E 1 553 1 9055E 4 92 3 1623E 5 246 2 6303E 2 400 2 18
86. 04 26 73E 00 9657E 13 15 725400 4082E 04 23 35E 01 Sc 50 4522E 02 96 83E 02 3956E 04 10 07E 00 9369E 13 15 49E 00 6108E 04 12 81E 01 Ti 45 2266E 02 57 30E 02 0459E 04 75 50E 01 T 1 8 Bremsstrahlung correction If the BREMSSTRAHLUNG keyword is used then the Bremsstrahlung correc tion to the gamma dose is calculated using either plane or point source formulae see Appendix A 10 4 for details and are printed as shown below for test4 THE BREMSSTRAHLUNG CORRECTIONS ARE CALCULATED FOR AN INFINITE PLANE SOURCE Bremsstrahlung dose rate from Ar 39 is 1 96842E 07 Sv h 1 96842E 05 Rems h This is 3 49951E 10 of the total dose rate 7 1 9 Sensitivity output The SENSITIVITY keyword causes the generation of summary sensitivity output in the output file and full details are sent to the sens output file to allow further post processing The summary output for test35 is shown below The first part of the sensitivity output summarises the irradiation steps over which the sensitivity calculation is performed i e all steps before the ZERO keyword in the input file In the summary output for test35 shown below there is only one irradiation step but more generally a table of the steps is displayed c f pathways output below This is followed by a summary of the number of sample calculations CCFE Page 109 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FisPACT Il User Manual the number of par
87. 05 CCFE 2010 L W Packer and J Ch Sublet The European Activation File EAF 2010 decay data library Technical Report CCFE R 10 02 CCFE 2010 L W Packer and J Ch Sublet The European Activation File EAF 2010 biolog ical clearance and transport libraries Technical Report CCFE R 10 04 CCFE 2010 CCFE Page 127 of 200 CCFE R 11 11 Issue 6 REFERENCES FISPACT II User Manual 14 15 16 18 19 23 24 25 26 J Ch Sublet P Ribon and M Coste Delclaux CALENDF 2010 User Manual Technical Report CEA R 6277 ISSN 0429 3460 CEA 2011 J W Eastwood and J G Morgan FISPACT MP FISPACT II with Multiple Particle Irradiation Technical Report CEM 130516 SD 9 Issue 1 Culham Elec tromagnetics Ltd June 2014 M Herman and A Trkov editors ENDF 6 Formats Manual Data Formats and Procedures for the Evaluated Nuclear Data File ENDF B VI and ENDF B VII volume BNL 90365 2009 Rev 2 Brookhaven National Laboratory November 2011 K Radhakrishnan and A C Hindmarsh Description and use of LSODE the Liv ermore solver for ordinary differential equations Technical Report LLNL Report UCRL ID 113855 LLNL 1993 J W Eastwood and J G Morgan The FISPACT II Detailed Design Document Technical Report CEM 100421 5D 3 Issue 1 Culham Electromagnetics Ltd November 2010 J W Eastwood and J G Morgan The FISPACT II 12 Release 1 00 Detailed Design Document Technical Report CEM 100421 SD 3
88. 0E 00 3084E 00 6 2783E 00 3 1353E 01 0847E 00 9363E 00 7285E 00 1 6978E 00 4 4508E 02 7096E 00 0046E 00 1974E 00 2 4673E 00 1 6447E 00 3418E 00 5687E 01 9T05E 00 7 3 5 Projectile spectrum This table shows the energy bin boundaries and the flux in each bin for the neutron spectrum used to collapse the cross section library The available energy groups are tabulated in Appendix B 1 NEUTRON SPECTRUM Group Upper Lower Flux Group Upper Lower index energy energy index energy energy 491800E 07 349830E 07 97424E 13 651550E 04 737830E 04 02505E 12 349830E 07 221380E 07 88579E 12 737830E 04 247430E 04 20028E 12 221380E 07 105150E 07 16576E 12 247430E 04 086700E 04 39238E 12 356830E 05 227710E 05 67138E 12 825490E 01 315700E 01 00249E 11 227710E 05 110880E 05 06565E 12 315700E 01 139870E 01 25430E 11 110880E 05 651550E404 20848E 12 139870E 01 000000 05 09818E 12 Spectrum type is GAM II flux spectrum identifier is EEF FW NORM 1MW M2 GAM II TOT 4 277E414 7 3 6 Decay spectral lines Decay spectral lines are listed for unstable nuclides The decay type Table on page 138 and spectrum type Table 11 on page 139 line energy and line intensity for all unstable nuclides are displayed where data are available see keyword SAVE LINES For unstable nuclides without data the text no spectral data is displayed DECAY RADIATION DISCRETE SPECTRA NUCLIDE NUCLIDE NUCLIDE SPECTRUM DE
89. 0E 2 58 8 0000E 2 304 T 7000E 2 159 7 7000E 2 160 6 7000E 2 59 6 7000E 2 305 5 9000E 2 161 5 8000E 2 60 5 8000E 2 162 5 0000E 2 61 5 0000E 2 306 4 3000E 2 163 4 2000E 2 62 4 2000E 2 164 3 5000E 2 63 3 5000E 2 307 3 2380E 2 308 3 2000E 2 309 3 0000E 2 165 3 0000E 2 64 3 0000E 2 166 2 5000E 2 65 2 5000E 2 310 2 0000E 2 167 2 0000E 2 66 2 0000E 2 311 1 5000E 2 168 1 5000E 2 67 1 5000E 2 312 1 0000E 2 169 1 0000E 2 68 1 0000E 2 170 6 9000E 3 313 5 5000E 3 171 5 0000E 3 69 5 0000E 3 314 3 0000E 3 172 3 0000E 3 315 1 1000E 4 316 1 0000E 5 176 1 0000E 5 101 1 0000E 5 173 1 0000E 5 70 1 0000E 5 Table 21 Energy group boundaries for the two 55 MeV high energy standard structures TRIPOLI 351 TRIPOLI 315 VITAMIN J 211 VITAMIN J 175 grp energy eV grp energy eV grp energy eV grp energy eV 1 5 5000E4 7 1 5 5000E 4 7 2 5 4000E4 7 2 5 4000E4 7 3 5 3000E 7 3 5 3000E 7 continued on next page Page 177 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT 1 User Manual continued from previous page TRIPOLI 351 TRIPOLI 315 VITAMIN 3J 211 VITAMIN J 175 grp energy eV grp energy eV grp energy eV grp energy eV 4 5 2000E 7 4 5 2000E 7 5 5 1000E 7 5 5 1000E 7 6 5 0000E 7 6 5 0000E 7 7 4 9000E 7 7 4 9000E 7 8 4 8000E 7 8 4 8000E 7 9 4 7000E 7 9 4 7000E 7 10 4 60
90. 1 FISPACT Collapsing EAF 2007 Note that the format of the EAF 2010 and earlier cross section libraries does not embed the number of energy groups or the group boundaries in the library file so it is not possible to confirm the consistency of the specified ebins with the cross sections being used If they are not consistent erroneous results may be calculated without any warning from FISPACT II CCFE Page 49 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT 1 User Manual 5 1 8 GRPCONVERT nestre ndstre This keyword allows the user to read a neutron or other projectile spectrum in an arbitrary number of groups nestrc and instruct FISPACT II to convert it into one of the presently allowed eleven standard structures ndstrc must therefore be 66 69 100 162 172 175 211 315 351 616 or 709 using any other value will result in an error message l he user must prepare a file containing the following data and connect it to the arb_flux input stream in the files file e nestrc 1 values representing the arbitrary energy boundaries starting with the highest energy e nestrc values representing the flux values cm7 s in each group starting with the high energy group e First wall loading MW m e Text string maximum of 100 characters identifying the spectrum Note that each of the above groups of items should start on a new line in the file but there should be no blank lines separating them T
91. 1 0 1 W 100 0 In this case pure tungsten is specified 1kg of tungsten containing the five stable isotopes in their natural abundances is to be used in the self shielding calculation Note that SSFFUEL and SSFMASS must not both be used in the same case in the input file It is not essential that the total of all elements is exactly 100 However it is recom mended that the user ensures that the total percentage of all elements equals 100 The SSFMASS keyword in this section applies to the collapse calculation initiated by the FISPACT keyword The keyword may also appear in the inventory calculation section in conjunction with further GETXS keywords 5 2 Initial Conditions This section of the input file follows the keyword FISPACT If an inventory calcula tion is to follow this section must contain information about the particular material elemental or isotopic composition and mass and the initial conditions for the irradi ation history the first timestep and flux value Keywords that control the initialisation of the rate equation solution process must be placed in this section If pathways calculations are to be performed this section must also contain the key words that initialise the pathway calculations Similarly the keywords used for sensi tivity calculations must also be placed in this section Keywords that set the flux amplitude should be used in this section but may also occur in the inventory calculation sectio
92. 1 1 and these cover only a reduced energy range For the remainder the UKFY4 2 library then further extends the range before a neighbouring fission yield is used This UKFY4 2 library using Wahl s systematics is also used for all other particle induced fission yields C 3 Variance and Covariance Above the upper energy of the resolved resonance range for each of the 2434 isotopes a Monte Carlo method in which the covariance data come from uncertainties of the nuclear model calculations is used A complete description of the procedure is given in Reference 64 For all isotopes the initial best set of results is produced by a TALYS 1 4 calculation with an adjusted input parameter set This set of results is stored in ENDF files MF 3 to MF 10 Next for each isotope many TALYS runs with random nuclear model parameters are performed which are used to generate uncertainties and correlations As well as correlation within the same reaction channels correlation between reaction channels is included All information on cross section covariance is stored in the MF 33 format starting at the end of the resonance range up to 200 MeV Short range self scaling variance components are also specified for each MT type CCFE Page 198 of 200 C 4 Probability Tables CCFE R 11 11 Issue 6 FisPACT II User Manual The data format used to store the variance covariance information has been made fully compliant with the ENDF 6 format description and
93. 1E 02 3 1 10489E 26 8 72720E 03 1 10302E 26 8 60142E 03 4 1 53124E 27 5 40532E 03 1 52796E 27 5 36675E 03 j atoms base atoms mean atoms unc 1 2 50290E 20 2 49955E 20 2 46164E 02 2 7 99801E 18 7 99665E 18 1 68690E 03 3 9 91006E 18 9 90588E 18 8 55649E 03 4 9 87505E 15 9 87505E 15 2 77468E 06 5 9 19707E 13 9 19707E 13 6 54647E 12 6 3 99705E 20 3 99705E 20 3 81181E 06 7 2 75193E 16 2 75193E 16 6 16791E 06 Correlation coefficients j i 1 2 3 4 1 9 66468E 01 TEARS TUS AAA Be ee Ite Bu xz 9 99810E 01 3 xm 1 00000E 00 Be Se 4 a 9 99993E 01 7 1 The Inventory Run output File CCFE R 11 11 Issue 6 FISPACT II User Manual 9 99911E 01 5 6 AS 9 60898E 01 7 9 66478E 01 SS Ee um 7 1 10 Uncertainty estimates Sensitivity analysis provides uncertainties from an ensemble of calculations A faster approach is to use a sum of squares estimate from the errors in reactions on the path ways from the initial inventory to the dominant nuclides at the end of the irradiation phase see Appendix A 13 The uncertainty estimates of the form shown in the next output extract are computed from pathways for an UNCERTAINTY keyword pa rameter of 1 or 3 Presented first for each of the dominant nuclide categories are total values and their uncertainties UNCERTAINTY ESTIMATES sections only Uncertainty estimates are based on pathway analysis for the irradiation phase Total Activity is 1 25070E 14
94. 2007 FISPACT II User Manual ENFA still works but is deprecated use GETDECAY instead ERROR The ERROR sub keyword of keyword OVER has been replaced by AD CROSS to avoid conflict with the keyword ERROR used in sensitivity calcu lations GETDECAY is a replacement for ENFA and its subordinate keywords TAPA ARRAY and LINA GETXS is a replacement for the COLLAPSE and NEWFILE keywords GRAPH now has an additional option to write output suitable for gnuplot INDEXPATH has been added to allow the user to create a reduced nuclide index file containing only those nuclides that lie on pathways from the initial inventory nuclides to the dominant nuclides at the end of the irradiation phase LOOPS is now obsolete and its use generates a warning message to use the updated UNCERTAINTY keyword instead LOOKAHEAD is new keyword used to fine tune pathways calculations MCSAMPLE has been added to control parameters for the Monte Carlo sampling used in sensitivity calculations MCSEED has been added to allow users to specify the pseudo random number se quence for sensitivity calculations MIND now affects only the inventory output not the calculation NEWFILE is now obsolete and is ignored apart from generating a warning message Its functionality has been implemented with the GETXS keyword in association with multiple cross section files in the files file NOHEAD has been replaced by NOHEADER NOSTAB has been replaced by NOSTABLE
95. 3183E 1 465 1 0965E 2 4 1 8197E 7 158 1 5136E 4 312 1 2589E 1 466 1 0471E 2 5 1 7378E 7 159 1 4454E 4 313 1 2023E4 1 467 1 0000E 2 6 1 6596E 7 160 1 3804E 4 314 1 1482E 1 468 9 5499E 3 7 1 5849E 7 161 1 3183E 4 315 1 0965E 1 469 9 1201E 3 8 1 5136E 7 162 1 2589E 4 316 1 0471E 1 470 8 7096E 3 9 1 4454E 7 163 1 2023E 4 317 1 0000E 1 471 8 3176E 3 10 1 3804E 7 164 1 1482E 4 318 9 5499E 0 472 7 9433E 3 11 1 3183E 7 165 1 0965E 4 319 9 1201E 0 473 7 5858E 3 12 1 2589E47 166 1 0471E 4 320 8 7096E 0 474 7 2444E 3 13 1 2023E47 167 1 0000E 4 321 8 3176E 0 475 6 9183E 3 14 1 1482E 7 168 9 5499E 3 322 7 9433E 0 476 6 6069E 3 15 1 0965E 7 169 9 1201E 3 323 7 5858E 0 477 6 3096E 3 16 1 0471E 7 170 8 7096E 3 324 7 2444E 0 478 6 0256E 3 17 1 0000E 7 171 8 3176E 3 325 6 9183E 0 479 5 7544E 3 18 9 5499E 6 172 7 9433E 3 326 6 6069E 0 480 5 4954E 3 19 9 1201E4 6 173 7 5858E 3 327 6 3096E 0 481 5 2481E 3 20 8 7096E 6 174 7 2444E 3 328 6 0256E 0 482 5 0119E 3 21 8 3176E 6 175 6 9183E 3 320 5 7544E 0 483 4 7863E 3 22 7 9433E4 6 176 6 6069E 3 330 5 4954E 0 484 4 5709E 3 23 7 5858E4 6 177 6 3096E 3 331 5 2481E 0 485 4 3652E 3 24 7 2444E4 6 178 6 0256E 3 332 5 0119E 0 486 4 1687E 3 25 6 9183E46 179 5 7544E 3 333 4 7863E 0 487 3 9811E 3 26 6 6069E 6 180 5 4954E4 3 334 4 5709E 0 488 3 8019E 3 27 6
96. 3E 4 102 1 1600E 7 280 3 9811E 3 458 1 0965E 0 636 3 0200E 4 103 1 1400E4 7 281 3 8019E 3 459 1 0471E 0 637 2 8840E 4 104 1 1200E 7 282 3 6308E 3 460 1 0000E 0 638 2 7542E 4 105 1 1000E 7 283 3 4674E 3 461 9 5499E 1 639 2 6303E 4 106 1 0800E4 7 284 3 3113E 3 462 9 1201E 1 640 2 5119E 4 107 1 0600E4 7 285 3 1623E 3 463 8 7096E 1 641 2 3988E 4 108 1 0400E 7 286 3 0200E 3 464 8 3176E 1 642 2 2909E 4 109 1 0200E4 7 287 2 8840E 3 465 7 9433E 1 643 2 1878E 4 110 1 0000E4 7 288 2 7542E 3 466 7 5858E 1 644 2 0893E 4 111 9 5499E 6 289 2 6303E 3 467 7 2444E 1 645 1 9953E 4 112 9 1201E 6 290 2 5119E 3 468 6 9183E 1 646 1 9055E 4 113 8 7096E4 6 201 2 3988E 3 469 6 6069E 1 647 1 8197E 4 114 8 3176E 6 202 2 2909E 3 470 6 3096E 1 648 1 7378E 4 115 7 9433E4 6 203 2 1878E 3 471 6 0256E 1 649 1 6596E 4 116 7 5858E 6 294 2 0893E 3 472 5 7544E 1 650 1 5849E 4 117 7T 2444E4 6 205 1 9953E 3 473 5 4954E 1 651 1 5136E 4 118 6 9183E 6 296 1 9055E 3 474 5 2481E 1 652 1 4454E 4 119 6 6069E 6 207 1 8197E 3 475 5 0119E 1 653 1 3804E 4 120 6 3096E 6 208 1 7378E 3 476 4 7863E 1 654 1 3183E 4 121 6 0256E 6 299 1 6596E 3 477 4 5709E 1 655 1 2589E 4 122 5 7544E 6 300 1 5849E 3 478 4 3652E 1 656 1 2023E 4 123 5 4954E46 301 1 5136E 3 479 4 1687E 1 657 1 1482E 4 124 5 2481E 6 302 1 4454E 3 480 3 9811E 1 658 1 0965E 4 125 5 0119E4
97. 3E 5 111 1 3183E45 265 1 0965E 2 419 9 1201E 2 573 7 5858E 5 112 1 2589E 5 266 1 0471E 2 420 8 7096E 2 574 7 2444E 5 113 1 2023E45 267 1 0000E 2 421 8 3176E 2 575 6 9183E 5 114 1 1482E4 5 268 9 5499E4 1 422 7 9433E 2 576 6 6069E 5 115 1 0965E 5 269 9 1201E 1 423 7 5858E 2 577 6 3096E 5 116 1 0471E 5 270 8 7096E 1 424 7 2444E 2 578 6 0256E 5 117 1 0000E 5 271 8 3176E 1 425 6 9183E 2 579 5 7544E 5 118 9 5499E 4 272 7 9433E 1 426 6 6069E 2 580 5 4954E 5 119 9 1201E 4 273 7 5858E 1 427 6 3096E 2 581 5 2481E 5 120 8 7096E 4 274 7 2444E 1 428 6 0256E 2 582 5 0119E 5 121 8 3176E 4 275 6 9183E 1 429 5 7544E 2 583 4 7863E 5 122 7 9433E4 4 276 6 6069E 1 430 5 4954E 2 584 4 5709E 5 123 7 5858E4 4 277 6 3096E 1 431 5 2481E 2 585 4 3652E 5 124 7 2444E 4 278 6 0256E 1 432 5 0119E 2 586 4 1687E 5 125 6 9183E 4 279 5 7544E 1 433 4 7863E 2 587 3 9811E 5 126 6 6069E 4 280 5 4954E 1 434 4 5709E 2 588 3 8019E 5 127 6 30906E 4 281 5 2481E 1 435 4 3652E 2 589 3 6308E 5 128 6 0256E 4 282 5 0119E 1 436 4 1687E 2 590 3 4674E 5 129 5 7544E 4 283 4 7863E 1 437 3 9811E 2 591 3 3113E 5 130 5 4954E 4 284 4 5709E 1 438 3 8019E 2 592 3 1623E 5 131 5 2481E 4 285 4 3652E 1 439 3 6308E 2 593 3 0200E 5 132 5 0119E 4 286 4 1687E 1 440 3 4674E 2 594 2 8840E 5 133 4 7863E44 287 3 9811E 1 441 3 3113E 2 595 2 7542E 5 134
98. 4 4 8050E 1 218 4 7850E 1 155 4 7851E 1 81 4 7850E 1 219 4 5520E 1 73 4 5517E 1 220 3 9810E 1 74 4 0169E 1 221 3 7270E 1 156 3 7267E 1 82 3 7266E 1 75 3 7267E 1 222 3 3890E 1 76 3 3720E 1 223 3 0510E 1 77 3 0511E 1 224 2 9200E 1 157 2 9023E4 1 83 2 9023E 1 225 2 7920E4 1 78 2 7608E 1 25 2 7700E 1 226 2 4980E 1 79 2 4981E 1 227 2 2600E4 1 158 2 2603E 1 84 2 2608E 1 80 2 2603E 1 228 2 0450E 1 229 1 9030E 1 81 1 9455E 1 230 1 7600E 1 159 1 7604E 1 85 1 7603E 1 231 1 6740E4 1 82 1 5928E 1 26 1 5970E 1 232 1 5230E 1 233 1 9710E 1 160 1 3710E 1 86 1 3709E 1 83 1 3710E 1 234 1 2590E 1 235 1 1220E 1 84 1 1225E 1 236 1 0680E 1 161 1 0677E 1 87 1 0677E 1 237 1 0000E4 1 85 9 9056E 0 27 9 8770E 0 238 9 1900E 0 86 9 1898E 0 239 8 9130E 0 240 8 3150E 0 162 8 3153E 0 88 8 3152E 0 87 8 3153E 0 241 7 9430E 0 242 7 5240E 0 88 7 5240E 0 243 7 0790E 0 244 6 4760E 0 163 6 4760E4 0 89 6 4758E 0 245 6 1600E 0 89 6 1601E 0 246 5 6230E 0 90 5 3464E4 0 247 5 0430E 0 164 5 0435E 0 90 5 0434E 0 91 5 0435E 0 248 4 6700E 4 0 249 4 4700E 0 250 4 1290E 0 92 4 1293E 0 93 4 0000E 0 28 4 0000E4 0 251 3 9280E 0 165 3 9279E 0 91 3 9278E 0 252 3 3810E 0 94 3 3808E 0 95 3 3000E 0 29 3 3000E 0 253 3 0590E 0 166 3 0590E 0 92 3 0590E4 0 continued on next page CCFE Page 175 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPA
99. 4 5709E 4 288 3 8019E 1 442 3 1623E 2 596 2 6303E 5 135 4 3652E 4 289 3 6308E 1 443 3 0200E 2 597 2 5119E 5 136 4 1687E 4 290 3 4674E 1 444 2 8840E 2 598 2 3988E 5 137 3 9811E4 4 291 3 3113E 1 445 2 7542E 2 599 2 2909E 5 138 3 8019E 4 292 3 1623E 1 446 2 6303E 2 600 2 1878E 5 139 3 6308E 4 293 3 0200E 1 447 2 5119E 2 601 2 0893E 5 140 3 4674E4 4 294 2 8840E 1 448 2 3988E 2 602 1 9953E 5 141 3 3113E44 295 2 7542E 1 449 2 2909E 2 603 1 9055E 5 142 3 1623E 4 296 2 6303E 1 450 2 1878E 2 604 1 8197E 5 143 3 0200E44 297 2 5119E 1 451 2 0893E 2 605 1 7378E 5 144 2 8840E 4 298 2 3988E 1 452 1 9953E 2 606 1 6596E 5 continued on next page CCFE Page 182 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FISPACT II User Manual continued from previous page LLNL 616 group structure erp energy eV grp energy eV grp energy eV grp energy eV 145 2 7542E 4 299 2 2909E 1 453 1 9055E 2 607 1 5849E 5 146 2 6303E 4 300 2 1878E 1 454 1 8197E 2 608 1 5136E 5 147 2 5119E44 301 2 0893E 1 455 1 7378E 2 609 1 4454E 5 148 2 3988E 4 302 1 9953E 1 456 1 6596E 2 610 1 3804E 5 149 2 2909E 4 303 1 9055E 1 457 1 5849E 2 611 1 3183E 5 150 2 1878E 4 304 1 8197E 1 458 1 5136E 2 612 1 2589E 5 151 2 0893E 4 305 1 7378E 1 459 1 4454E 2 613 1 2023E 5 152 1 9953E 4 306 1 6596E
100. 5 R Sc 44 S This generic pathway is the sum of 3 pathways The path floor is pmin as a percentage of the number of atoms of the target nuclide 4Sc and max_depth is set by nmaz For an interpretation of the output see page 5 2 44 SENSITIVITY zxsens ansen1 insen3 insen4 parent i daughter i i 1 insen3 nuclide j j 1 insen4 This keyword allows sensitivity calculations to be performed The sensitivity Monte Carlo calculation is undertaken over all the irradiation steps and is initiated by the ZERO keyword Time dependent flux amplitude flux spectra and cross sections are permitted in sensitivity runs If zsens LAMBDA then the sensitivity coefficients with respect to decay constant are calculated If xsens SIGMA then the sensitivity coefficients with respect to cross section are calculated However only one of these options can be specified for a case the keyword must not be input twice In the current version the LAMBDA option is not available The cut off value ansen1 is the magnitude of the correlation coefficient lt 1 0 value below which results are not printed typical value may be 0 8 The independent variables for the monte carlo calculations are the reactions defined by insen3 parent daughter pairs To include fission use the name Fission or number 0 for the daughter nuclide name For each of the insen4 nuclides specified the sensitivity of that nuclide to each of the insen3 cross sections o
101. 5E 5 374 5 2481E 1 552 1 4454E 2 19 2 8000E 8 197 1 8197E4 5 375 5 0119E 1 553 1 3804E 2 20 2 4000E 8 198 1 7378E 5 376 4 7863E 1 554 1 3183E 2 21 2 0000E 8 199 1 6596E 5 377 4 5709E 1 555 1 2589E 2 22 1 8000E 8 200 1 5849E 5 378 4 3652E 1 556 1 2023E 2 23 1 6000E 8 201 1 5136E 5 379 4 1687E 1 557 1 1482E 2 24 1 5000E 8 202 1 4454E 5 380 3 9811E 1 558 1 0965E 2 25 1 4000E 8 203 1 3804E 5 381 3 8019E 1 559 1 0471E 2 26 1 3000E 8 204 1 3183E 5 382 3 6308E 1 560 1 0000E 2 27 1 2000E 8 205 1 2589E 5 383 3 4674E 1 561 9 5499E 3 28 1 1000E 8 206 1 2023E 5 384 3 3113E4 1 562 9 1201E 3 29 1 0000E 8 207 1 1482E 5 385 3 1623E 1 563 8 7096E 3 30 9 0000E4 7 208 1 0965E 5 386 3 0200E 1 564 8 3176E 3 31 8 0000E 7 209 1 0471E 5 387 2 8840E 1 565 7 9433E 3 32 7 5000E 7 210 1 0000E 5 388 2 7542E 1 566 7 5858E 3 33 7 0000E 7 211 9 5499E 4 389 2 6303E 1 567 7 2444E 3 34 6 5000E 7 212 9 1201E 4 390 2 5119E 1 568 6 9183E 3 35 6 0000E 7 213 8 7096E 4 391 2 3988E 1 569 6 6069E 3 continued on next page CCFE Page 183 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT 1 User Manual continued from previous page CCFE 709 group structure erp energy eV grp energy eV grp energy eV grp energy eV 36 5 8000E 7 214 8 3176E 4 392 2 2909E 1 570 6 3096E 3 37 5 6000E 7 215 7 9433E 4 393 2 1878E 1
102. 6 303 1 3804E 4 3 481 3 8019E 1 659 1 0471E 4 126 4 7863E4 6 304 1 3183E 3 482 3 6308E 1 660 1 0000E 4 127 4 5709E4 6 305 1 2589E 3 483 3 4674E 1 661 9 5499E 5 128 4 3652E 6 306 1 2023E 3 484 3 3113E 1 662 9 1201E 5 129 4 1687E 6 307 1 1482E 3 485 3 1623E 1 663 8 7096E 5 130 3 9811E4 6 308 1 0965E 3 486 3 0200E 1 664 8 3176E 5 131 3 8019E4 6 309 1 0471E 3 487 2 8840E 1 665 7 9433E 5 132 3 6308E 6 310 1 0000E 3 488 2 7542E 1 666 7 5858E 5 133 3 4674E4 6 311 9 5499E 2 489 2 6303E 1 667 7 2444E 5 134 3 3113E46 312 9 1201E 2 490 2 5119E 1 668 6 9183E 5 135 3 1623E 6 313 8 7096E 2 491 2 3988E 1 669 6 6069E 5 136 3 0200E4 6 314 8 3176E 2 492 2 2909E 1 670 6 3096E 5 137 2 8840E4 6 315 7 9433E 2 493 2 1878E 1 671 6 0256E 5 138 2 7542E4 6 316 7 5858E 2 494 2 0893E 1 672 5 1544E 5 139 2 6303E 6 317 7 2444E 2 495 1 9953E 1 673 5 4954E 5 continued on next page CCFE Page 185 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT 1 User Manual continued from previous page CCFE 709 group structure erp energy eV grp energy eV grp energy eV grp energy eV 140 2 5119E 6 318 6 9183E 2 496 1 9055E 1 674 5 2481E 5 141 2 3988E 6 319 6 6069E 2 497 1 8197E 1 675 5 0119E 5 142 2 2909E 6 320 6 3096E 2 498 1 7378E 1 676 4 7863E 5 143 2 1878E 6 321 6 0256E 2 499 1 6596E 1 677 4
103. 6 03 10 22 May 2013 Log FILES file files 7 2 The Inventory Run runlog File CCFE R 11 11 Issue 6 FISPACT II User Manual Log fileroot inventory The numbers after the unit names are the internal unit numbers input 5 inventory i crossunc 7 EAF2010data eaf un 20100 asscfy 8 EAF2010data eaf n asscfy 20100 A copy of the run monitoring information see MONITOR keyword is written to the runlog Settings keywords are simply echoed and action keywords e g ATOMS are followed by summary messages for the actions they initiate NOHEADER MONITOR 1 GETXS 0 GETDECAY 0 FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 load cross sections load decay data collapse fission yields run reset cross section MASS 1 0 1 TI 100 0 FLUX 4 27701E 14 MIND 1 E5 GRAPH 2 2 1 1 4 UNCERTAINTY 2 ATOMS load initial values run output inventory HAZARDS load hazards data HALF ATWO load a2 data TIME 2 5 fill rate equation matrix for cooling fill rate equation matrix for irradiation start pathstep recording initialise dominant analysis test for gas kerma and dpa data YEARS ATOMS run add rateeq for pathways CCFE Page 115 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual run irradiation init run irradiation step run add pathstep run output inventory FLUX 0 ZERO TIME 1 MINS ATOMS run pathways initialisation run pathways uncertainty run cooling
104. 78E 1 554 1 8197E 4 continued on next page CCFE Page 181 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual B EAF LIBRARY DATA continued from previous page LLNL 616 group structure erp energy eV grp energy eV grp energy eV grp energy eV 93 3 0200E 5 247 2 5119E 2 401 2 0893E 1 555 1 7378E 4 94 2 8840E 5 248 2 3988E 2 402 1 9953E 1 556 1 6596E 4 95 2 7542E 5 249 2 2909E4 2 403 1 9055E 1 557 1 5849E 4 96 2 6303E 5 250 2 1878E 2 404 1 8197E 1 558 1 5136E 4 97 2 5119E45 251 2 0893E 2 405 1 7378E 1 559 14454E 4 98 2 3988E 5 252 1 9953E 2 406 1 6596E 1 560 1 3804E 4 99 2 2909E 5 253 1 9055E 2 407 1 5849E 1 561 1 3183E 4 100 2 1878E45 254 1 8197E 2 408 1 5136E 1 562 1 2589E 4 101 2 0898E45 255 1 7378E 2 409 1 4454E 1 563 1 2023E 4 102 1 9953E 5 256 1 6596E 2 410 1 3804E 1 564 1 1482E 4 103 1 9055E4 5 257 1 5849E 2 411 1 3183E 1 565 1 0965E 4 104 1 8197E4 5 258 1 5136E 4 2 412 1 2589E 1 566 1 0471E 4 105 1 7378E 5 259 1 4454E 2 413 1 2023E 1 567 1 0000E 4 106 1 6596E 5 260 1 3804E 2 414 1 1482E 1 568 9 5499E 5 107 1 5849E45 261 1 3183E 2 415 1 0965E 1 569 9 1201E 5 108 1 5136E 5 262 1 2589E 2 416 1 0471E 1 570 8 7096E 5 109 1 4454E4 5 263 1 2023E42 417 1 0000E 1 571 8 3176E 5 110 1 3804E 5 264 1 1482E 2 418 9 5499E 2 572 7 943
105. 81 3 6883E4 5 38 3 6883E 5 80 3 3370E 5 82 3 3373E4 5 39 3 3373E 5 81 3 0200E4 5 83 3 0197E4 5 40 3 0197E 5 29 3 0197E4 5 8 3 0250E 5 82 2 9850E 5 84 2 9849E4 5 83 2 9720E4 5 85 2 9721E 5 84 2 9450E 5 86 2 9452E4 5 85 2 8730E 5 87 2 8725E 5 86 2 7320E4 5 88 2 7324E 5 41 2 7323E45 30 2 7324E 5 87 2 4720E 5 89 2 4724E 5 42 2 4723E 5 31 2 4724E 5 88 2 38520E 5 90 2 3518E4 5 89 2 23870E4 5 91 2 2371E 5 43 2 2370E 5 90 2 1280E 5 92 2 1280E 5 91 2 0240E 5 93 2 0242E4 5 44 2 0242E4 5 92 1 9250E 5 94 1 9255E 5 93 1 8320E 5 95 1 8316E 5 45 1 8315E 5 32 1 8316E 5 9 1 8300E 5 94 1 7420E 5 96 1 7422E 5 95 1 6570E 5 97 1 6573E 5 46 1 6572E 5 96 1 5760E 5 98 1 5764E 5 97 1 5000E 5 99 1 4996E 5 4T 1 4995E 5 98 1 4260E 5 100 1 4264E 5 99 1 3570E 5 101 1 3569E 5 48 1 3568E 5 100 1 2910E 5 102 1 2907E 5 101 1 2280E 5 103 1 2277E 5 49 1 2277E4 102 1 1680E 5 104 1 1679E 5 103 1 1110E4 5 105 1 1109E 5 50 1 1109E 5 34 1 1109E4 5 10 1 1100E 5 104 9 8040E 4 106 9 8037E4 4 105 8 6520E 4 107 8 6517E 4 51 8 6516E 4 106 8 2500E 4 108 8 2503E 4 107 8 2300E 4 35 8 2298E 4 108 7 9500E 4 109 1 9499E4 4 5 133 1 2277E 5 continued on next page CCFE Page 172 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FisPACT II User Manual continued from previous page TRIPOLI 315 VITAMIN J 175 G
106. 94E 07 SECS OR 2 5000E 00 YEARS ELAPSED TIME IS 2 500 y NUCLIDE ATOMS GRAMS Bq b Energy a Energy g Energy DOSE RATE INGESTION INHALATION Bq A2 kW kW kW Sv hr DOSE Sv DOSE Sv Ratio 8 11507E 21 1 358E 02 0 000E 00 0 000E 00 0 00E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 3 39818E 20 1 137E 03 0 000E 00 0 000E 00 0 00E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 1 92678E 17 9 650E 07 3 432E 08 3 139E 10 0 00E 00 0 000E 00 0 000E 00 1 442E 02 8 924E 02 8 581E 06 5 62944E 15 2 819E 08 0 000E 00 0 000E 00 0 00E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 2 57627E 21 1 712E 02 0 000E 00 0 000E 00 0 00E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 3 37819E 05 1 569E 17 0 000E 00 0 000E 00 0 00E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 The inventory contains up to eleven columns of data excluding the nuclide identifier and flags giving values at the end of the step indicated by the time line The first seven of these are always printed and their contents are defined in Table 7 In that table Ar atomic weight of isotope i amu Na Avogadro constant mol Eg B decay energy for isotope i eV Eo i a decay energy for isotope i eV E y decay energy for isotope i eV C conversion from eV to kJ 107 e Note that the atomic weights read from the eaf_decay file are in units of neutron masses but these are converted and stored internally in amu Note also that the column h
107. AMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 109 7 4990E 4 110 7 2000E 4 110 7 2025E 4 111 6 7380E 4 111 6 7380E4 4 52 6 1378E A 36 60 7380E 4 11 6 7340E 4 112 6 1730E 4 113 5 6560E 4 112 5 6562E 4 114 5 5170E 4 37 5 5166E 4 115 5 2480E 4 113 5 2475E 4 53 5 2474E 4 116 4 9390E 4 117 4 6310E 4 114 4 6309E 4 118 4 3590E 4 119 4 0870E 4 115 4 0868E 4 54 4 0867E4 A 38 4 0868E 4 12 4 0850E 4 120 3 6980E 4 39 3 6979E 4 121 3 4310E 4 116 3 4307E 4 122 3 1830E 4 117 3 1828E 4 55 3 1827E4 123 3 1620E 4 124 3 0730E 4 125 2 9850E 4 126 2 9010E 4 40 2 9283E 4 127 2 8500E 4 118 2 8501E 4 128 2 8180E 4 129 2 7380E 4 41 2 7394E 4 130 2 7000E 4 119 2 7000E4 4 131 2 6610E 4 132 2 6060E 4 120 2 6058E 4 133 2 5850E 4 134 2 5120E 4 135 2 4790E 4 121 2 4788E 136 2 4410E 4 137 2 4180E 4 122 2 4176E 4 138 2 3580E 4 123 2 3579E 4 139 2 3040E 4 140 2 2390E 4 141 2 1870E 4 124 2 1875E4 4 142 2 1130E 4 143 2 0540E 4 4 144 1 9950E 4 145 1 9310E 4 125 1 9305E 4 57 1 9304E 4 146 1 7780E 4 147 1 6620E 4 43 1 6616E 4 148 1 5850E 4 149 1 5030E 4 126 1 5034E4 150 1 3830E 4 151 1 2730E 4 152 1 1710E 4 127 1 1709E 4 59 1 1709E 4 153 1 1140E 4 45 1 1138E 4 128 1 0595E 4 T T 4 4 56 2 4787E 4 42 2 4788E 4 13 2 4780E4 4 4 58 1 5034E 4 44 1 5034E 4 14 1 5030E 4 154 1 0080E 4 155 9 1190E 3
108. ANCE e 5 2 59 UNCERTAINTY e 5 200 UNCTIYPE ucc 442445645 e kon aes 5 2 61 USEFISSION 22s 5 2 02 WALL 453 2 uk se Re heces S a SU S adea 86 Inventory Calculation Phase uo 4 oy ee y A XE SUR OS 5 9 1 ATOMS ea 3 util BES xe Id e Peg m tex BAS 5 93 2 EAFVERSI N 2 624 2 48 2a 9x ok 93 X RO a Dio IND Lux cs e E Ee ewe EEE Xeno oue OE SOR CER E 5 9 4 ENDPULSE sedia s s demere 9 24 PUR a Reds 530 FLUX 44e RR noto diese doe de X Ro e m Ree os 5 3 6 FULLXS 2 2 2 2 2 2 242 2225222 25 a CGETAS 22533 kw Soe cee XE ola eundem E Sues RIS xm 5 3 8 GRPCONVERT 2 2 2 2 2 22225D52 190 5 3 9 LOGLEVEL e 5 3 10 NOSTABLE ce arresi Dodl NOTI coa a a dde e a eo RR RR eR os 5 93 12 NOTI uu eoo lio a aa a dd 53 19 NOUS iss oot ox GR Gees Rl e RENE AR do ORO S He X S 91 5 9 14 NOTA aa n dere rt dn x E eur ed rh HR ee ie s 5 5 15 OVER mida a xaxd a gie Wow eae a 5 3 16 PARTITION 2 2 2 2 2 2222 s 5 3 17 PATHRESET 2s 5 3 18 PROBTABLE 2 2 2 2 2 2 2 2D202 D5D2 0 5 9 10 PULSE 2 co km oso oum om RO ok R Porch Ox dos 0 93 20 RESULT x gnc uk vas Be EUER S EO da Rd CCFE Page 9 of 200 CCFE R 11 11 Issue 6 CONTENTS FISPACT II User Manual 5 3 21 SPECTRUM 1 e 94 Died DOP MOOSE 22 2x 32 4 y A Ro yn e Red e d 94 5 3 23 SSFDILUTION 2 es 94 5 9 24 SORE UBL a tw ee SH Im e
109. CAY LINE ENERGY INTENSITY NAME ZAI NUMBER TYPE TYPE eV eV H 3 10030 3 beta b 1 85710E 04 1 85710E 04 He 6 20060 6 beta b 3 50700E 06 3 50700E 06 Li 5 30050 7 no spectral data 7 4 Probability Table Collapse Run CCFE R 11 11 Issue 6 FISPACT II User Manual i 8 30080 10 beta b 1 29650E 07 1 29650E 07 alpha a 1 56600E 06 1 56600E 06 i 9 30090 11 beta b 2 32000E 06 9 28000E 04 beta b 5 67000E 06 8 50500E 04 beta b 1 08300E 07 1 08300E 06 beta b 1 11770E 07 3 80018E 06 beta b 1 36060E 07 6 87103E 06 6 40060 12 no spectral data 7 4 Probability Table Collapse Run The cross section collapse with probability table data to compute the self shielding factor and the effective collapsed cross sections differs from the standard collapse c f Section 4 2 in that T 2 a mapping for the probability table data directory must be added to the files file the reading of the probability table data must be activated by including the PROBTABLE keyword in the library preparation section of the input file the set of parent nuclides or elements to which the self shielding factor is to be applied is specified by the SSFCHOOSE keyword the mixture of nuclides to be included in the dilution computation must be specified using either the SSFMASS of SSFFUEL keywords Note that the values specified for these may be different from the M ASS or FUEL mixtures specified in the inventory run
110. CT II User Manual continued from previous page TRIPOLI 315 VITAMIN J 175 GAMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 254 2 7680E 0 96 2 7679E 0 97 2 7200E 0 98 2 6000E 0 30 2 6000E 0 99 2 5500E 0 255 2 3720E 0 167 2 3824E4 0 93 2 8823E 0 256 2 3600E 0 100 2 3600E 0 257 2 1300E 0 101 2 1300E4 0 102 2 1000E 0 31 2 1000E 0 258 2 0200E 4 0 103 2 0200E 0 259 1 9300E 4 0 104 1 9300E 0 260 1 8550E 0 168 1 8554E 0 94 1 8554E 0 261 1 8400E 0 105 1 8400E4 0 262 1 7550E 0 106 1 7550E4 0 263 1 6700E 0 107 1 6700E 0 264 1 5900E 0 108 1 5900E 0 265 1 5100E 4 0 109 1 5000E 0 32 1 5000E 0 110 1 4750E4 0 266 1 4450E 0 169 1 4450E 0 95 1 4450E 0 111 1 4450E 0 267 1 4400E 0 268 1 3700E 0 112 1 3700E 0 113 1 3375E 0 269 1 3050E 4 0 114 1 3000E 0 33 1 3000E 0 270 1 2350E 0 115 1 2350E4 0 271 1 1700E 0 116 1 1700E4 0 117 1 1500E 0 34 1 1500E 0 272 1 1250E 0 170 1 1254E 0 96 1 1253E 0 118 1 1254E4 0 35 1 1230E 0 273 1 1100E4 0 119 1 1100E 0 120 1 0970E 0 36 1 0970E 0 274 1 0900E 0 275 1 0800E 0 276 1 0700E 0 121 1 0710E 0 37 1 0710E 0 122 1 0450E 0 38 1 0450E 0 277 1 0350E 0 123 1 0350E4 0 124 1 0200E 0 39 1 0200E 0 278 1 0100E 0 125 9 9600E 1 40 9 9600E 1 279 9 8600E 1 126 9 8600E 1 127 9 7200E 1 41 9 7200E 1 128 9 5000E 1 42 9 5000E 1 280 9 3000E 1 129 9 3000E 1 130 9 1
111. CULHAM CENT FUSION ENER HO RE G Y CCFE R 11 11 Issue 6 June 2014 Jean Christophe C Sublet James W Eastwood J Guy Morgan The FISPACT II User Manual CCFE R 11 11 Issue 6 FISPACT II User Manual This document is intended for publication in the open literature It is made available on the un derstanding that it may not be further circulated and extracts or references may not be published prior to publication of the original when applicable or without the consent of the Publications Officer CCFE Library Culham Science Centre Abingdon Oxon OX14 3DB UK Enquiries about Copyright and reproduction should be addressed to the Culham Publications Officer CCFE Library Culham Science Centre Abingdon Oxon OX14 3DB UK CCFE Page 2 of 200 CCFE R 11 11 Issue 6 The FisPACT ll User Manual Jean Christophe C Sublet James W Eastwood J Guy Morgan June 2014 UK Atomic Energy Authority Culham Science Centre Abingdon Oxfordshire OX14 3DB 1 Culham Electromagnetics Ltd Culham Science Centre Abingdon OX14 3DB UKAS AQ UKAS Ka ats he MUS ER 150 9001 a E 150 14001 CCFE Page 3 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual Contact Dr Jean Christophe Sublet UK Atomic Energy Authority Culham Science Centre Abingdon Oxfordshire OX14 3DB United Kingdom Telephone 44 0 1235 466400 Facsimile 44 0 1235 463435 email Jean Christoph
112. DES does not attempt this error estimation Consequently users should set rtol and atol cautiously A 14 5 Runtime error reporting FISPACT II traps any error returns from LSODES and reports them to the user with the same error logging system used elsewhere in the code Extensive testing of FISPACT II demonstrates that LSODES is robust and no abnormal returns from it should be expected Exceptionally the abnormal error returns that may be encountered are listed in Ta ble A fuller explanation of the meanings of these error returns may be found in the exten sive comments at the head of the source files dlsodes f and slsodes f CCFE Page 167 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual A THE MODEL Table 19 The abnormal error returns from LSODES ISTATE Brief explanation User action 7 sparse solver problem should not Report error to developers happen 6 zero variable with pure relative er rerun with atol 40 ror control 5 repeated convergence failures bad Report error to developers Jacobian 4 repeated error test failures bad in Report error to developers put 3 illegal input Report error to developers 2 requested accuracy too great rerun with larger rtol and or atol 1 excessive work too many internal rerun with larger rtol and or steps atol CCFE Page 168 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual B EAF Library Data F
113. DF library data that can be read by Version 2 of the code 2 What FISPACT II Does FISPACT II is a practical activation transmutation engineering prediction tool The four principal tasks that it undertakes are CCFE Page 16 of 200 2 1 Library Data Preparation CCFE R 11 11 Issue 6 FISPACT 1 User Manual 1 extraction reduction and storage of nuclear and radiological data from the EAF or ENDF library files 2 construction and solution of the rate equations to determine the time evolution of the inventory in response to different irradiation scenarios These scenarios include a a cooling only calculation b a single irradiation pulse followed by cooling c multiple irradiation pulses where only flux amplitudes change followed by cooling d multi step irradiation where flux amplitude flux spectra and cross sections may change followed by cooling 3 computation and output of derived radiological quantities 4 subsidiary calculations to identify the key reactions and decays and to assess the quality of the predictions The four main subsidiary items are a pathways analysis n reduced model calculations b uncertainty calculations from pathways c d Lom monte carlo sensitivity and uncertainty calculations These items are described further in the following subsections 2 1 Library Data Preparation The library preparation task comprises reading and collapsing
114. E 4 29 7 0000E 7 70 6 4000E 6 111 8 7500E 5 151 6 5000E4 4 30 6 0000E 7 71 6 2000E 6 112 8 5000E 5 152 6 0000E4 4 31 5 5000E 7 72 6 0000E 6 113 8 2500E 5 153 5 5000E 4 32 5 4000E 7 73 5 8000E 6 114 8 0000E 5 154 5 0000E 4 33 5 0000E 7 74 5 6000E 6 115 7 7500E 5 155 4 5000E 4 34 4 5000E 7 75 5 4000E 6 116 7 5000E 5 156 4 0000E 4 35 4 0000E 7 76 5 2000E 6 117 7 2500E 5 157 3 5000E 4 continued on next page CCFE Page 179 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual B EAF LIBRARY DATA continued from previous page CCFE 162 group structure grp energy eV grp energy eV grp energy eV grp energy eV 36 3 5000E 7 77 5 0000E 6 118 7 0000E4 5 158 3 0000E 4 37 3 0000E 7 78 4 8000E 6 119 6 7500E 5 159 2 5000E 4 38 2 8000E 7 79 4 6000E 6 120 6 5000E 5 160 2 0000E 4 39 2 6000E 7 80 4 4000E 6 121 6 2500E 5 161 1 5000E4 4 40 2 4000E 7 81 4 2000E 6 122 6 0000E 5 162 1 0000E 4 41 2 2000E 7 82 4 0000E 6 Table 24 Energy group boundaries for the LLNL 616 group structure LLNL 616 group structure erp energy eV grp energy eV grp energy eV grp energy eV 1 2 0000E 7 155 1 7378E 4 309 1 4454E 1 463 1 2023E 2 2 1 9953E 7 156 1 6596E 4 310 1 3804E 1 464 1 1482E 2 3 1 9055E 7 157 1 5849E 4 311 1
115. Equation 11 We use the following definitions path A path is a chain of different nuclides connecting the source nuclide to the target nuclide loop A loop is a closed chain of different nuclides connecting a nuclide to itself Loops formed by the cyclic permutation of the nuclides in the loop are considered to be the same loop pathway A pathway is the combination of a single path with zero or more loops These are illustrated in Figure 9 The full directed graph has one vertex per nuclide and one edge for each off diagonal term in the rate equations The EAF 2010 data have 2233 vertices nuclides and CCFE Page 158 of 200 A 12 Pathways CCFE R 11 11 Issue 6 FISPACT II User Manual 1 2 3 4 a e 0 90 path 2 3 5 2 b e 0 0 0 loop 1 4 c e pathway 7 Figure 9 a A path is a linear chain of nuclides connected by edges b a loop is a cyclic chain of nuclides and c a pathway is the combination of a single path with zero or more loops approximately 120000 edges non zero elements in A If fission can be omitted this reduces to about 42000 and for a cooling period this drops to less than 5000 c f Section A 1 The brute force approach to finding paths by examining all the paths to descendants of a given source nuclide for these numbers of edges rapidly becomes impracticable because of the combinatorial explosion of the number of alternatives to be examined as path lengths increase FIS
116. Examples of the use of the LOOKAHEAD keyword may be found in fispQA Tst 709 test127 i and fispQA2010 Tst 211 test79 i 5 2 27 MASS totm indx2 sym i zp i i 1 indx2 This keyword allows the input of the total mass totm kg and the number indz2 of elements in the material to be irradiated For each element the chemical symbol sym i e g Fe and the percentage by weight zp i are then read This keyword enables elements to be input with the number of atoms of each isotope calculated by FisPACT II using natural abundance data that are stored internally If an element whose natural abundances are not known is selected then FISPACT II will issue a fatal error message Computations for these cases must use the FUEL keyword The MASS keyword is the recommended method of inputting materials unless special isotopic compositions are required An example of the use of this keyword is In this case the composition of a stainless steel ignoring impurities and minor elements is specified 1kg of the steel containing the seven listed elements is to be irradiated Note that FUEL and MASS must not both be used in the same case in the input file Note it is not essential that the total of all elements is exactly 10096 however if the total was say 80 and 1 kg was specified for totm then only 800 g of material would be considered in the calculation It is recommended that the user ensures that the total percentage of all elements equals
117. F libraries ngr 1 ngr is used to retain backward compatibility jb is specified in the same manner as ja above Note that if a fission reaction is required then jb should be either Fission or 0 ALAM thalf units thalf is the new half life of the nuclide and units specifies the time unit SECS 1 seconds MINS 2 minutes HOURS 3 hours DAYS 4 days YEARS 5 years The units are specified either by name SECS MINS etc or by number 1 2 etc ADCROSS jb errfcz jb is the daughter of the reaction and errfcz is the new error factor for the cross section ADLAM dthalf dthalf is the new error factor for the half life Examples of the uses of the OVER keyword are OVER BE9 ACROSS HE6 1 05490E 2 Here the 1 group cross section for the reaction Be n a 8He is given the value 10 549 mb for all subsequent calculations in the run OVER C14 ALAM 3000 0 5 CCFE Page 74 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual Here the half life of C is given the value 3000 0 years for all subsequent calculations in the run OVER C14 ADCROSS C13 1 10 Here the error factor for the C n 2n C reaction is set to 1 10 for all subsequent calculations in the run Note that the arrayx and collapx files are not altered so that in subsequent runs the cross section half life or error factor will revert to its original value Note that the OVER keyword must occur after the GETXS and GETDECAY keywords t
118. ISPACT II requires connection to several data libraries before it can be used to calcu late inventories While any libraries in the correct format could be used the code has been designed to use the European Activation File and this library is the recommended source of cross section data The following libraries are required e Cross section data for projectile induced reactions where the projectile defaults to neutron or may be one of the other four options given using the PRO JECTILE keyword and the energy group structure is one of those listed in Appendix e Uncertainty data for neutron induced reactions e Projectile spectrum data These may be one of the standard fluxes files pro vided with the library see Appendix B 3 or may be a user generated one see page 50 e Decay data e Fission yield data for projectile induced reactions where the projectile may be neutron deuteron or proton e Biological hazard data e Legal transport data e Clearance data e Gamma absorption data The libraries are described in more detail below It is a user s choice to select from the 2003 2005 2007 or 2010 library versions B 1 Cross section Group Structure There are nine standard group structures are used for the European Activation File and two standard group structures in the ENDF format data in all these structures can be read into FISPACT II Table 20 lists the group structures for the five original cases with upper
119. If the cross section data collapse and decay data condense are undertaken in separate runs of FISPACT II it is necessary to include the EAFVERSION 8 command in both runs to obtain both new cross section data and new decay data This allows CCFE Page 38 of 200 oma 10 Ct RB QO Ne 4 7 Developing New input Files CCFE R 11 11 Issue 6 FISPACT II User Manual the flexibility of using old cross section data with new decay data and vice versa If the GETXS 1 and GETDECAY 1 commands appear in the same input file then EAFVERSION should appear at most once and the cross section data and decay data are either both ENDF format or both EAF format The file connection diagram for a collapse run using the ENDF format data is obtained by replacing the crossec and crossunc streams in Figure I by the single input stream xs endf Similarly the ENDF format file connection diagram for a condense run is obtained from Figure 2 by respectively replacing the decay and fissyld streams by dk endf and fy endf and by removing the asscfy stream Fission yield associations are not used with the new TENDL 2011 TENDL 2012 and TENDL 2013 data 4 7 Developing New input Files The above examples give an initial guide on how to use FISPACT II to undertake inventory calculations Examining the sets of test cases in directories fispQA2010 and fispQA see Section 6 together with the definitions and usage examples of the keywords see Section 5 provide further guida
120. In this example an irradiation phase of 2 5 years is specified TIME 2 5 YEARS ATOMS The cooling phase is started by the ZERO keyword This resets the time origin causes the pathways analysis to be performed over all the steps preceding it and initiates the saving of data for the graphs NOTE there must be at least two cooling steps for the graph output to be plotted NOTE there must be no more than one ZERO keyword in an input file but there may be irradiation as well as cooling steps after the ZERO keyword MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS YEARS ATOMS To run the inventory calculation type fispact inventory This should cause the output shown below to appear at the terminal window The set tings keywords are simply echoed but the action keywords in this example ATOMS ZERO and END list the actions they initiate The final line gives cpu timing and a summary of the number of errors or warnings issued MONITOR 1 GETXS 0 GETDECAY 0 FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 load cross sections load decay data collapse fission yields run reset cross section MASS 1 0 1 CCFE Page 35 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual 4 GETTING STARTED TI 100 0 FLUX 4 27701E 14 MIND 1 E5 GRAPH 2 2 1 1 4 UNCERTAINTY 2 ATOMS load initial values run output inventory HAZARDS load hazards data HALF ATWO load a2 data TIME 2 5 fill rate equation matrix for cooling fill rat
121. M 06441E 05 Sv kg CCFE Page 104 of 200 7 1 The Inventory Run output File CCFE R 11 11 Issue 6 FisPACT II User Manual The TOTAL ACTIVITY FOR ALL MATERIALS item gives total activity in Bq and the TOTAL ACTIVITY EXCLUDING TRITIUM is the total with tritium activity excluded The HEAT PRODUCTION items are the sums over all materials of the respective a P and powers the total of these three powers and the total with the contribution of tritium decay excluded The NUMBER OF FISSIONS is a count of the change of the number of nuclides that may undergo fission from the number in the the initial inventory These nuclides are identified as those with the MT 18 reaction on their list of reactions c f Table on page 140 BURN UP OF ACTINIDES gives the percentage of the initial number of fissionable nuclides that have been burnt up Note that NUMBER OF FISSIONS may become negative if for example there are no nu clides with MT 18 initially but ones are created by irradiation of the initial inventory All nuclides with MT 18 reactions are counted even if their reactions are excluded be cause USEFISSION is absent or reactions are excluded by the FISYIELD keyword or reactions are excluded because their fission yield data are not available The remaining items in the summary list depend upon the use of the ATWO CLEAR and HAZARDS keywords and on whether the DENSITY keyword was used If the ATWO keyword is used
122. NDL data this make take values 709 or 162 default 709 CCFE Page 42 of 200 4 8 Compressed ENDF Library Files CCFE R 11 11 Issue 6 FISPACT II User Manual 4 the save type This takes a value 0 5 that specifies what data are to be saved default 1 5 the name of files file default files The save types are cross section only cross section and variance cross section variance and covariance resonances and cross section resonances cross section and variance resonances cross section variance and covariance orWNrF Any other value defaults to 1 For most applications cross sections only or cross sections and variance neutron irradiation are sufficient Covariance data are only available for neutron irradiation and are only needed if the COVARIANCE keyword is being used The resonance data are only used if the SSFGEOMETRY keyword is to be used An example of the use of compress_xs_endf that creates a binary compressed library tal2013 n bin containing 709 group cross sections variance and covariance data for neutron irradiating flux using the T ENDL2013 library is as follows compress xs endf tal2013 n n 709 2 The fifth argument is not present and so will take its default value files This file must contain the mappings for ind_nuc and for xs_endf For example index of nuclides to be included ind nuc ENDFdata TENDL2013data tendl13 decayl2 index Library cross section data xs_endf ENDFda
123. PACT II User Manual A 5 2 ENDF data The fission yield data like the covariance data are on coarser energy grids than the flux and cross sections To collapse the fission yield the weights are calculated using N N Wy gt gt S i Y hi 55 i 1 i 1 where there are k 1 K fission yield energy groups The yields are collapsed using K Y 2 Y WAY 56 k 1 The variance of the collapsed fission yield is given by K var OE 57 k 1 where Pj are the tabulated 1c errors in the ENDF file The fractional uncertainty is A var Y In the present version of the FISPACT II the fission yield uncertainty is not used A 6 Gamma Activation The set of reactions allowed for gamma activation is identical to the set of 90 reactions for neutron activation The table for these reactions can be obtained by replacing the projectile n by y and decreasing all the values of AA by 1 in Table 12 A 7 Proton Activation The set of reactions allowed for proton activation is identical to the set of 90 reactions for neutron activation The table for these reactions can be obtained by replacing the projectile n by p and increasing all the values of AZ by 1 in Table A 8 Deuteron Activation The set of reactions allowed for deuteron activation is identical to the set of 90 reactions for neutron activation The table for these reactions can be obtained by replacing the projectile n by d and increasing all the values of AZ by 1 and a
124. PACT II uses a much faster technique based on iteration on a single visit tree where the tree is pruned using edge weighting to reduce the number of graph edges to be retained in the search for significant pathways 41 The weights are computed using coefficients of the rate equation matrix A and are used to eliminate paths and loops as follows e Each edge on a path has weight lt 1 If the product of the weights along a path falls below the path threshold then the path is discarded e Each edge on a loop has weight lt 1 The weight of the loop Wi is the product of the edge weights around the loop If W 1 W falls below the loop threshold then the loop is discarded The retained paths are significant paths and the retained loops are significant loops Significant paths and significant loops are combined to give pathways Loops are added to paths according to the following criteria 1 the loop has one or more nuclides in common with the path 2 when the loop is added to the path it does not create a second path to the target this is to avoid double counting of paths CCFE Page 159 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual 3 when the loop is added to the path the fractional increase in the target inventory due to the path is greater than the loop floor The resulting pathway is discarded as not significant if the fraction of the target in ventory due to the pathway i e the path a
125. al inventory and the USEFISSION keyword has not been used then warning messages are written to both the output and runlog files 5 2 62 WALL wall This keyword allows the input of the total neutron first wall loading wall in units of MW m for a fusion device This is converted to a flux value by using data read from the neutron spectrum file The neutron spectrum file 1uxes contains a value of the first wall loading e g 4 15 MW m7 The energy integrated flux e g 1 80 x 10 nem s which is approximated by the sum of neutrons in all the groups is calculated and equated to the wall loading during library processing Note that it is the user s responsibility to ensure that this wall loading is correct when the spectrum file is constructed If a wall loading of 2 0 MW m was input then a flux value of 2 0 4 15 x 1 80 x 10 ncm s would be used in the calculations WALL is a convenient alternative to using FLUX for the irradiation of first wall materials but great care must be exercised if it used for irradiations with other than first wall spectra In these cases the flux specified for the region must be that which would be present if the first wall loading shown in the file was present on the first wall It is recommended that FLUX is always used in preference to WALL unless the user has a run that makes its use essential It should be noted that the wall loading describes the power that impinges on the first wall not what is act
126. amma doses see the SPEK keyword and Appendix A 10 3 For the EAF data fission yield data for actinides where data are available are read from the EAF fission yield library and for the remainder a neighbouring fission yield is used as specified by the fission association file Appendix B 5 More extensive fission yield data for more nuclides are available in the TENDL 2013 data in the ENDF library and so the fission association and surrogate daughter schemes are not used for these new data The twenty four decay types ten spectrum types and ninety reaction types recognised by FisPACT II when reading the EAF and ENDF library data are summarised in Ta bles 10 11 and 12 in Appendix A The additional seven MT values for total cross sections for gas production eight for kerma and four for dpa that may appear in the TENDL 2013 data sets are listed in Table 13 Other MT values that are recognised by the code but are silently ignored are listed in Table The mapping of reaction MT numbers to the CALENDF group MT numbers implemented in FISPACT II are summarised in Table Printed summaries of the related library data may be output using the PRINTLIB CCFE Page 18 of 200 2 2 Inventory Calculation CCFE R 11 11 Issue 6 FisPACT II User Manual keyword see page 77 2 2 Inventory Calculation Library data preparation provides the cross sections and decay constants needed to construct the coefficients for the rate equations The rate equatio
127. ansport index eaf a2 Transport of radioactive material from place to place is governed by regulations set up by the IAEA Reference gives details of A2 values for certain radionuclides Using these values it is possible to work out how much of a particular mixture of radioactive materials can be packed into a type of container and safely transported Data from this reference for the nuclides listed are transferred to eaf a2 with the default prescription given in reference 25 used for all radionuclides not explicitly listed References 13 25 document the eaf a2 libraries FISPACT II can use these data to show the A2 limit for individual nuclides and the effective A2 value for the irradiated material B 6 3 Clearance index eaf clear Disposal of radioactive material in special repositories is expensive Regulations exist which determine activity levels for nuclides such that materials can be cleared or disposed of as if they are not radioactive Clearance data are being investigated by the IAEA and recommendations are available Reference gives details of suggested clearance values for certain radionuclides while an earlier report reference 55 gives a formula that allows values for other nuclides to be calculated Data from these references for the nuclides listed are transferred to eaf clear with the default pre scription used for all radionuclides not explicitly listed References document the eaf clear libraries FISPACT II can us
128. ases provide a check on whether your installation is working correctly In each of the test directories is a script fisprun that runs all the test cases in the directory The results generated by executing this script should match those in the testresults directory apart from run timestamp minor roundoff discrepancies and for sensitivity calculations only differences arising from different random numbers So if you see error messages from your runs check against the reference data to see if that is what is expected Test input cases for standard collapse condense and inventory runs including pathways and uncertainties are covered in the Getting Started section page 26 Examples of different ways of using the code are illustrated in the fispoA2010 and fispQA directo ries For example in fispQA2010 see test37 for sensitivity calculation test18 for multi pulse irradiation test65 for time dependent collapsed cross sections test97 8 for reduced nuclide set calculations test for ROUTES investigation test for PATH calculation test for IRON calculation test10 for OVER calculation The subdirectories of directory fispQA illustrates features added in Version 2 20 Direc tories Tst_162alpha Tst 162deut Tst 162gamm and Tst 162prot respectively contain CCFE Page 98 of 200 CCFE R 11 11 Issue 6 FisPACT II User Manual illustrations of calculations using the TENDL 2013 libraries for a d y and p projec tiles The
129. at survive step 2 and extract from this paths and loops that have one or more nuclides on one of the paths to the targets A branch of the tree is terminated when e a loop is found e the weight of the path is below path floor e the length of the path reaches max depth Building of the tree terminates when there are no more parent nuclides in the digraph queue CCFE Page 160 of 200 A 13 Uncertainty Estimates CCFE R 11 11 Issue 6 FisPACT II User Manual Step 4 prune loops retaining only those loops where the weight of the loop is above loop floor and the combination of loops and path give a single candidate path way whose weight bound is above path floor Step 5 integrate the rate equations for candidate pathways to get the actual weight of the path storing pathways above the path 1oor threshold in decreasing weight order and discarding those pathways below the threshold A 13 Uncertainty Estimates The pathways analysis is used to identify the pathways from the initial inventory nuclides to the target dominant nuclides at the end of the irradiation phase together with the number of atoms created at target nuclide t due to the reaction and decay chain along path p to that nuclide These together with uncertainties in the reaction cross sections and decay half lives associated with the edges of the pathways are used in FIsPACT II to provide estimates of the uncertainties Given a set of target nuclides S then the uncer
130. atures 3 8 Keyword Changes Getting Started 4 1 Introduction 4 2 Cross section Collapse 4 8 Decay Data Condense 4 4 Library Summary Printing 4 5 Inventory Calculation 4 06 ENDF format Library Data 4 7 Developing New input Files 4 8 Compressed ENDF Library Files 4 9 Reduced Nuclide Index Control File Keywords 5 1 Library Data Preparation 5 1 1 CLOBBER 5 1 2 COVARIANCE 5 13 EAFVERSION 5 1 4 FISPACT 5 L5 JEFU LLXS esca alase a a ee 5 16 GETDECAY 5l GEUXS 2444 246 ia geyri 5 18 GRPCONVERT 5 19 LOGLEVEL 5 1 10 MONITOR 5 1 11 NOERROR 5 1 12 NOFISS 5 1 13 NOHEADER 5 1 14 PROBTABLE 5 1 15 PROJECTILE 5 1 16 SAVELINES CCFE Page 7 of 200 CCFE R 11 11 Issue 6 CONTENTS FISPACT II User Manual Dl lf SPER oeo acia a amp ar te o que aa oque pe ke A 5 1 18 SSFCHOOSE 2 2 4922zn RR Rs 5 11 19 SSFDILUTION a uu ic deno ERO A OR ee ee e o 5 1 20 SORPPUEL e x ga ico Rees wks a Ge D ERES 5 1 21 SSEGEOMETRY 000 ee ee 56 51 22 SOPMADSS Lue dee er td RUE E ae pou x B cR ee 8 5 2 Imtial Conditions lt lt lt 6x Robo kw an Xem ae ox 5 21 ATOMS 4 2k hk e E REO ORE E
131. ause the pathways to the late time dominant nuclides are not included in the uncertainty calculation If the loss of accuracy is due to only a few late time dominant nuclides then the LOOKAHEAD Section 5 2 26 keyword provides a simple means of including all the late time dominant nuclides In some cases particularly when there are actinides in the material then LOOKAHEAD leads to a slow computation because too many nuclides get included in the pathways calculation The PATHRE SET Section keyword provides an alternative in these cases This keyword causes the pathways calculation to be redone for the dominant nuclides at the time interval preceding the keyword and by reducing fopxr and increasing the number of occurrences of PATHRESET combinatorial growth can be avoided whilst retaining important pathways at each time step Uncertainty estimates that are printed for both the old and new set of pathways at the points where PATHRESET are used indicate whether the reset is needed to achieve convergence of the error estimate A description and examples of the uncertainty and pathways output generated by using this keyword may be found in Sections 7 1 10 7 1 11 and V 1 12 Appendices and outline the methods of calculation 5 2 60 UNCTYPE iuncty 1 This keyword allows the user to specify the type of uncertainty contributions to include when calculating the uncertainties of the radiological quantities If iuncty is set to 1 or if the keyword i
132. ay also appear in the inventory calculation section in conjunction with further GETXS keywords 5 1 21 SSFGEOMETRY type length1 lt length2 gt This keyword introduces the use of the universal sigmoid curve model of self shielding 21 22 23 to account approximately for the reduction of the neutron flux by cross section resonances The first integer parameter type defines the type of target geometry and the following one or two real parameters specify the size of the target in units of cm Permitted values of type are 1 4 with interpretations as follows CCFE Page 56 of 200 5 1 Library Data Preparation CCFE R 11 11 Issue 6 FISPACT II User Manual type Target shape length1 length2 1 foil thickness not used 2 wire radius not used 3 sphere radius not used 4 cylinder radius height where foil targets are taken to be of infinite transverse extent and wires are taken to be infinitely long The self shielding factors are calculated using the resolved resonances of the nuclides specified with the SSFFUEL keyword or indirectly using the SSFMASS keyword In the latter case the natural abundance data stored internally in FISPACT II are used to calculate the numbers of atoms of the individual nuclides For foil type 1 or wire type 2 targets the numbers of atoms specified are interpreted as being per unit area or length respectively Note that SSFGEOMETRY and PROBTABLE must not both be used in a particu
133. be included as often as required This causes the late time dominant nuclides to be included in the uncertainty calculations There are three values for the showpathways argument 1 display pathways for a target nuclide for which pathways have not been displayed at earlier times O do not display pathways but use the pathways in uncertainty estimates 1 display pathways for all dominant nuclides at each pathways reset An example of the use of this keyword is TIME 6 8E10 ATOMS PATHRESET 0 TIME 2 2E11 ATOMS For further discussion on this see the UNCERTAINTY keyword on page Examples of its use may be found in fispQA Tst 709 test128 i and fispQA2010 Tst 211 test80 i 5 3 18 PROBTABLE multxs 0 usepar 1 The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 14 Its use here is in conjunction with a subsequent GETXS keyword CCFE Page 92 of 200 5 3 Inventory Calculation Phase CCFE R 11 11 Issue 6 FisPACT II User Manual 5 3 19 PULSE npulse This keyword is used to start the loop construct in the input file npulse is the number of times that the keywords between PULSE and ENDPULSE are repeated Using FISPACT II it is possible to nest this pair of keywords to an arbitrary depth and there is now no limit on npulse for any loop This facility is included so that a series of identical pulses off time and on time can be represented easily in the inp
134. broadr cross check mato unresr cross check recent ne p y lecritp thermr A e heatr sixpack gasp activate purr merger Ka groupr me 4 s BE cietni P groupie 4 K Lg Figure 13 A schematic of the processing sequence using NJOY PREPRO and CAL ENDF to transfer regularly to technology the feedbacks of extensive validation verification and benchmark activities from one release to the next T ENDL 2013 is the fifth gen eration of such a library and as such has benefited from the previous releases since TENDL 2008 but also from the EAF 2007 and EAF 2010 V amp V processes 61 The cross section data are provided adequately in two universal group structures a CCFE 709 scheme for the neutron induced cross sections and a CCFE 162 scheme for the non resonant p d and y induced cross sections The data format used is fully compliant with the ENDF 6 manual specification handled on an isotopic basis and so allows many existing utility codes further to manipulate visualise or check any aspects of the pre processed files The data files are produced using a complex but robust complementary sequence of modules of the processing codes NJOY12 021 and PREPRO 2013 62 During the processing outputs from verification and validation steps are regularly taken in order to establish the validity of all computed derived data To be able to account for Doppler broadening effects the processed files are give
135. cattering 2 101 absorption no outgoing neutron 102 103 107 18 fission total 18 4 inelastic scattering emitting one neutron A11 15 multiple neutron production excluding fission 5 16 17 37 The data provided by CALENDF are cross section and probability values depending on four parameters c x on o p g x n 25 P r on P p g 2 n 26 where p parent nuclide number g energy group number x macro partial or total index n quadrature index In the expressions below we suppress the explicit display of dependence of cross section on the parent nuclide p and energy group g except in the formulae for dilution The CCFE Page 144 of 200 A 4 Neutron Activation CCFE R 11 11 Issue 6 FisPACT II User Manual infinite dilution d oc cross section for a given parent energy group and component is a z d EE En f c E dE Y P z n o z n 27 max man min n 1 When a nuclide is a part of a homogenous mixture of nuclides then the effective cross sections in the resonance regions are reduced and are parameterised using the dilution cross section d 30 En Pla n o x n or n d o x d 28 nai P z n o n d where the total cross section is given by the sum of the macro partials X en Y o a n 29 gl The total cross section for nuclide p in energy group g at dilution d is given by X o d 5 c x dp 30 a 1 The probability table data from CALENDF are
136. collapsed library collapx file specified in the files file If librs is 1 then the second parameter ebins gives the number of energy bins to be used in collapsing the cross section data from the EAF or ENDF library files and fluxes or arb flux files specified in the files file If bas is 1 then the ENDF data are read from the compressed binary version of the ENDF data stored in the file specified by xs endfb in the files file The value libzs 1 is not valid for EAF libraries For information on the preparation of the compressed binary ENDF data files see Section 4 8 on page 42 The GETXS keyword may also be used in the initial conditions and inventory calcu lation phases for handling time dependent projectile spectra and temperature changes in cross sections see pages and B8 The number of energy groups ebins must be consistent with the number of groups in the supplied library file The permitted numbers of groups for cross section data are currently EAF 2010 66 69 100 172 175 211 315 351 616 ENDF 162 616 709 Each set of energy dependent cross sections is then combined in a weighted sum with the supplied projectile spectrum to produce a one group effective cross section library which is used directly in subsequent runs Note that if no uncertainty data are supplied in the library as for the deuteron and proton induced reactions then the keyword NOERROR must be used An example of the use of this keyword is GETXS 1 21
137. corresponding new style files that can be found in the later fispQA directories The fispQA test set provides tests that use the new ENDF format libraries introduced in Version 2 for cross sections fission yields and decay and the new CALENDF output files for the probability tables used in calculating self shielding Note that some of the test cases issue warnings and some terminate with fatal error messages The purpose of these test cases is to illustrate the CCFE Page 97 of 200 CCFE R 11 11 Issue 6 6 TEST CASES FISPACT 1 User Manual errors that are issued if obsolescent keywords are used or if keywords are used incorrectly in the input file To find examples of the use of a particular keyword go to the fispQA2010 or fispQA directory and use grep to search for the keyword for example UNCERTAINTY fispQA2010 grep UNCER Tst i Tst 066 test112 i UNCERT 2 Tst 066 test113 i UNCERT 0 Tst 100 test4 i UNCERT 1 1 0E 4 1 0E 2 20 3 Tst 100 test8 i UNCERT 3 Tst 100 test9 i UNCERT 3 Tst 100 test9 i UNCERT 0 Tst 172v test32 i UNCERT 2 One can then go to the relevant test directory e g Tst 100 and view the input files files and say test8 i to see the context of the use of the keyword and then view what the result of running that case is by looking in the Tst_100 testresults directory for the output files test8 10g test8 out etc In addition to providing a useful guide to using FISPACT II these test c
138. ction for one of the kerma cross sections listed in Table Specific values of this energy per kilogram and per cm are obtained by scaling the total kerma using the initial mass and density Gas production rates in s are given by Nn GAS RATE 4 Nia 6 i 1 where 37 is the collapsed total gas production cross section in cm A list of the total gas production cross sections recognised by the code is given in Table 13 on page 143 If there is gas production from decays then there will be corresponding rates e g GAS RATE a decay printed If any of the kerma dpa or gas appm rates are zero then their production rates are not printed The final part of the summary output table is the gas atoms parts per million for the five secondary gas nuclides APPM OF H APPM OF H APPM OF H APPM OF He 644 46 26 987 0 15302E 01 0 44706E 03 204 59 APPM OF He CCFE Page 107 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual 7 1 5 Elemental inventory The composition of material by element is the next table displayed The column headings for this are number of atoms of the element number of gram atoms number of grams 8 power output Curie MeV and kW y power output Curie MeV and kW and o power output Curie MeV and kW COMPOSITION OF MATERIAL BY ELEMENT ALPHA ATOMS GRAM ATOMS CURIES MeV CURIES MeV kW CURIES MeV 4551E421 1 4040E 02 4718E 02 5 2945E 05 1385E 10 0 0000E
139. d be set to 3 and for an alpha library nproj should be set to 4 A neutron library uses the default value of 1 An example of the use of this keyword in the collapse of a deuteron library is MONITOR 1 PROJ 2 NOERROR GETXS 1 211 FISPACT COLLAPSE EAF_20070 WITH IFMIF END END OF RUN 5 1 16 SAVELINES This keyword causes the spectral line energies and intensities to be stored when the decay library data are being condensed Note that the spectral lines output option chosen with the PRINTLIB 5 command will produce spectral line output only if the SAVELINES command is used as de scribed above 5 1 17 SPEK This keyword causes the calculation of an approximate y spectrum for nuclides in the decay library which have no spectral data These nuclides are flagged by an amp in the inventory output and in the output of library data produced in a run with the keyword PRINTLIB 5 1 18 SSFCHOOSE ncho 0 nprint 0 sym j j 1 ncho This keyword is used to specify the nuclides for which the self shielding factors are computed using the probability table data see Appendix on page 144 ncho gives the number of element or nuclide names that follow the keyword The symbols sym may either be element names e g Ti or nuclide names e g W182 If an element name is given then all the naturally occurring isotopes of that element are included in the list of nuclides to which the self shielding correction is
140. data are identified For runs where the cross section data change further information on EAF source and flux files is displayed at the end of the output file There are no platform specific messages as FISPACT II is written in standard conforming Fortran and the same source is used for Unix Linux Mac OS and Windows versions Occurrences of the ATOMS keyword in the input file cause the output at the end of the step of 1 table keys first ATOMS only 2 the time line 3 iron information if the IRON keyword used see fispQA2010 test8 for an ex ample 4 the inventory comprising CCFE Page 100 of 200 7 1 The Inventory Run output File CCFE R 11 11 Issue 6 FISPACT II User Manual a the heading line b a line for each nuclide with non negligible inventory see MIND keyword c nuclide table totals d inventory summary 5 inventory by element if NOCOMP is not used 6 gamma spectra 7 gamma dose totals 8 dominant nuclides if NOSORT is not used 9 Bremsstrahlung corrections if the BREMSSTRAHLUNG keyword is used see fispQA2010 test4 for an example 7 1 2 Table key Prior to the first inventory tables output initiated by the ATOMS keyword the fol lowing key is printed NB IN FOLLOWING TABLES MEANS CONVERGENCE NOT REACHED FOR NUCLIDE amp MEANS GAMMA SPECTRUM IS APPROXIMATELY CALCULATED MEANS NUCLIDE IS STABLE MEANS NUCLIDE WAS PRESENT BEFORE IRRADIATION This key list
141. diation calculations including those where the flux spectrum changes from pulse to pulse It can now read ENDF style data libraries in addition to the EAF libraries and the present version uses the latest TALYS based TENDL 2011 TENDL 2012 and TENDL 2013 evaluated nuclear data libraries together with probability table data from CALENDF for including self shielding in the calculations These libraries allow additional projectiles and nuclides to be included and make possible additional kerma dpa and appm diagnostics This document is its User Manual It first outlines what calculations the code performs and how the code differs from FISPACT 2007 It has a getting started section to provide a basic introduction to new users It explains the use and provides examples of all the keywords used in the input file to specify a FISPACT II run and describes how all the data files are connected It introduces the test cases provided with the code and gives a guide to interpreting the physical output and logging output from the code It also introduces subsidiary programs for printing output and for compressing the ENDF cross section libraries Three appendices are provided the first outlines the physical and mathematical models implemented in the code The second summarises the EAF nuclear data used by the code giving background information on the data files and examples of neutron spectra suitable for various applications The final appendix describ
142. e 12 j SHI 1e 11 1 al 1 l 1 sl 1 sl 1 Load 1 i xu 1 1e 06 1e 05 0 0001 0 001 0 01 0 1 1 10 file name test81 gra Time after irradiation years run timestamp 13 21 51 18 February 2014 Figure 5 Graphical output produced using the gnuplot visualisation package The third parameter guncrt allows the user to specify if uncertainty data should be 1 or should not be 0 written to the graph file If the uncertainty data are written then the plotting routines can display the uncertainties on all five types of plots The axes are scaled automatically in the gnuplot plt file The minimum time is set to the start of the logarithmic decade in which the first cooling step is displayed The value of the radiological quantity at the start of the cooling time t 0 is plotted on the ordinate of this graph The graph command will fail if there is not at least one cooling step If different display options are required then the user may edit the p1t files to match their preferences An example of the use of this keyword is CCFE Page 66 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual GRAPH 3 0 1 124 In this case data on activity dose rate and ingestion dose are written to a file in standard format with uncertainty data included From this file three graphs can subsequently be plotted 5 2 19 GROUP igamgp 0 This keyword specifies the binning of the discrete photon spectral line
143. e FISPACT keyword enters a labelling line and marks the end of the library preparation section of the input control file The PRINTLIB 4 line causes a summary of the collapsed cross sections to be printed The END keyword marks the end of the input To perform the collapse type fispact collapse You should see the output collapse cpu time 0 472 secs No errors warnings where the cpu time will be changed to that appropriate to your computer You will find that new files collapse log collapse out and collapsed cross section data have appeared in the present working directory The first two are text files and the third is a binary file of data that can be used as input for later runs The log file contains a summary of the files linked to the program monitoring echo of the keywords and actions a list of the files used and a summary of the cpu times The output file contains the header summary the collapsed cross sections and uncer tainties run identification information and a summary of the files used If you type fispact collapse again you will get output CCFE Page 30 of 200 4 2 Cross section Collapse CCFE R 11 11 Issue 6 FisPACT II User Manual Inspection of collapse log in this case will reveal the error message file unit name output file name collapse out Fatal files m files open 9 cannot open new file for writing The run failed because the repeat run tried to write output to an existin
144. e SubletQccfe ac uk This work was funded by the RCUK Energy Programme under grant EP 1501045 CCFE is the fusion research arm of the United Kingdom Atomic Energy Authority Neither the authors nor the United Kingdom Atomic Energy Authority accept respon sibility for consequences arising from any errors either in the present documentation or the FISPACT II code or for reliance upon the information contained in the data or its completeness or accuracy CCFE is certified to ISO 9001 and ISO 14001 Date of Issue 4 August 2014 Issue number 6 Authorization Name Signature Position Prepared by Dr J W Eastwood T 5 f Lo e ISPACT II Developer Approved by Dr J G Morgan py FisPACT II Developer Released by Dr J C C Sublet Nuclear Data Manager CCFE Page 4 of 200 CCFE R 11 11 Issue 6 FisPACT II User Manual Executive Summary FISPACT II is an inventory code capable of performing modelling of activation trans mutations and burn up induced by neutron proton alpha deuteron or gamma parti cles incident on matter It is a completely new inventory code initially designed to be a functional replacement for FISPACT 2007 but now extended to have substantially more capability FISPACT II is written in object fortran and has full dynamic memory allocation It has improved algorithms for the ode solver pathways uncertainty and sensitivity calculations All these can be used in multi pulse irra
145. e effective self shielding factor for the collapsed cross section ane p y ssf p y SHBG S OD 41 A 4 4 Self shielding of resonant channels using the universal curve model Starting from Release 2 10 FISPACT II provides a second method of accounting for self shielding in thick targets with a variety of geometries This can be used as an alternative to the probability table method described in the previous section it is not possible to use both descriptions of self shielding simultaneously In a series of papers 21 22 23 the authors Martinho Goncalves and Salgado de scribed a universal sigmoid curve model of self shielding to account for the reduction of the neutron flux by cross section resonances in the context of neutron activation analysis They based their development on earlier experimental and theoretical work by Baumann 34 The Martinho et al model initially described the effect of a single resonance peak in a pure target consisting of a single nuclide The self shielding factor Gres is approx imated as a simple function of a single dimensionless length parameter that depends on the physical size and shape of the target as well as the peak cross section at the resonance and the resonance widths for elastic scattering and radiative capture CCFE Page 147 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual The final form of the model accommodates a group of isolated resonances of a p
146. e equation matrix for irradiation start pathstep recording initialise dominant analysis YEARS ATOMS run add rateeq for pathways run irradiation init run irradiation step run add pathstep run output inventory FLUX 0 ZERO TIME 1 MINS ATOMS run pathways initialisation run pathways uncertainty run cooling step run add pathstep run output inventory run pathways uncertainty TIME 1 HOURS ATOMS run cooling step run add pathstep run output inventory run pathways uncertainty TIME 1 DAYS ATOMS run cooling step run add pathstep run output inventory run pathways uncertainty TIME 7 DAYS ATOMS run cooling step CCFE Page 36 of 200 4 5 nventory Calculation CCFE R 11 11 Issue 6 FISPACT II User Manual run add pathstep run output inventory run pathways uncertainty TIME 1 YEARS ATOMS run cooling step run add pathstep run output inventory run pathways uncertainty END END run output summary run closedown deallocate and closedown inventory cpu time 0 379 secs No errors warnings The files inventory log inventory out inventory gra and inventory plt are also created by this run To convert the inventory gra file to a postscript output file type gnuplot inventory plt Figure 3 shows the first page of the resulting plots IRRADIATION OF TI EEF FW 1 0 MW M2 1e 15 T T T T i T Activity Bq kg S Uncertainty value t half for nuclide tert4 F 7 4 o E o a
147. e first line is the error message identifier It comprises five fields each terminated by a colon These fields are 1 error number 2 error severity 3 module 4 subprogram 9 point There are six error severities only three of which are of concern to users Fatal Close down immediately Serious Close down if 10 or more serious errors Warning Flag information to user The module subprogram and point identifiers uniquely identify the line in the code from which the error message was issued Each error message has between one and three lines of descriptive information In some cases values relevant to error messages are output in the lines preceding the error message These take the form Log name value Log projectile 2 00001 Fatal rundata m read lib keys 7 CCFE Page 117 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual NOERROR keyword needed for projectile 1 FATAL ERROR run terminated 7 3 The Printlib Run output File The printlib output consists of the six blocks of data illustrated below These are selected by the PRINTLIB keyword as described in Section 5 2 41 on page 77 7 3 1 Decay data The summary of the decay data for each nuclide is printed with thirteen nuclides listed per page For each nuclide its internal identifier number the decay constant A s71 and the half life in appropriate units for stable nuclides is printed are given follo
148. e half lives greater CCFE Page 112 of 200 7 1 The Inventory Run output File CCFE R 11 11 Issue 6 FISPACT 1 User Manual Target nuclide Sc 44 99 557 of inventory given by 8 paths path 1 20 048 Ti 46 R Sc 45 R Sc 44 S 98 16 n np 100 005 n 2n 1 84 n d path 2 12 567 Ti 46 R Sc 45 R Sc 44m b Sc 44 S 98 16 n np 100 005 n 2n 100 00 IT 1 84 n d 0 00 n n path 3 11 143 Ti 46 R Sc 45m d Sc 45 R Sc 44 S 96 62 n np 100 005 IT 100 00 n 2n 3 38 n d Shown below each edge is a list giving the percentage contributions that each reaction and decay make towards the total rate for the edge for primary products If the edge daughter is a secondary then the isomeric state of the primary product of the reaction or decay is also displayed Significant loops are displayed directly after their path with the percentages of their part of the total path percentage 7 1 12 Generic pathways All pathways differing only by an isomeric decay IT edge are regarded as the same generic pathway and are shown in the generic pathways list Individual pathways with details of the reactions and decays on each edge may be found by referring to the individual pathways The generic pathway path displays a path number the percentage of the target nuclide atoms generated along the pathway and the source to target edges Below each path is
149. e these data to show the clearance index for individual nuclides and for the irradiated material B 7 Absorption Data eaf abs The photon mass attenuation coefficient u p and the mass energy absorption coeffi cient en p for all elements with Z 1 100 have been produced using the XGAM program from the National Institute of Standards and Technology The database covers energies of photons X ray y ray and bremsstrahlung from 1 keV to 100 GeV and has been processed into a 24 group structure 1keV 20 MeV identical to the FisPACT II y group structure The present compilation follows that used in FISPACT 2007 and is an extension of the calculations of Seltzer 57 It replaces the values given in Hubble which were used in earlier FISPACT versions CCFE Page 194 of 200 CCFE R 11 11 Issue 6 FisPACT II User Manual The present data differ from the Hubble set in the following respects 1 The first 100 elements are included compared to the 40 selected elements previ ously covered 2 All edge energies are included and identified and values of u p and Len p are given just above and below each discontinuity to facilitate accurate interpolation 3 Somewhat different values for the atomic photoeffect cross section have been used for Z 2 54 4 For compounds and mixtures values for 4 p can now be obtained by simple addition i e combining values for the elements according to their proportions by weight Radiative los
150. e universal sigmoid curve approximation for self shielding SSFMASS is a new keyword used to specify indirectly the mixture of nuclides to be used in the self shielding calculations TABn now accepts arbitrary Fortran unit numbers which are ignored The unit numbers actually used are chosen internally and are reported in the log file TIME now accepts an optional keyword SECS for the time units TOLERANCE is a new keyword used to introduce convergence parameters for the LSODES solver used to compute the inventories UNCERTAINTY has a changed option to set numerical parameters relevant to the improved pathways calculation The new option is introduced by a value of 1 for the first parameter and use of the previous value of 4 for this parameter now generates a fatal error message The meanings of values 0 3 for the first parameter are unchanged USEFISSION isa new keyword that causes fission reactions specified by the FISYIELD keyword and for which yield data are available to be self consistently included in the matrix describing the inventory equations It should be used whenever actinides or other heavy elements that are transmuted to actinides are speci fied in the target material Its absence leads to a much faster calculation which remains accurate when actinides are not present or produced CCFE Page 25 of 200 CCFE R 11 11 Issue 6 4 GETTING STARTED FISPACT II User Manual 4 Getting Started 4 1 Introduction The
151. eadings b Energy a Energy and g Energy used by FISPACT 2007 are retained despite the fact that the columns contain powers kW The remaining four columns are specified by using the HAZARDS CLEAR ATWO and HALF keywords The contents of these columns are defined in Table 8 Note that the clearance index is defined as a dimensionless quantity in FISPACT II rather than as a quantity of dimension mass kg as used in FISPACT 2007 so different results will be seen for input masses different from 1kg In that table eins ei factors to convert activity of an ingested or inhaled nuclide into the dose in Sv received by an average person over 50 years These factors are tabulated in eaf hazards CCFE Page 102 of 200 7 1 The Inventory Run output File CCFE R 11 11 Issue 6 FisPACT II User Manual Table 7 Entries in columns 1 7 of the inventory output table column description value units 1 number of atoms JN 2 mass NiAri Na g 3 activity A NjAi Bq 4 B power Aj Eg iC kW 5 a power Ai ES C1 kW 6 y power A E Ci kW 7 dose rate Eq 58 or Eq 61 Sv h t Lj specific activity in Bq kg below which a material is given clearance for dis posal Values of L are tabulated in eaf clear Mot total mass of material kg A activity level for safe transport Values of Ag in TBq are tabulated in eaf_a2 C conversion factor from TBq to Bq 101 Table 8 Keywords and the entries that
152. ebruary 2009 J W Eastwood and J G Morgan The FISPACT Phase 2 Software Specification Document Technical Report CEM 081203 SD 2 Issue 2 Culham Electromag netics Ltd September 2009 J W Eastwood and J G Morgan The FISPACT II 12 Software Specification Document Technical Report CEM 100421 SD 2 Issue 4 Culham Electromag netics Ltd May 2011 J W Eastwood and J G Morgan The FISPACT II 12 Software Specification Document Technical Report CEM 120504 SD 2 Issue 5 Culham Electromag netics Ltd September 2012 V K Decyk C D Norton and B K Szymanski How to support inheritance and run time polymorphism in Fortran 90 Comput Phys Commun 115 9 17 1998 V K Decyk and H J Gardner Object oriented design patterns in Fortran 90 95 mazevl mazev2 and mazev3 Comput Phys Commun 118 611 620 2008 R A Forrest J Kopecky and J Ch Sublet The European Activation File EAF 2007 neutron induced cross section library Technical Report UKAEA FUS 535 UKAEA 2007 R A Forrest The European Activation File EAF 2007 deuteron and proton induced cross section libraries Technical Report UKAEA FUS 536 UKAEA 2007 R A Forrest The European Activation File EAF 2007 decay data library Tech nical Report UKAEA FUS 537 UKAEA 2007 J Ch Sublet L W Packer J Kopecky R A Forrest A J Koning and D A Rochman The European Activation File EAF 2010 neutron induced cross sec tion library Technical Report CCFE R 10
153. ecays The flow from the parent nuclide at vertex j to the daughter nuclide at vertex i along the directed edge ji is given by the sum of decay and reaction flows To maintain correct accounting decays or reactions leading to daughter nuclides not included in the set being considered are assigned to a fictitious sink nuclide and secondary decay products are assigned to the appropriate gas nuclides Reaction cross sections depend on the projectile energy and the source data for cross sections give values for a set of energy groups In the code an effective collapsed cross section is computed as an average cross section weighted by projectile fluxes in each energy group ae z a Eno BY s Ex 10 where al Ex is the cross section at projectile energy group k 4 Ex is the integrated projectile flux in energy group k and the sums are over all energy groups k A consequence of the modelling assumptions underlying FISPACT II is that the imposed projectile flux is not modified by the reactions and decays in the target material Then the decay rates and cross sections appearing in Equation are all independent of the nuclide numbers N and the equation can be rewritten compactly as dN rx AN 11 where the matrix A is independent of the inventory N In FIspact II the projectile flux is constant during each time interval so that A is also piecewise constant in time Furthermore the matrix A is sparse Its sparsity pattern
154. ed in a library of files CCFE Page 44 of 200 CCFE R 11 11 Issue 6 FisPACT II User Manual Table 5 Pages on which the Keywords recognised by FISPACT IT are defined page keyword page keyword page keyword ACROSS ADCROSS ADLAM AINPUT ALAM ARRAY ATOMS ATWO BREMSSTRAHLUNG CLEAR CLOBBER COLLAPSE CONV COVARIANCE 60 CULTAB DAYS 60 DENSITY DOMINANT 60 DOSE EAFVERSION END ENDPULSE ENFA ERROR FISCHOOSE FISPACT FISYIELD FLUX FUEL FULLXS GENERIC GETDECAY GETXS GRAPH GROUP GRPCONVERT HALF 69 HAZARDS HOURS 69 INDEXPATH 69 IRON LEVEL LINA LOGLEVEL 69 LOOKAHEAD LOOPS MASS MCSAMPLE MCSEED MIND MINS MONITOR NEWFILE NOCOMP NOERROR NOFISS NOHEADER NOSORT NOSTABLE NOTI NOT2 NOT3 NOT4 OVER PARTITION PATH PATHRESET PRINTLIB PROBTABLE PROJECTILE PULSE RESULT ROUTES SAVELINES SECS SENSITIVITY SEQNUMBER SEQUENTIAL 80 SORTDOMINANT 80 SPECTRUM SPEK 80 SPLIT SSFCHOOSE SSFDILUTION SSFFUEL SSFGEOMETRY SSFMASS TAB1 TAB2 TAB3 TAB4 TAPA TIME TOLERANCE UNCERTAINTY UNCTYPE 86 USEFISSION 86 WALL YEARS 96 ZERO CCFE Page 45 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual 2 initial conditions specifying physical numerical and housekeeping conditions for a calculation 3 inventory calculation phase specifying a sequence of timesteps including one or more irradiation steps optionally separated by cooling steps with further cooling steps optionally follo
155. ed later in this manual see Section 5 3 on page and Appendix A 2 3 Subsidiary Calculations The standard inventory calculation employing EAF 2010 library data uses all 2233 nuclides and 66256 reactions that are catalogued in these libraries If the TENDL 2011 TENDL 2012 and TENDL 2013 library data are used the number of nuclides increases to 3873 with a corresponding increase in the number of reactions The dominant nuclides at the end of a sequence of irradiation pulses can be readily identified from lists of nuclides ordered by various radiological quantities derived from the inventory These lists do not show which dominant nuclide arose from which initial target nuclide and by what path The subsidiary calculations in FISPACT II provide tools for the user to probe the re actions and decays in detail Unlike FISPACT 2007 pathways and sensitivity analyses can be undertaken for a series of irradiation pulses rather than for just a single irra diation pulse Pathways calculations can be to arbitrary depths and automatically identify loops that make significant differences to the contribution of the paths on which they lie The pathways calculation identifies how much of the inventory of each of the dominant nuclides came from which initial nuclide and by what chains and loops of reactions and decays Specific routes and paths can be probed independently from the dominant nuclide lists and specific cross sections and decay rates can
156. energy limits of 20 MeV CCFE Page 169 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT II User Manual Name Number of groups WIMS 69 GAM II 100 XMAS 172 VITAMIN J 175 TRIPOLI 315 The method of presentation in Table is designed to make clear in which energy ranges particular structures have most groups and will therefore give a good repre sentation of the cross sections Each group energy displayed in eV is the maximum energy of the group with the minimum being given by the upper energy of the next group The final entries are the minimum energy of the final group Table P1 lists the two original high energy structures VITAMIN J4 211 groups and TRIPOLI 351 groups which are still limited to 55 MeV and below Table P1 omits the lower energy groups below 20MeV which are the same as the VITAMIN J and TRIPOLI groups A further four group structures have been added to provide an increased upper energy bounds of 200 MeV and 1 GeV to allow the addition of o and y induced reactions and to provide for more precise modelling of reaction thresholds and the resonance ranges These additional groups are Name Number of groups LANL 66 CCFE 162 LLNL 616 CCFE 709 The CCFE 162 structure was introduced for studies of charged particle projectiles and induced activation and transmutation The CCFE 709 group structure is an engineered extension of the LLNL 616 structure It has 50 tally bins pe
157. ent daughter pairs the number of nuclides to be output and the type of analysis SENSITIVITY ANALYSIS FOR IRRADIATION PHASE no of steps irradiation time 7 88940E 07 secs flux 4 27701E 14 n cm 2 s Number of samples 400 Number of pd edges 4 Number of nuclides 7 Sensitivity to cross section errors Next are tables of the parent daughter pair properties a list of the nuclides whose sensitivities are being evaluated and the type of distribution and distribution cutoff assumed for the Monte Carlo calculations Base cross section data index parent daughter sigma Sigma unc a zai nuc_no i nuc_no cm 2 1 220460 233 i 210460 219 0 39039E 25 0 35942E 01 2 220460 233 i 210461 220 0 10142E 25 0 35942E 01 3 220480 235 i 210480 222 0 11049E 25 0 87272E 02 4 220480 235 i 210470 221 0 15312E 26 0 54053E 02 Output nuclides nuc no 219 221 222 223 224 205 7 232 Normal x cutoff 3 0000 3 0000 std dev The summary output tables give the input and output mean and fractional standard deviations of reaction or decay rates and the resulting output inventory means and fractional uncertainties The final summary output tables give values of the Pearson correlation coefficients that are above the threshold specified by the keyword argument rnsens1 See also Appendix on page 156 i sigma base sigma unc base sigma mean sigma unc 1 3 90391E 26 3 59421E 02 3 87468E 26 3 38911E 02 2 1 01424E 26 3 59421E 02 1 01360E 26 3 4755
158. ent neutron data for 472 nuclides or elements from 1 H 1 to 100 Fm 255 for more details visit https www oecd nea org dbforms data eva evatapes jeff 32 CCFE Page 200 of 200
159. er Verlag 1968 The JEFF 3 1 3 1 1 Radioactive Decay Data and Fission Yields Sub libraries OECD NEA 2009 JEFF Report 20 O N Jarvis Low activity materials reuse and disposal Technical Report AERE R 10860 Atomic Energy Research Establishment 1983 J W Eastwood Using Graph Theory Methods for Enumerating Pathways Technical Report CEM 081203 TN 1 Culham Electromagnetics Ltd September 2010 A C Hindmarsh 2001 http www netlib org odepack opkd sum CCFE Page 129 of 200 CCFE R 11 11 Issue 6 REFERENCES FISPACT II User Manual 43 44 45 46 47 48 51 52 93 54 95 56 A C Hindmarsh L R Petzold and A H Sherman 2005 http www oecd nea org tools abstract detail USCD1229 J Kopecky D Nierop and R A Forrest Uncertainties in the European Activation File EAF 3 1 Subfile EAF UN 3 1 Technical Report ECN C 94 015 Stichting EnergieonderZoek Centrum Nederland ECN 1994 The JEFF 2 2 Radioactive Decay Data OECD NEA 1994 JEFF Report 13 E Brown and R B Firestone Table of Radioactive Isotopes John Wiley and Sons 1996 R W Mills An initial study of providing energy dependent fission product yields Technical Report JEFF Doc 1157 OECD NEA 2006 International Commission on Radiological Protection editor Dose Coefficients for Intakes of Radionuclides by Workers Number 68 in ICRP Publicatio Perga mon Press Oxford 1995 International Commi
160. ersion SLSODES may provide sufficient accuracy This software is presented as a set of Fortran 77 library routines with an interface defined by CCFE Page 164 of 200 A 14 Method of Solution of Rate Equations CCFE R 11 11 Issue 6 FisPACT II User Manual the subroutine argument list of the top level driver routine The present development treats DLSODES and SLSODES as black boxes and no significant modifications to their internal details have been made In the following these two variants of the solver will be referred to collectively as LSODES The use by the solver of the sparsity structure of the matrix describing the rate equa tions is very significant practically since it yields a reduction in running time by a large factor because during the calculation LSODE performs many solutions of linear sys tems of equations derived from the matrix which for general matrices requires O N arithmetic operations The stiffness of the system of equations limits the choice of numerical method LSODES uses the backward differentiation formula method also known as Gear s method When the equations are not stiff other methods are feasible and LSODES uses an implicit Adams method For simplicity of implementation FISPACT II always calls on LSODES to apply Gear s method there is no easy rapid way of determining whether or not a system of equations is stiff so an automatic selection of method does not seem to be possible limited
161. es 15 Included with the FISPACT II Version 2 20 distribution are over 270 test input files per library that exercise the code options and datasets FISPACT II has been compiled using Intel Oracle gfortran and g95 Fortran compilers and has been shown to give the same results apart from roundoff errors on Unix Linux Mac OS and Windows machines 1 1 Structure of the Document Section contains a brief summary of what FISPACT II does and Section 3 summarises the differences from FISPACT 2007 Section 4 provides an introductory guide to the new user by walking through some sim ple example runs of FISPACT II the data for which may be found in the getting started subdirectory of the test data tree provided with the code It also describes how to speed up calculations by using compressed ENDF libraries and reduced nuclide indexes Section 5 on page 44 contains a summary of all the keywords that users may use in the run control files Section 6 shows where to find examples of uses of these keywords in input files and the resulting output files in the fispQA2010 and fispQA test directories A guide to interpreting the output from FISPACT II is given in Section 7 on page 99 The first appendix page expands on the details of the model used in the code and the decay reaction and spectrum types recognised by the software The second appendix page 169 summarises the EAF library data The final appendix page 195 summarises the EN
162. es the ENDF data forms introduced in Version 2 of the code CCFE Page 5 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual Revision History Revision Date Issued by Comment 0 0 28 October 2010 2 4 31 May 2013 2 5 30 June 2014 2 6 4 Aug 2014 1 0 1 December 2010 1 1 10 May 2011 1 2 21 September 2011 2 1 14 March 2012 2 2 18 July 2012 2 3 27 September 2012 J G Morgan J W Eastwood J W Eastwood J W Eastwood J W Eastwood J W Eastwood J W Eastwood J W Eastwood J W Eastwood J W Eastwood Initial Template Partial document for alpha release Revised and expanded for 1 00 beta release Revised and expanded for Release 1 00 Revised and expanded for Release 2 00 Minor corrections for Release 2 00 01 Minor corrections for Release 2 00 02 Revised and expanded for Release 2 10 00 Revised and expanded for Release 2 20 00 Revised for Release 2 20 01 CVS document source revision Name Release 2 20 01 RCSfile sd7 tex v Revision 1 74 Author jim Date 2014 08 08 10 07 01 CCFE Page 6 of 200 CONTENTS CCFE R 11 11 Issue 6 FISPACT II User Manual Contents Summary 1 Introduction 1 1 Structure of the Document What FISPACT II Does 2 1 Library Data Preparation 2 2 Inventory Calculation 2 3 Subsidiary Calculations Differences from FISPACT 2007 3 1 New Features 0 0 3 2 Obsolete Fe
163. es the modifications made for EAF 2010 The uncertainty data are greatly simplified but complete no covariance information is provided However the file enables FISPACT II to give broad brush estimates of uncertainties CCFE Page 188 of 200 B 3 Neutron Flux Sample Data CCFE R 11 11 Issue 6 FisPACT II User Manual B 3 Neutron Flux Sample Data The collapsed cross sections depend strongly on the nature of the projectile spectra and so it is important to use the appropriate spectrum together with the appropriately weighted cross section data The majority of neutron application spectra stem from light water assemblies mock ups or reactors where the integral responses are strongly if not solely influenced by the energy ranges of the fission spectra and thermal maxwellian Fusion spectra that have been obtained from magnetic confinement MCF or inertial confinement fu sion ICF present typical D D 2 5 MeV or D T 14 MeV peaks sometimes accompa nied by a higher energy tail but also showing rather different slowing down profiles Accelerator driven beam spectra are important in their role in nuclear data acquisition and materials research but also for medical therapeutic and diagnostic applications In essence the particle spectrum profile through the collapsing process emphasises the energy region of most importance for each application Transferring data from one application or energy range to another should be done with grea
164. es would probably be masked by the activity of the iron In order to remove the background this keyword causes the iron matrix to be replaced by a matrix of a fictitious stable nuclide with no induced reactions so that the printed inventories and dose rates refer only to the impurities An example of the use of this keyword is IRON MASS 1 0 2 Fe 99 9999 Ag 1 0E 4 In this run corresponding to the irradiation of 1 ppm of silver impurity in iron the output will be due only to the reactions on silver However the y dose rate will represent decays of silver isotopes in an iron matrix rather than in solid silver 5 2 25 LOGLEVEL level 2 See Section on page 51 for more information 5 2 26 LOOKAHEAD This keyword is used for fine tuning of the pathways and uncertainty calculations If it is present then the pathways and uncertainty calculations do a look ahead over the entire cooling phase and add any dominant nuclides that appear in the late time CCFE Page 69 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual dominant nuclide lists to the list of target nuclides created at the ZERO time for use in the pathways calculation This keyword should be used with care as it may lead to slow calculations or even code failure through heap exhaustion because of large numbers of target nuclides in the pathways calculation For further discussion on this see the UNCERTAINTY keyword on page 83
165. eyword Input syntax errors Run terminating test 142hb1 FATAL ERROR run terminated for details see runlog file test142b log This run has generated three syntax error messages for the file test142b 1 The first error is on line 16 Inspection of the FUEL keyword on line 11 shows that it specifies four nuclides whereas five are listed with the fifth on line 16 To correct this change the 4 to 5 on line 11 After detecting an error the syntax checker attempts to recover by skipping the tokens in the input stream until the next keyword is detected and then continuing the checking This may as in this example provide the opportunity to correct more than one error using the output of one run depending on the effectiveness of the recovery The second error detected is for DOSE This is an instance of a keyword that has a variable number of arguments If you consult Section on page you will see that if the first argument of DOSE has value one then the second argument should not be present Deleting the second argument 1 0 corrects this error The final error arises because of keyword ambiguity A minimum of three characters are needed for a keyword provided the abbreviation is unique The MIN on line 36 could be an abbreviation for either MIND or MINS and so an error message is issued It is clear from the context on line 36 that the keyword should be MINS and so adding an S fixes this error The run is always ter
166. f alternative ENDF format library data sets This caused a major overhaul of the data input parts of the software and a huge expansion of the number of nuclides and reactions that can be treated better fission yield data and cross section data in more energy groups up to higher energies can now be used see Appendix C on page 195 This version can also handle more irradiating projectiles y n p d o and provides additional diagnostic outputs kerma dpa and gas appm rates if the ENDF format library contains the required input data CCFE Page 15 of 200 CCFE R 11 11 Issue 6 2 WHAT FISPACT II DOES FISPACT 1 User Manual Features added in Release 2 10 are self shielding using the universal sigmoid curve approximation c f Appendix A 4 4 processing of covariances between differ ent reactions extended pathways analysis features handling additional isomeric states and the capability to use the TENDL 2012 library data The present release Release 2 20 includes ENDFB VIT 1 JEFF 3 2 and JENDL 4 0 nuclear data libraries and the capability to read and process them It can compress these and the TENDL libraries to allow faster collapse calculations There is a new interface module that allows inventories to computed for mul tiple flux spectra and returned to a calling program There is also a separate FrsPACT MP program that allows inventory calculations to be performed for simulataneous irradiation by several different projectil
167. finitions different from those in the ENDF manual 16 FisPACT II reads and discards these data A 3 Decay Modes The code will allow 27 decay modes by which the parent nuclide j can decay to daughter nuclide i T hese are listed in Table The index IRT is the index used in the code The index RTYP is the ENDF 6 reaction type code used for reaction product code MT 457 Sec 8 3 page 8 5 The table also includes two unused IRT codes and another to indicate an unknown decay mode so the maximum IRT is 26 The decay constant A appearing in Equation 7 is the sum of the decay constants for the transmutation of nuclide j to i In terms of the directed graph the edge shown in Figure 6 corresponds to the combination of a subset of 23 possible decay edges from j to i Figure 8 Decay processes e g a decay may produce secondary gas nuclides that are included in the inventory calculation Some of the decay modes listed in Table have secondary gas nuclides that are included in the inventory calculation and the number of these is NSEC and their names are included in the right hand column of the table This is illustrated in Figure 8 The primary reaction leads to a secondary edge in the directed graph and other products from these decays are regarded as debris that is ignored When there are gas nuclide secondaries then a secondary edge from nuclide j to the gas nuclide is associated with the decay from j to i There may be 0 1 or 2 secondar
168. for a variety of targets of different shapes sizes and compositions Six nuclides that exhibit strong resonances were used individually not as mixtures The model was then extended by Martinho et al 22 who defined an effective length y for cylinders of finite height but a more significant extension was provided by Sal gado et al 23 who defined an average Gres by assigning weights to each resonance CCFE Page 148 of 200 A 4 Neutron Activation CCFE R 11 11 Issue 6 FisPACT II User Manual and forming an average of the individual Gres factors calculated for each resonance individually The weight of resonance i is I gl res where T is the neutron scattering width g is the statistical factor 2J 4 1 2 21 4 1 J is the spin of the resonance state I is the spin of the target nucleus Then the effective self shielding factor is m Ni WiGres zi Gres where each z is calculated from Eq using the effective length of the target y and the resonance parameters for resonance 7 49 This model has been generalised further in two ways to make it suitable for application in FISPACT II First the average self shielding factor is computed from the resonance parameters given in the resolved resonance range defined in the ENDF File 2 data for a subset of the nuclides specified with the SSFFUEL or SSFMASS keywords It is assumed that the resonances for the mixture of nuclides are separated in energ
169. g file The error reporting first gives the internal file name output and the file to which it is connected collapse out The error message has a number 00001 a severity Fatal a record of from where it was issued module files m subprogram files open point 9 and a brief explanation of the error FISPACT II is designed to avoid accidental overwriting of existing output files To run the program with the fileroot collapse all the output files collapse lt ext gt see Table 1 and the collapxo file must first be moved away or deleted If you prefer the default option of overwriting old run data used by FISPACT 2007 then you can add the CLOBBER keyword to the top of the input file Figure 1 shows the input and output files used in a collapse run where the file unit names are mapped to real files according to the specifications in the files file In cases where self shielding using the probability table method is included there are additional inputs through the prob tab file unit gt Figure 1 The files used by FISPACT II in the cross section collapse run example The files file maps the internal file names shown in the figure to the actual files used by the run CCFE Page 31 of 200 CCFE R 11 11 Issue 6 4 GETTING STARTED FISPACT II User Manual 4 3 Decay Data Condense The next stage is to condense the decay and fission data The input file for this is condense i NOHEADER SPEK GETDECAY 1 FISPACT
170. hat obtain the library data to be modified 5 2 39 PATH nlink indzp i i 1 nlink 1 This keyword allows a particular pathway consisting of nlink reactions and decays to be specified The nlink 1 nuclides in the pathway are input using their identifiers e g Tel29m For backwards compatibility the R and D have been retained but are not used Any character e g X could be used instead All reactions and decays between a given parent and daughter nuclide are retained and the path calculation gives a breakdown of the percentage of the inventory of the daughter due to each reaction and decay that leads to it from the specified parent This keyword is necessary only if a special investigation of pathway information is needed Pathway data can be generated automatically for all the dominant nuclides by using the UNCERTAINTY keyword PATH might be used for a particularly com plicated pathway not generated automatically or to investigate nuclides only formed in small amounts Path inventories are calculated over all the timesteps until the ZERO keyword is encountered It is possible when using this keyword to produce first a standard inventory and then the numbers of atoms of the daughters are specified in subsequent runs using the RE SULT keyword No inventory then needs to be calculated for these runs investigating the pathways An example of the use of this keyword is CCFE Page 75 of 200 CCFE R 11 11 I
171. he group wise cross sections including only those resonances with peaks in the relevant energy bin Then this array of energy dependent self shielding factors is applied to each energy dependent cross section before the cross section col lapse The principle underlying this model of self shielding is that the resonances perturb the spectrum of the applied neutron flux Consequently the self shielding factors should modify the cross sections for all reactions However the effect of self shielding varies from reaction to reaction because of the differing energy dependencies of the cross sections A 5 Fission The EAF libraries have very little induced fission yield data and relatively few nuclides At most the fission yield data is in three energy ranges and an extrapolation procedure is used to fill in missing data To assess the effects of fission of actinides without fission yield data fission associations were defined using the asscfy data stream so that actinides without fission yield data use the data of a nuclide with similar properties Surrogate daughters were introduced to fill in where daughter nuclides are not included Subsection describes these The new TENDL 2013 ENDF libraries have fission yield data for many more nuclides and these data are tabulated in energy bins in the same manner as for cross sections and covariances With data available for many more nuclides the fission association and surrogate daughter algorithms
172. he output file will contain information about the conversion what fraction of the input groups are included in each output group and details of the input and the output spectra The converted spectrum is written to the file connected to the fluxes stream named in the files file this contains the standard information for a fluxes file e ndstrc values representing the flux values cm s 1 in each group starting with the high energy group e First wall loading MW m e Text string maximum of 100 characters identifying the spectrum Note that although the text string can contain 100 characters only the first 22 will be used as the spectrum identifier so these should provide an unambiguous description The conversion is done on an equal flux per unit lethargy basis e g if one of the input groups is split into two or more groups in the converted spectrum then the fraction of particles in each output group is determined by the ratio of each lethargy interval of the output structure to the total lethargy interval of the input structure There is a restriction on the number of arbitrary energy groups this must be greater than 2 An example of the use of this keyword is CCFE Page 50 of 200 5 1 Library Data Preparation CCFE R 11 11 Issue 6 FisPACT II User Manual GRPCONVERT 99 172 In this case a spectrum in 99 groups is converted into the XMAS 172 group structure 5 1 9 LOGLEVEL level 2 The error logging modu
173. he sum of all the previous cooling time intervals after the keyword ZERO Examples of the use of this keyword are ZERO TIME 2 5 YEARS ATOMS TIME 7 5 YEARS ATOMS Following irradiation the start of cooling is specified by the keyword ZERO Invento ries at the elapsed times of 2 5 and 10 years are output 5 3 32 WALL wall This keyword may also be used in the initial conditions section of the input file see Section 5 2 62 5 3 33 ZERO This keyword is used to reset the time value to zero after an irradiation After ZERO the output will show COOLING TIME rather than TIME in the title for the interval It also sets the flux to zero but the FLUX keyword should also be used This keyword must be used after an irradiation if the keyword GRAPH is also used in the input file This keyword initiates the calculation and output of pathways as specified by the UNCERTAINTY ROUTES or PATH keyword in the initialisation phase If neither ZERO nor RESULT keywords are present then no pathways information will be output NOTE Irradiation steps can be specified after the ZERO keyword has been specified if so desired This allows one to investigate pathways for a subset of the irradiation steps or to get graphical output for irradiation steps CCFE Page 96 of 200 5 4 Miscellaneous CCFE R 11 11 Issue 6 FisPACT II User Manual 5 4 Miscellaneous Comments can now be placed throughout the input file 5 4 1 com
174. here their use has changed error and warning messages are written to the log file Most obsolete keywords can be ignored by the new code or are replaced with their new equivalents so that runs of FISPACT II can still be conducted with historical input files Note that the new keyword reader will recognise a truncated keyword if three or more leading letters of the keyword given uniquely identify it The reader will not recognise keywords that have extra letters at the end e g NOFISSION will not be recognised as NOFISS The following list gives a summary of the changes in keywords CLOBBER is new keyword that allows existing output files to be overwritten By default FISPACT II terminates with a fatal error rather than overwriting an existing output file COLLAPSE still works but is deprecated use GE TXS instead CONV is now obsolete and its use generates a warning message instructing users to use TOLERANCE instead COVARIANCE is a new keyword used to instruct FISPACT II to read and condense covariances between different reactions DOMINANT is now obsolete use UNCERTAINTY and SORTDOMINANT instead to control dominant nuclide output EAFVERSION is now used to distinguish between EAF and ENDF format data libraries version 8 causes reading of the new style ENDF library files while 7 or less causes the reading of EAF libraries EAF 2007 and EAF 2010 data CCFE Page 23 of 200 CCFE R 11 11 Issue 6 3 DIFFERENCES FROM FISPACT
175. hese are collapsed and stored for use in evaluating uncertainties and sensitivities see Appendices and B 2 5 ENDF The cross section data in the TENDL 2013 data sources come in two group structures the CCFE 709 group scheme for neutron induced cross sections and the CCFE 162 group scheme for p d a and y induced reactions The cross sections are collapsed in the same manner as is used for the EAF data but the cross section uncertainties are found using the ENDF 6 LB 5 covari ance data in the manner described in Appendix The EAF cross section and uncertainty data for the 616 group structure have been converted to the same format to provide Verification and Validation for the new data input and processing A preliminary ENDF data compression step can be used c f Section 4 8 to provide a binary version of the ENDF data that gives a much faster collapse calculation The effect of self shielding on collapsed cross sections may be introduced using either the probability table method or the universal sigmoid curve method In the condensing task decay constants branching ratios and discrete decay spectra are read from the EAF Appendix B 4 or ENDF Appendix C 5 decay data files The y and X ray lines are used to construct 24 group spectra for use in computing gamma doses from the inventories In cases where the y spectrum data are not available then approximate spectra may be constructed for the purpose of estimating g
176. his keyword stops uncertainty information from being used It should be used if a cross section library with no uncertainty component is being collapsed or if such a collapsed library is used with the UNCERTAINTY keyword This keyword can still be used so long as only pathway data are required Note that if this keyword is used with the ERROR keyword then the user must supply values of the fractional error ermat If output of the data libraries is requested with the PRINTLIB keyword and no uncertainty data exist then NOERROR must be used In all cases the keyword must come near the top of the input file before the keyword FISPACT 5 1 12 NOFISS This keyword stops the fission yield data from being input and processed during the preparation of the arrayx file It causes substantial speedup of the calculation but will cause errors in the inventory predictions if fission is important It is advisable not to use this keyword if the initial inventory contains actinides 5 1 13 NOHEADER This keyword stops the printing of the header and user information at the beginning of the output and is useful to reduce the amount of printed output 5 1 14 PROBTABLE multxs 0 usepar 1 This keyword causes the probability tables to be read for the nuclides specified by the elements listed by the SSFCHOOSE keyword The self shielding is applied to the existing cross section value as a multiplicative factor if multxs is set to 1 If it is set t
177. hose keywords that cause actions a summary of the actions is also echoed to the terminal The second part specifies the initial conditions In this case the MASS keyword specifies 1 kg of one element Ti and the number of atoms of each isotope of Ti is computed from internal tables of natural abundances ASS 1 0 1 Ti 100 0 FLUX 4 27701E 14 ND 1 E5 GRAPH 2211 4 UNCERT 2 ATOMS HAZARDS HALF ATWO The FLUX keyword specifies the energy integrated neutron flux em s The next two keywords are output selectors The MIND keyword gives the threshold inventory for a nuclide to be displayed in the output tables The GRAPH keyword causes the generation of a gnuplot data and gnuplot command file for plotting total activity and ingestion dose during the cooling period The UNCERTAINTY keyword causes pathways analysis to be undertaken for the irradiation period and for uncertainties to be output The ATOMS keyword leads to the initial state being printed to the output file The remaining three keywords are output selectors to control the output printed for each time interval of the calculation In this case the selections are output of ingestion and inhalation doses HAZARDS half lives HALF and transport limits ATWO CCFE Page 34 of 200 4 5 Inventory Calculation CCFE R 11 11 Issue 6 FisPACT II User Manual The remainder of the file specifies the irradiation and cooling phases of the inventory calculation
178. hs loops and pathways llle 159 10 Sample neutron spectra 2 eh 191 11 Sample neutron spectra 2 2 e 192 12 Magnetic confinement fusion neutron spectra 00 193 13 Processing using NJOY PREPRO and CALENDF List of Tables 1 Filename extensions for user input and output files 2 Mapping of internal unit names to external EAF library files 3 Mapping of internal unit names to external ENDF directories 4 Mapping of internal unit names to other input data files 5 Pages on which the Keywords recognised by FISPACT II are defined 6 Gamma spectrum energy groups 68 7 Main inventory table entries lee 103 8 Optional inventory table entries o lll 103 9 Atomic displacement energies used to compute DPA 10 Decay Types MT 457 recognised by the code 138 11 Decay Radiation Types MT 457 recognised by the code 139 12 Neutron induced reactions recognised by the code 140 13 Additional MT numbers for Gas production Dpa and Kerma assessment 143 14 Additional MT numbers for reactions that are silently ignored 143 15 CALENDF MT number 2 2 242 2 2 2 202 4 16 The types of target geometry recognised by FisPACT II 148 17 Maximum y energies for various decay modes 155 18 The largest decay rates in the EAF library
179. iance data Table 4 Mapping of internal unit names to external flux files and to intermediate cross section and decay files unit unit description of file name number arb_flux 3 Energy group structure projectile spectrum and wall loading for arbitrary group structure fluxes 20 Projectile spectrum and wall loading for a standard group structure arrayx 13 Input and output condensed decay library collapxi 12 Input collapsed cross section library collapxo 17 Output collapsed cross section library all the files needed for you to perform the runs described in the following subsections In getting started there is a files file that contains the following absorp ind_nuc Crossec crossunc fluxes fluxes gamma attenuation data EAF2007data eaf abs 20070 index of nuclides to be included EAF2010data eaf index 20100 100 group Library cross section and uncertainty data neutron induced GAM II library EAF2010data eaf n gxs 100 fus 20100 EAF2010data eaf un 20100 EEF first wall fluxes collapsed cross section data in and out collapsed cross section data collapsed cross section data collapxi collapxo Library decay data decay CCFE Page 28 of 200 EAF2010data eaf dec 20100 001 4 2 Cross section Collapse CCFE R 11 11 Issue 6 FisPACT II User Manual Library fission data asscfy EAF2010data eaf n asscfy 20100 fissy
180. ides are given with a code representing the reaction between them The cross section codes are listed in Table 12 on page and the diagnostic cross section codes are given in Table 13 on page 143 CROSS SECTIONS IN BARNS The cross section for the specified reaction is given in barns followed by the error in percent 147E 03 1 0E 01 H 902E 04 1 1E 00 He 453E 02 2 4E 01 Li n 2n H 322E 02 1 6E 00 H n g He 493E 05 3 6E 01 He n nd He 076E 02 2 1E 01 Li n g 523E 05 1 0E 01 n p 142E 02 5 0E 01 n g i 524E 04 9 3E 00 861E 04 2 9E 00 Li 573E 03 1 9E 02 Be 572E 03 9 5E 00 Be n g Li 816E 04 8 2E 00 Li n p Li 204E 02 5 0E 01 Be n 2n Be 128E 01 3 1E 00 Be n d 784E 03 2 9E 00 n d i 113E 02 2 9E 01 n g 667E 04 9 6E 00 2 2 3 3 6 6 049E 01 1 0E 01 Li 7 n 2n Li 857E 03 1 6E 00 Li 7 n na 166E 02 2 1E 01 7 7 7 7 9 9 CCFE Page 119 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual 7 3 4 Bremsstrahlung candidates The fourth section contains the table of Bremsstrahlung candidates using the criteria described in Appendix The user may select nuclides from this table for input with the BREMSSTRAHLUNG keyword BREMSSTRAHLUNG CANDIDATES NUCLIDE AV BETA MeV AV GAMMA MeV NUCLIDE AV BETA MeV AV GAMMA MeV NUCLIDE AV BETA MeV HALF LIFE 1 DAY He 1 5613E 00 5 6441E 03 2046E 00 2983E 02 i 6963E 00 Be 4 6473E 00 1 4188E 00 6150E 00 000
181. ies associated with a decay type see Table 10 for details A 3 1 Heating Heating from decay is computed using the average decay energies for light particles electromagnetic radiation and heavy particles that are included in the data in the decay file CCFE Page 137 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual A THE MODEL Table 10 Decay Types MT 457 recognised by the code The column labelled Code EC is the number of secondaries is the description used in output from FISPACT II NS and Secs is an abbreviation for Secondaries IRT RTYP Description AZ AA Code NSEC Secs 1 1 8 decay 1 0 b 0 2 2 B decay or electron capture 1 0 b 0 3 3 isomeric transition IT 0 0 IT 0 4 4 a decay 2 4 a 1 He 5 5 neutron emission 0 1 n 0 6 6 spontaneous fission SF 999 999 SF 0 7 7 proton emission 1 1 p 1 1H 8 8 not used 0 0 0 9 9 not used 0 0 0 10 10 unknown 0 0 0 11 1 5 8 decay neutron emission 1 1 b n 0 12 1 4 8 decay a emission 1 4 b a 1 He 13 2 4 B decay a emission 3 4 bta 1 He 14 2 7 B decay proton emission 2 l btp 1 1H 15 3 4 IT followed by a emission 2 4 IT a 1 He 16 1 1 double 8 decay 2 0 b b 0 17 1 6 8 decay followed by SF 999 999 b SF 0 18 7 7 double proton emission 2 2 pp 2 1H tH 19 2 2 double 8 or electro
182. ile see Section 5 1 21 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 2 52 SSFMASS totm indx2 sym i xp i i 1 indx2 The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 22 In this section it will apply to the actions of the next occurrence of the GET XS keyword 5 2 53 TABI ia This keyword causes the inventory data in columns 1 and 2 the number of atoms and grams of each nuclide to be written to an external file TAB1 Note that the stream CCFE Page 81 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FisPACT Il User Manual number a is now ignored Both NOT1 and TAB1 may be used several times during a run to restrict and restore the output as required 5 2 54 TAB2 ib This keyword causes the inventory data in columns 3 and 7 the activity Bq and dose rate Sv h of each nuclide to be written to an external file TAB2 Note that the stream number ib is now ignored Both NOT2 and TAB2 may be used several times during a run to restrict and restore the output as required 5 2 55 TAB3 ic This keyword causes the inventory data in columns 8 and 9 the ingestion and inhalation dose Sv of each nuclide to be written to an external file TAB3 Note that the stream number ic is now ignored Both NOT3 and TAB3 may be used several times during a run to restrict and restore the output as required 5 2 56 TAB4 id
183. in the mixture are in the apply list of SSFCHOOSE but in general the apply list will contain a subset of the nuclides in the mixture The output from a run using this dataset has the labelling and heading information followed by the probability table data initialisation output that specifies the method of calculation chosen and the source data directory for the probability table data PROBABILITY TABLE INITIALISATION Self shielding factors are computed using partial cross sections Library infinite dilution values are replaced by self shielded probability table values Probability table data directory PTdata tp294 Temperature 294K Printed next is a list of the parent nuclides to which the self shielding factor correction is applied and the name of the files containing the probability table data used Probability Table Application List Nuclide Data File W 182 W182 294 tpe W 183 W183 294 tpe W 184 W184 294 tpe W 186 W186 294 tpe and then a summary of the material mixture used in the dilution calculation CCFE Page 122 of 200 7 5 Universal Curve Self Shielding Collapse Run CCFE R 11 11 Issue 6 FISPACT II User Manual Material Mixture List Nuclide Percentage W 182 26 534 W 183 14 319 W 184 30 680 W 186 28 467 A full list of collapsed cross sections can be obtained using PRINTLIB The collapse run simply summarises the reactions whose cross sections are changed by self shielding The table fo
184. in the input file then table items TOTAL Bq A2 RATIO and EFFECTIVE A2 are displayed where N n Aj TOTAL Bq A2 RATIO a 2 xe and EFFECTIVE A2 is the ratio of the total activity to TOTAL Bq A2 RATIO If the CLEAR keyword is used in the input file then the A2 values are replaced by N CLEARANCE INDEX 2 MiotLi The HAZARDS keyword causes the total ingestion and inhalation doses and the total doses excluding the contribution from tritium to be printed The DENSITY keyword causes the density in g cm to be printed The following output fragment is from Tst_709 test120 using TENDL 2013 data that contains kerma dpa and appm cross sections see Table 13 These output appear at this point in the output only for irradiation steps where there is a non zero flux amplitude Total Displacement Rate n Ddiss Total Displacement Rate n Dinel Total Displacement Rate n Del 6 21857E 17 Displacements sec 2 13196E 18 Displacements sec 3 12211E 18 Displacements sec 5 76667E 08 Displacements Per Atom sec 1 97703E 07 Displacements Per Atom sec 2 89522E 07 Displacements Per Atom sec 1 81982E 00 DPAfyear 6 23903E 00 DPA year 9 13662E 00 DPA year CCFE Page 105 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual Total Displacement Rate n Dtot 7 11931E 18 Displacements sec 6 60194E 07 Displacements Per Atom sec 2 08342E 01 DPAl year 37761E
185. ing the course of an irradiation or where the dependence of cross sections on energy changes significantly due to temperature changes When GETXS is used in the inventory calculation section of the input file its actions are performed immediately so all settings that are to apply to the reading of new cross sections must be declared before the use of GETXS If the projectile spectra at a series of irradiation times are known then it is possible to prepare the corresponding collapx files prior to the inventory calculation An input file that would achieve this is GETXS 1 69 lt lt first collapse gt gt SPEK GETDECAY 1 lt lt condense decay data gt gt FISPACT THREE COLLAPSES AND CONDENSE GETXS 1 69 lt lt second collapse gt gt GETXS 1 69 lt lt third collapse gt gt END END OF COLLAPSE CCFE Page 88 of 200 5 3 Inventory Calculation Phase CCFE R 11 11 Issue 6 FISPACT II User Manual The cross section files and flux files for each of these collapses are specified in the order in which they are used in the files file first collapse input fluxes FLUXES 01 crossec EAF2007data eaf n gxs 069 fis 20070 crossunc EAF2007data eaf un 20070 output collapxo COLLAPX 01 second collapse input fluxes FLUXES 02 crossec EAF2007data eaf_n_gxs_069_fis_20070 crossunc EAF2007data eaf_un_20070 output collapxo COLLAPX 02 third collapse input fluxes FLUXES 03
186. ion with code mt as summarised in Table 12 these data are tabulated in file crossec The second sum in Equation is the production of secondary gas nuclide from the reaction producing nuclide k from j where s is the number of secondaries of nuclide 7 per reaction Table 12 Neutron induced reactions recognised by the code Projectile Products MT AZ AA NSEC Secondaries n total 1 0 0 0 n E 2 0 0 0 n nonel 3 0 0 0 nin 4 0 0 0 n O 5 0 n 2nd IL 3 1 H n 2n 16 0 1 0 n 3n 17 0 2 0 n F 18 0 n na 22 2 4 1 He n n3a 23 6 12 3 4He He He n 2na 24 2 5 1 He n 3na 25 2 6 1 He n np 28 1 1 1 H n n2a 29 4 8 2 He He n 2n2a 30 4 9 2 He tHe n nd 32 1 2 1 H n nt 33 1 3 1 3H n nh 34 2 3 1 He n nd2a 35 5 10 3 7H He He n nt2a 36 5 11 3 3H He He continued on next page CCFE Page 140 of 200 A 4 Neutron Activation CCFE R 11 11 Issue 6 FisPACT II User Manual continued from previous page Projectile Products MT AZ AA NSEC Secondaries n 4n 37 0 3 0 n 2np 41 1 2 1 1H n 3np 42 1 3 1 1H n n2p 44 2 2 2 H H n npo 45 3 5 2 H He n 102 0 1 0 n p 103 1 0 1 1H nid 104 1 1 1 7H nt 105 1 2 1 H n h 106 2 2 1 He nio 107 2 3 1
187. is to be read 5 1 4 FISPACT Title This keyword reads a 72 character title beginning with an containing information about the particular run This title is also used to label the graphs but for the graph title only the first 40 characters are used Note that the keyword is the divider that separates the library input from the initial conditions and irradiation sequence details It is the action keyword that triggers the execution of the queued actions from the library data preparation section of the input file 5 1 5 FULLXS This keyword causes the full energy dependent group cross sections to be stored when the cross section library data are being collapsed 5 1 6 GETDECAY libdecay This keyword has one integer parameter libdecay which is set to zero to read decay data from an existing condensed decay library arrayx file or to one to condense decay and fission data from the EAF or ENDF library files specified in the files file For example to get cross section data from a collapsed library and decay data from a condensed library GETXS 0 GETDECAY 0 FISPACT Irradiation of SS316 steel CCFE Page 48 of 200 5 1 Library Data Preparation CCFE R 11 11 Issue 6 FISPACT II User Manual 5 1 7 GETXS libzs lt ebins gt This keyword has two integer parameters If the first parameter libxs is set to zero then the second parameter should be omitted and cross section data are read from the existing
188. lar case An example of the use of this keyword is SSFMASS 0 000193 1 W 100 0 SSFGEOMETRY 1 0 01 In this case pure tungsten is specified A foil 0 1 mm thick containing 0 193 gem of tungsten with the five stable isotopes in their natural abundances is to be used in the self shielding calculation The SSFGEOMETRY keyword in this section applies to the collapse calculation initiated by the FISPACT keyword The keyword may also appear in the inventory calculation section in conjunction with further GET XS keywords 5 1 22 SSFMASS totm indx2 sym i xp i i 1 indx2 This keyword allows the input of the total mass totm kg and the number indz2 of elements in the material to be used in the self shielding calculation For each ele ment the chemical symbol sym i e g W and the percentage by weight zp i are then read This keyword enables elements to be input with the number of atoms of each isotope calculated by FISPACT II using natural abundance data that are stored internally If an element whose natural abundances are not known is selected then FISPACT II will issue a fatal error message Computations for these cases must use the SSFFUEL keyword CCFE Page 57 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual The SSFMASS keyword is the recommended method of inputting materials unless special isotopic compositions are required An example of the use of this keyword is SSFMASS
189. ld EAF2007data eaf n fis 20070 condensed decay and fission data in and out arrayx condensed decay and fission data Library regulatory data hazards EAF2010data eaf haz 20100 clear EAF2010data eaf clear 20100 a2data EAF2010data eaf a2 20100 The purpose of the files file is to provide a mapping of the unit names used within the program to the names of the EAF library and to other input output data files used by the program but whose names are not constructed from lt fileroot gt Tables 2 3 and 4 summarise the file names and the files they map to The file numbers are aliases for the file names that provide backwards compatibility with FISPACT 2007 FILES files The ENDF 6 format input data libraries use a single file for each parent nuclide where the file name is related to the nuclide name In this case the entry in the files file gives the name of the library directory in which the separate data files for each nuclide are stored Table 3 lists the internal unit names and numbers that are used to map to the nuclide data in the cases of the probability table data and the ENDF format data for cross sections decays and fission yields 4 2 Cross section Collapse The EAF and ENDF libraries contain cross sections as functions of the energy of the in coming projectile For the example we are considering here eaf_n_gxs_100_fus_20100 the incoming projectile is the neutron this is the default but other projecti
190. le in FISPACT II provides error messages identifying the point in the code from which the message is issued together with information identifying its severity and its nature In some cases values are output before the error message for further clarification Six error severities are defined by the value of level fatal error serious error error warning error information debug information logging info oP WN Fr The default is to write messages for severity 2 error warning and higher The LOGLEVEL keyword allows the amount of information written to the runlog file to be varied For example will cause extra information to be output that may help identifwaqy the cause of problems LOGLEVEL may appear repeatedly throughout the input file to increase or decrease the amount of diagnostic information as required 5 1 10 MONITOR monit 0 The progress of a FISPACT II run can be monitored by printing the various keywords as they are read in the input file and reporting the actions they initiate to the standard output The default is not to print this information but it can be switched on by setting monit to 1 For both settings of monit the keywords and their actions are written to the runlog file An example of the use of this keyword is CCFE Page 51 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual In this case the keywords in the input are echoed to standard output 5 1 11 NOERROR T
191. les may be selected using the PROJECTILE keyword FISPAcT II reduces each of these cross sections to a single value by taking an average weighted by the spectrum of the incoming projectile flux collapsing see Equation on page 140 The input file in the getting_started directory that instructs FISPACT II to collapse the cross sections and the cross section uncertainties is collapse i GETXS 1 100 FISPACT COLLAPSE 100_99 WITH FW EEF lt lt print summary of collapsed cross sections gt gt PRINTLIB 4 CCFE Page 29 of 200 CCFE R 11 11 Issue 6 4 GETTING STARTED FISPACT II User Manual END END OF RUN The first argument 1 of the GETXS keyword tells the program to collapse cross sections and uncertainties from the EAF library files connected to the crossec and crossunc file unit names and write the collapsed values to the file connected to collapxo The second argument specifies that the 100 group GAM II energy bins should be used NOTE You must ensure that the correct projectile and the correct number of energy groups are specified as there is no information in the EAF library files for the program to check that consistent values are chosen The ENDF format alternative to the EAF library does contain internal information about the projectile and energy group boundaries If this library is used consistency checking is possible and the new code will issue a fatal error message if discrepancies are found Th
192. library data preparation section of the input file see Section 5 1 21 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 3 26 SSFMASS totm indx2 sym i xp i i 1 indx2 The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 22 In this section it will apply to the actions of the next occurrence of the GETXS keyword 5 3 27 TABI ia This keyword may also be used in the initial conditions section of the input file see Section 5 2 53 5 3 28 TAB2 ib This keyword may also be used in the initial conditions section of the input file see Section 5 2 54 5 3 29 TAB3 ic This keyword may also be used in the initial conditions section of the input file see Section 5 2 55 5 3 30 TAB4 id This keyword may also be used in the initial conditions section of the input file see Section 5 2 56 5 3 31 TIME t This keyword allows the input of the irradiation or cooling time interval t in seconds by default The value of the time may be followed by one of the following keywords CCFE Page 95 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FisPACT Il User Manual SECS MINS HOURS DAYS or YEARS so that time units other than seconds may be used Note that it is important when inputting times that it is the interval time not the total elapsed time that is specified Thus for cooling steps the time printed on the inventory is t
193. ll the values of AA by 1 in Table 12 CCFE Page 152 of 200 A 9 Alpha Activation CCFE R 11 11 Issue 6 FisPACT II User Manual A 9 Alpha Activation The set of reactions allowed for alpha particle activation is identical to the set of 90 reactions for neutron activation The table for these reactions can be obtained by replacing the projectile n by o and increasing all the values of AZ by 2 and all the values of AA by 3 in Table A 10 Gamma Radiation In addition to the activity of irradiated materials another measure of acceptability is the dose rate from emitted y rays FISPACT II uses two approximate estimates of the y dose rate due to irradiation by neutrons contact dose from the surface of a semi infinite slab or dose at a given distance from a point source For both measures the contribution of high energy P particle bremsstrahlung to the total dose rate can be significant and this may be output using the BREMSSTRAHLUNG keyword The formulae used for these are discussed in the following sub subsections A 10 1 Contact gamma dose rate Equation shows the formula used to calculate the y dose rate at the surface of a semi infinite slab of material it is taken from Jaeger 38 B al Ha Ei l D C5 2 us E Es 58 where D surface y dose rate Sv h N number of energy groups in the y spectrum histogram Ej mean energy of the i th energy group c f Table 6 on page 68 lla mass energy absor
194. m is the dimension of the seed array and seed i are the dim integers to seed the pseudo random number generator An example of the use of this keyword is MCSEED 8 437395160 1404128605 572505362 1187264075 454383258 525702629 973594203 1758310677 The value of dim may be found by looking at the log file for a run undertaken without this keyword For example a run using FISPACT II compiled using the Intel Fortran compiler gave the log message run sensitivity Log dimension of seed array 2 Log seed value 502091259 Log seed value 493 In this case two integer values are needed to get reproducible pseudo random numbers CCFE Page 71 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT 1 User Manual 5 2 30 MIND mind 1 This keyword allows the input of a parameter indicating the minimum number of atoms which are regarded as significant for the output of the inventory It is usually not important to consider a few atoms of a nuclide The default value is 1 but this means that inventory tables with an extremely large number of unimportant nuclides will be output and it is recommended that a value such as 10 be used for the mind parameter It is possible to use a parameter value less than 1 if information on a wide range of nuclides is required Note that the value of mind corresponds to the amount of material specified it does not refer to number of atoms for a unit mass A request for a small mind will p
195. ment gt gt Comments may be included anywhere apart from within the text input lines beginning with enclosed by double angle brackets lt lt gt gt These comments may cover several lines of the input file An example of the use of this construction is TIME 2 5 YEARS 6 Test Cases fispQA2007 fispQA2010 fispQA2012 and fispQA Supplied with the FISPACT II software are 110 test input files in the fispQA2007 directory 185 test input files in the fispQA2010 directory 256 test input files in the fispQA2012 and 270 test input files in the fispQA directory together with test output and log files to illustrate the running of FISPACT II for a variety of irradiation and cooling scenarios using all the different EAF and ENDF library files and illustrating the use of all the keywords listed in Section The tests are grouped according to the cross section libraries used Directory Tst 100 uses the 100 group GAM II cross section data and so forth The fispQA2007 tests are based on the test set that was distributed with FISPACT 2007 These are included to show the compatibility with the old style files and INPUT files The fispQA2010 directory repeats some of the 2007 tests using the new format files file and input control files and adds new tests for the new capabilities and data libraries that were added for Versions 0 and 1 of FISPACT II We strongly recommend that you do NOT use the obsolete old style input files but instead use
196. minated with a fatal error message if errors in the input file are detected Correcting all three of the errors detected for test142b i gives input file test142c i and rerunning produces ERROR in INPUT file Detected at argument 10 on line 16 of keyword FUEL on line 11 Argument value is Ti60 Expected argument type is nuclide name Unrecognised nuclide argument Input syntax errors Run terminating test142c FATAL ERROR run terminated for details see runlog file test142c log CCFE Page 41 of 200 CCFE R 11 11 Issue 6 4 GETTING STARTED FisPACT Il User Manual The error has occurred because FISPACT II does not have the nuclide Ti60 in its index of nuclides This example of finger trouble where 6 was entered instead of 5 could not be detected by the earlier run because the context of the token Ti60 was corrupted by the previously incorrect argument count given with the FUEL keyword Going to line 16 changing the 6 to 5 gives a syntactically correct input file Deleting the CLOBBER and MONITOR keywords saving the input file as test142 i and rerunning gives the concise terminal output that signifies a successful run test142 cpu time 2 11 secs No errors warnings provided that the collapx and arrayx files have already been prepared by running collapx i and arrayx i This example and further illustrations of messages generated by input file errors may be found in the subdirectory Tst input errors of fispQA2010 4
197. n Most output selection keywords must also be placed in this section although some may also occur in the inventory calculation section The ATOMS or SPECTRUM keywords may appear in this section to produce output describing the initial inventory before irradiation The initial conditions section is terminated by the first occurrence of the TIME key word which sets the first timestep and causes the start of the solution of the inventory CCFE Page 58 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual equations Exceptionally the section may also be terminated by the END keyword for runs that do not require an inventory calculation 5 2 4 ATOMS When it is used in the initial conditions section of the input file this keyword causes the initial inventory to be printed to the output file This keyword may also be used in the inventory calculation section of the input file see Section 5 2 2 ATWO This keyword causes data on the legal limits of activity for transport of radioactive material to be read the calculations to include these data to be performed and the results for individual nuclides and summed values to be output for all timesteps 5 2 3 BREMSSTRAHLUNG arg nuclb j j 1 iarg This keyword allows the input of the number iarg of nuclides and the identifiers nuclb j for each of the nuclides The identifier should be specified using the format Te129m When the output is generated
198. n at three temperatures 293 6 600 and 900 degree Kelvin Data for a small small number of nuclides are taken from sources different from TENDL Table lists those nuclides from different sources in the TENDL 2013 li brary CCFE Page 197 of 200 CCFE R 11 11 Issue 6 C TENDL LIBRARY DATA FISPACT II User Manual Table 26 Non TENDL evaluations in TENDL 2013 nuclide source date evaluators 5 B 10 LANL APROG G M Hale P G Young 5 B 11 LANL MAY89 P G Young 4 Be 9 LLNL LANL OCTO09 G Hale Perkins et al Frankle 6 C 0 JAERI AUG83 K Shibata 9 F 19 CNDC ORNL OCTO03 Z X Zhao C Y Fu D C Larson Leal 1 H 1 LANL OCTO05 G M Hale 1 H 2 LANL FEB97 P G Young G M Hale M B Chadwick 1 H 3 LANL NOVO01 G M Hale 2 He 3 LANL MAY90 G Hale D Dodder P Young 2 He 4 LANL SEP10 G Hale 3 Li 6 LANL APRO6 G M Hale P G Young 3 Li 7 LANL AUGSS P G Young 7 N 14 LANL JUN97 M B Chadwick P G Young 7 N 15 LANL SEP83 E Arthur P Young G Hale 8 O 16 LANL DECO5 Hale Young Chadwick Caro Lubitz C 2 Fission Yield Data The fission yield data need to be provided for each actinide and incident particle The files are supplied in an ENDF 6 format and are read by FISPACT II with no further processing The default library provided is based on the JEFF 3 1 1 library for neutron induced fission Only 19 of the many nuclides that have fission have any fission yield data in JEFF 3
199. n capture 2 O b b 0 20 1 55 8 and double neutron emission 1 2 b 2n 0 21 1 555 67 and triple neutron emission 1 3 b 3n 0 22 1 5555 8 and quadruple neutron emission 1 4 b 4n 0 23 5 5 double neutron emission 0 2 2n 0 24 5 55 triple neutron emission 0 3 3n 0 25 2 77 B decay double proton emission 3 2 b 2p 2 1H tH 26 2 777 8 decay triple proton emission 4 3 b 3p 3 1H tH tH 27 2 6 P decay followed by SF 999 999 bt SF 0 CCFE Page 138 of 200 A 4 Neutron Activation CCFE R 11 11 Issue 6 FisPACT II User Manual Table 11 Decay Radiation Types MT 457 recognised by FisPACT II The column headed Code is the description used in output from FISPACT II STYP Radiation Type Code 0v gamma rays gamma 1187 beta rays beta 2 ec 8 electron capture and or positron emission ec beta 3 not known not known 4 a alpha particles alpha 5n neutrons n 6 SF spontaneous fission fragments SF Tip protons p 8 e discrete electrons e 9 x X rays and annihilation radiation x A 3 2 Gamma spectrum A 22 or 24 group histogram is generated by nearest grid point binning of the intensities of discrete gamma and X ray lines STYP 0 or 9 contained in the data from the decay file A 3 3 Neutron yield The spontaneous fission neutron yield STYP 5 is accumulated using the decay yields contained in the decay file A 4 Neutron Activation The main a
200. n described in Section 7 1 1 Header and run information The output file always begins with a header identifying the version of the code and the CVS repository export Tag for the Release version If the NOHEADER keyword is absent this header is followed by a summary of the information given in Section 3 of this manual Transmutation Activation Inventory Code United Kingdom Atomic Energy Authority and CCFE Page 99 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual Culham Electromagnetics Limited Release 2 20 June 2014 Authors James Eastwood and Guy Morgan Copyright c 2009 14 UK Atomic Energy Authority and Culham Electromagnetics Limited Printed after the header information is the box containing unique identifying informa tion for the run RUN IDENTIFICATION INFORMATION INITIAL CROSS SECTION DATA Collapsed library timestamp 18 16 44 2 August 2014 EAF source library label EAF 2010 100Gp THE EUROPEAN ACTI FLUX file label EEF FW NORM 1MW M2 GAM II TOT 4 DECAY DATA Condensed library timestamp 18 16 45 2 August 2014 EAF source library label EAF 2010 THIS RUN timestamp 18 16 47 2 August 2014 fileroot testl name of FILES file files FISPACT title IRRADIATION OF TI EEF FW 1 0 MW M2 See the testl log file and summary details at the end of this file for further information on files used by this run Note that only the initial cross section
201. n neutron D T 1 I 1 r 1 i fi ji fi i107 5 91 5 99 5 9 5 9 5 99 5 91 5 99 5 99 107 Neutron energy eV d TU Dresden D T Neutron spectra CERN H4IRRAD 400 GeV c NHLDPM 10 T T T T T T T 10 10 10 10 Neutron energy eV f CERN H4IRRAD Sample neutron spectra B 6 Radiological Data CCFE R 11 11 Issue 6 FisPACT II User Manual T L flux per lethargy Neutron flux per lethargy 4 i A a S T 1 Neutron ux pel a lo a a o E 1 107 919 5 09 5 9 5 9 5 9 Sa 5 9 5 99 107 107 919 5 99 5 9 5 9 95 995 St 10 95 99 5107 Neutron energy eV Neutron energy eV a ITER D T b ITER D D Figure 12 Magnetic confinement fusion neutron spectra in eaf d xs which have fission cross sections have any fission yield data in UKFY 4 0 at relevant energies For the remainder a neighbouring fission yield is used For the EAF 2010 library the file eaf d asscfy 20100 connected to the stream asscfy contains these associations B 5 3 Proton eaf p fis and eaf p asscfy eaf p fis is taken completely from the UKFY 4 0 fission yield library and FISPACT II reads the file in ENDF B VI format with no pre processing Only 19 of the 90 nuclides in eaf p xs which have fission cross sections have any fission yield data in UKFY 4 0 at relevant energies For the remainder a neighbouring fission yield is used For the EAF 2010 library
202. n page 153 and for the dose from a point source at a specified distance POINT SOURCE see Equation on page 154 DOSE RATE PLANE SOURCE FROM GAMMAS WITH ENERGY 0 20 MeV IS 5 63098E 04 Sieverts hour 5 63098E 06 Rems hour If most of the dose rate is produced by nuclides with approximate y spectra then the following warning message will be given CCFE Page 108 of 200 7 1 The Inventory Run output File CCFE R 11 11 Issue 6 FISPACT II User Manual WARNING gt 20 OF DOSE FROM NUCLIDES WITH NO SPECTRAL DATA TREAT DOSE AND GAMMA SPECTRUM WITH CAUTION 7 1 7 Dominant nuclides At each step the inventory is sorted into descending order of radiological quantities and tables of nuclides at the tops of these lists are printed see SORTDOMINANT key word In all cases dominant nuclides as measured by activity total heat production dose rate gamma heating and beta heating are displayed If the HAZARDS keyword is used nuclides are also sorted by ingestion and inhalation dose and CLEAR adds columns with sorting by clearance index DOMINANT NUCLIDES NUCLIDE ACTIVITY PERCENT NUCLIDE HEAT PERCENT NUCLIDE DOSE RATE PERCENT NUCLIDE INGESTION PERCENT Bq ACTIVITY kW HEAT Sv hr DOSE RATE Sv INGESTION 2507E 14 6006E 02 Total 6310E 04 3853E 05 er 3656E 13 34 91E 00 4964E 02 69 33E 00 Sc 48 1357E 04 73 45E 00 4216E 04 53 57E 00 4682E 13 19 73E 00 3894E 03 23 30E 00 Sc 46 3625E 04 24 20E 00 7023E
203. nce on how to use the keywords in the input files to specify the desired calculations FisPACT II offers the user more help in developing new input files than FISPACT 2007 does because it has new input file syntax checking and error reporting This is illustrated by the erroneous example test142a input file It is in the subdirectory Tst input errors of directory fispQA2010 and is shown below CLOBBER NOHEAD MONITOR 1 GETXS 0 GETDECAY 0 FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 DE FUE Ti4 ITY 4 54 4 00619E24 18148E23 28210E24 91755E23 Ti6 79178E23 MI E5 GRAPH 320123 FLUX 4 27701E14 TOLERANCE 1 1 0E8 2 0e 3 UNCERT 3 SENSITIVITY SIGMA 1E 10 2 1 Ti48 Sc48 S L 6 7 Ti48 Ti49 0 D Rh oF tO to ja CCFE Page 39 of 200 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 CCFE R 11 11 Issue 6 4 GETTING STARTED FisPACT Il User Manual Ti49 Sc48 Sc48 ATOMS DOSE 1 1 0 ATOMS TIME 1 MIN ATOMS TIME 1 HOURS ATOMS TIME 1 DAYS ATOMS TIME 7 DAYS ATOMS The syntax checker tries to report as many errors as it can in one pass but interaction of errors may lead to more than one test run being needed to locate all the syntax errors The CLOBBER and MONITOR keywords are included at the top of the input file to help in the debugging of the input file CLOBBER is used to eliminate the need to clear up the files generated by failed tests before reruns are undertaken
204. nd energy group g is computed using a weighted sum over all the nuclides q 1 Q in the mixture The fraction f of the mixture is nuclide q Nuclides in the mixture may or may not be included in the list of nuclides to which the self shielding correction is to be applied Nuclides to which self shielding corrections are applied must be in the mixture list The first approximation is given using the total cross sections from the cross section library Q LIB tot oO D d p 9 Y A F a il q 1 p DFq where Y gh S g 5 a MP p g y 34 1 y Over the energy range for which the probability table data are available for those nuclides in the mixture for which self shielding corrections are being applied the ap proximation given by Eq is iteratively refined using Q tot i S0 LIB tot a q 9 d e 35 9 2 10 q 9 got q g 00 i tot i i 1 5 g _ LIB tot a p 9 d p g d pg 5 c p 9 T TS 36 Replacement of LIB data If there is only one reaction MT in the CALENDF macro partial group then the replacement formulae would be given by replacing the c P values in the above equations by the infinite dilution cross sections obtained from the CALENDF data When there is more than one reaction in the macro partial set then the dilution effect has to be apportioned according to the LIB reaction cross sections If the partial self shielding scaling factor option is chosen then the
205. nd its loops is below the path floor The pruning weights are computed from the coefficients of the rate equation matrix The formulae used for single and for multiple pulse irradiation are those derived in 41 A 12 1 Algorithm Pathways are calculated from a single source nuclide to multiple target nuclides and then they are sorted into target nuclide order for output For each source nuclide lists of paths and loops ordered in decreasing importance are found Parent to daughter rather than daughter to parent data ordering is used to simplify the extraction of adjacency information and of reaction rates from the compact storage structure used to store the rate equation matrix coefficients The computation of the significant paths and loops for a given source nuclide uses a five step process designed to prune unnecessary searches and thereby reduce computational effort Step 1 build a breadth first search BFS tree representation of the digraph that visits as parents only once all significant nuclides that are descendants of the source nuclide Significant nuclides are ones that are descendants of the source nuclide that may be reached by a path whose weight is above the path_floor threshold Step 2 repeatedly search the BFS tree of Step 1 to find all graph edges that lie on paths from the source nuclide to the target nuclide or lie on loops that intersect these paths Step 3 build a brute force BFS tree using those edges th
206. nergetic electrons but it is assumed that the same expressions are valid for the emission of P particles which have a continuous energy distribution if the mean energy is used for Ep The value of Z used in Equation is calculated from Z Zij 66 j where Zj atomic number of the j th element and fj atomic fraction of the j th element i e number of atoms of j total number of atoms A 10 5 Bremsstrahlung candidates Only a subset of all the nuclides in the decay library needs to be considered for bremsstrahlung production Nuclides may make a contribution to the y dose rate because of bremsstrahlung emission from energetic 8 particles The following criteria are applied by the code to the EAF decay library Appendix B 4 to give the nuclides displayed by the PRINTLIB 4 keyword option the nuclide is radioactive with a half life gt 0 1 years or in the case of a short lived nuclide the half life of the parent gt 0 1 years e the nuclide is radioactive with a half life lt 5 0 x 1016 years e the nuclide has an average P energy average y energy e the nuclide has an average G energy gt 0 145 MeV A 11 Monte Carlo Sensitivity Estimation FISPACT II uses a Monte Carlo approach to sensitivity analysis A series S of inventory calculations is performed with the set of J independent variables X7 i 1 I s 1 S chosen from distributions with means X and standard deviations AX These runs produce a set
207. ng the TIME keyword This keyword may also be used in the initial conditions section of the input file see Section 5 3 2 EAFVERSION neafv 7 The primary use of this keyword is in the library data preparation section of the input file see Section In this section EAFVERSION must be used before the keyword GETXS to which it refers as it determines which input streams from the files file are used to read the nuclear data 53 3 END Title This keyword may also be used in the initial conditions sections of the input file see Section 5 2 9 CCFE Page 87 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FisPACT Il User Manual 5 3 4 ENDPULSE This keyword terminates a loop construct that was started by PULSE The actions for all keywords between PULSE and ENDPULSE are repeated npulse times where npulse is the parameter following PULSE 5 3 5 FLUX fluz2 This keyword may also be used in the initial conditions section of the input file see Section 5 2 13 5 3 6 FULLXS This keyword may also be used in the library data preparation section of the input file see Section 5 3 7 GETXS libxs lt ebins gt This keyword must be used in the library preprocessing section see page 49 to collapse cross sections or to input previously collapsed cross sections It may be also used in the inventory calculation phase to compute new collapsed cross sections where the projectile spectrum changes significantly dur
208. nly one of the four actinides 298U 39pu 74 Pu and 2Pu as parent 5 2 12 FISYIELD nyld symb i i 1 nyld gt When actinides are included in the list of input elements and USEFISSION is spec ified then by default only U235 U238 and Pu239 will produce fission products when they undergo fission If nyld 0 then no fission products are produced from any of the CCFE Page 62 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual actinides If nyld is a positive integer then only the actinides that are specified in the list of identifiers symb e g Am242m produce fission products If nyld is a negative integer then all actinides except those that are specified in the list of identifiers symb e g Am242m produce fission products This facility is included so that information on the irradiated actinides alone can be obtained Also when investigating the properties of various actinides it may be useful to be able to restrict which of these produce fission products Note that fissionable isotopes that have no fission yield data in the selected library do not undergo fission Examples of the use of this keyword are FISYIELD 0 None of the actinides will produce any fission products when fissioned FISYIELD 2 U235 Pu239 Only 7 U and 9Pu will produce any fission products when they undergo fission FISYIELD 2 U238 Am241 All actinides except 28U and Am will produce fission prod
209. ns describe the trans mutation of the initial inventory by nuclear reactions induced by the projectiles and by spontaneous radioactive decay see Appendix A The inventory calculation then proceeds by 1 setting the physical initial conditions of the target 2 setting the output selections 3 specifying the subsidiary calculations 4 computing the irradiation steps 5 performing the subsidiary calculations 6 computing the cooling steps 7 computing summary data Output is written as it becomes available at each step The sequence of steps performed in the calculation follows the sequence of steps spec ified in the user s input file controlling the run The duration of a step is specified by the user and typically ranges from fractions of a second to many years During each step the irradiating flux amplitude cross sections and decay rates are kept constant Also it is assumed that the imposed projectile flux is not modified by the reactions and decays in the target material In consequence the rate equations are linear and have constant coefficients for each step The material is homogeneous infinite and infinitely dilute but in some circumstances self shielding can be accommodated in the model and the description of the evolution of the nuclide numbers is reduced to a stiff set of ordinary differential equations see Appendix on page 133 Unlike FisPACT 2007 1 Fispact II does not use the equilibrium approximation fo
210. nt precision of 15 decimal places Fl This approach to floating point precision is not compatible with the Fortran 77 code of LSODES which uses default REAL and DOUBLE PRECISION declarations in the single and double precision versions respectively FISPACT II uses Fortran 95 intrinsic functions to determine the precision provided by default REAL and DOUBLE PRECISION floating point variables and chooses the one that provides 15 decimal places Unusually a platform may achieve this precision with default REAL and on such a platform FISPACT II would automatically use SLSODES rather than DLSODES A 14 4 Error estimation and step control The LSODES solver controls the accuracy of its calculations by refining its internal timesteps to satisfy a criterion placed on its estimate of the error Estimates are produced separately for each component of the solution vector but these are combined into a single measure of the error using a root mean square norm The acceptance criterion is based on the sum of relative and absolute tolerances so that for the dominant nuclides in a FISPACT II calculation the error is determined by the chosen rtol parameter while for the minor nuclides the tolerance is relaxed by the addition of the atol parameter This avoids the problems that would occur for a pure relative error estimate in the case of zero or very small inventories The solver returns the error estimates of the individual component
211. o define the weight for group i as N Wi i gt i 12 i 1 The collapsed cross section c f Eq 10 X is given by N X gt WX 13 Covariances for cross sections X and Y grouped in energy bins i 1 Nx j 1 Ny are Cov X Yj The collapsed covariance arising from these is given by mE Nx Ny Cov X Y 3 3 W W Cov X Yj 14 i 1 j 1 Cov X Y is not presently used in FIsPACT II but is planned to be used in future in the monte carlo sensitivity calculations T he case of interest at present is that where reactions X and Y are the same and then the collapsed variance is given by N var Cov X X 5 WiW Cov Xi X 15 ij l 3If fluxes are in different energy groups then the GRPCONVERT keyword can be used to remap them to the appropriate groups c f Section 5 1 8pn page 50 CCFE Page 135 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual For EAF data the uncertainty is defined at the three standard deviation point A 3Vvar X 16 and for the TENDL 2013 data it is defined as A vvar X 17 The covariance data are less complete than the cross section data Each covariance data energy group contains several cross section energy groups and in some cases the data in different energy groups are assumed to be uncorrelated The covariance data in the EAF and TENDL 2013 libraries that FISPACT II recognises are the ENDF 16 NI type data with LB 1 5 6 or 8 The
212. o zero then the cross section values in the energy CCFE Page 52 of 200 5 1 Library Data Preparation CCFE R 11 11 Issue 6 FisPACT II User Manual groups for which there are probability table data are replaced with apportionment determined from the EAF or ENDF library data The second argument defines the way in which the self shielding factors SSF are computed usepar 0 use the total cross section to calculate one SSF for each nuclide and apply this factor to all relevant cross sections usepar 1 use macro partial cross sections to calculate a separate SSF for each macro partial and apply it to relevant reactions contributing to that macro partial If the 616 group infinitely dilute cross sections in the EAF data library and the CAL ENDF probability table data were fully consistent these two methods of calculation would give the same answers However the EAF data do not contain the elastic scattering cross sections and so cannot give the correct total cross section The CAL ENDF probability tables do give the correct total cross sections but only provide cross sections for sets of macro partials and so have to use the EAF data to apportion the cross sections when they are used to replace EAF values when self shielding is in cluded The replacement option is the recommended one but both options are included so that the user can assess the uncertainty of the effective collapsed self shielding fac tor n expert mode
213. ol then the last values specified will be used The atol parameter is significant in relaxing the accuracy requirement on the results for the minor constituents of an inventory and to avoid excessive demands on the solver If accurate results are required for minor constituents of the inventory indicated by the setting of a small mind parameter then atol should be reduced as well An example of the use of this keyword is TOLERANCE 0 5 0E3 1 0E 3 In this case the absolute and relative tolerances for the main inventory calculation are reduced by a factor of two compared with the default values See Appendix for more information 5 2 59 UNCERTAINTY iuncer 0 path floor 0 005 loop floor N 510 01 maz depth 10 iuncer gt This keyword allows user control of the uncertainty estimates and pathway information that are calculated and output for each time interval This is primarily specified by the parameter iuncer 0 The allowed values are resets default values for a particular run and permits other values to be specified by the following parameters which can be present only for this value of uncer 0 no pathways or estimates of uncertainty are calculated or output 1 only estimates of uncertainty are output although all the pathway information is calculated 2 both estimates of uncertainty and the pathway information are output 3 only the pathway information is output 4 now generates a fatal error message path
214. ollows The EAF TENDL infinitely dilute values old sigma are replaced by the sigmoid curve effective cross sections new sigma The effective self shielding factor is the ratio of new to old values Factors greater than 90 00 are omitted from the table parent nuclide Na 22 Ar 37 K 37 K 42 K 44 Sc 43 Sc 44 44m 46 47 47 49 49 51 55 57 58m 59 53m daughter mt nuclide Na Ar K K K Sc Sc Sc Sc V V Cr Mn Mn Mn Mn Mn Fe 22 38 38m 43 45 44 45 45 47 48 48 50 50 52 56 58 59 60m 54 old sigma barns 51796E 00 11255E 03 13898E 03 93336E 03 63969E 03 22430E 03 83744E 03 27846E 02 25279E 02 T1165E 03 30012E 03 25513E 03 73542E 03 29548E 03 10650E 02 05890E 03 83010E 03 33656E 03 70607E 03 new sigma barns 20106E 00 29233E 03 71960E 03 29917E 03 14525E 03 78151E 03 01304E 03 33443E 03 22951E 03 16265E 03 35352E 03 54519E 03 26221E 03 10514E 03 75274E 03 27035E 04 05237E 03 70209E 03 83581E 03 self shielding factor 19s 82 80 76 86 62 62 57 73 87 13 81 78 85 88 87 81 89 81 12 68 39 43 42 68 29 37 67 24 33 52 13 31 14 55 92 99 51 parent nuclide cl K K K Sc Sc Sc Sc Sc V V Mn Mn Mn Mn Fe Co 36 37 38 43 42m 43 44 44m 46m 45 48 50 51 56 58 58 53 daughter nuclide cl K K K Sc Sc Sc Sc
215. ompute collapsed covariances between different reactions if covariance data are available in the reaction data files Tables of the collapsed covariances and correlations may be printed using the print 4 option with the PRINTLIB keyword 5 1 3 EAFVERSION neafv 7 This keyword is used to select the format of the nuclear data libraries to be read It is not needed if the EAF 2007 or EAF 2010 libraries are to be used neafv is an integer indicating the EAF library version For backwards compatibility the default value of neafv is 7 indicating the EAF 2007 or EAF 2010 libraries The new value of 8 is used to indicate the ENDF format libraries that are now an alternative to EAF Note that FISPACT II has not been validated for earlier versions of the EAF library than EAF 2007 EAFVERSION must be used before the FISPACT keyword as it determines which input streams from the files file are used to read the nuclear data This keyword may also be used in the initial conditions and inventory calculation sections of the input file if the cross sections or decay rates are to be changed during the course of a run see Sections and An example of the use of this keyword is EAFVERSION 8 NOERROR PROJ 2 GETXS 1 162 CCFE Page 47 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual FISPACT TENDL 2011 gxs 162 deuteron 1 MeV 200 MeV The 162 group ENDF format cross section library for a deuteron projectile
216. on and improved integral tests Nucl Inst and Methods in Phys Research A 589 85 108 April 2008 A J Koning Hilaire S and Duijvestijn M Talys 1 4 a nuclear reaction program http www talys eu 2012 NRG Nuclear Research and Consultancy Group J Ch Sublet and R A Forrest EASY II 12 decay data library Technical Report CCFE R 12 18 CCFE 2012 J Ch Sublet and R A Forrest EASY II 12 biological clearance and transport indices library Technical Report CCFE R 12 19 CCFE 2012 M B Chadwick et al ENDF B VII 1 Nuclear Data for Science and Technology Cross Sections Covariances Fission Product Yields and Decay Data Nuclear Data Sheets 112 12 28872996 Dec 2011 K Shibata Iwamoto O T Nakagawa N Iwamoto A Ichihara S Kunieda S Chiba K Furutaka N Otuka T Ohsawa T Murata H Matsunobu A Zuk eran S Kamada and J Katakura JENDL 4 0 A New Library for Nuclear Science and Engineering J Nucl Sci Technol 48 1 1 30 2011 J Katakura JENDL FP Decay Data File 2011 and Fission Yields Data File 2011 Technical Report JAEA Data Code 2011 025 JAEA Mar 2012 CCFE Page 131 of 200 CCFE R 11 11 Issue 6 REFERENCES FISPACT II User Manual This page has been left intentionally blank CCFE Page 132 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual APPENDICES A The Model A 1 The Rate Equations FIsPACT II follows the evolution of the inventory of nuclides in a target ma
217. ons is obtained from the uncertainty file 5 2 11 FISCHOOSE ncho fischo i i 1 ncho FISCHOOSE affects the choice of actinides included in the pathways analysis not the actinides included in the activation calculation USEFISSION and FISYTELD are the keywords to use to alter the treatment of actinides in the activation calculation When actinides are included as trace elements in a material then dominant nuclides that can be formed as a result of the fission of an actinide will be considered in the calculation of pathway information Although uranium and thorium may have been the only actinides input neutron induced reactions and decay will create many other fissionable actinides and the user may wish to specify which of these actinides are considered as possible parents when calculating the pathways By default all actinides are considered but by setting ncho and specifying the identifiers of the actinides the user can limit the nuclides to be included In most cases minor actinides are unlikely to have significant impact on the total radi ological quantities and so are unlikely to be part of the important pathways Also this keyword only affects the calculation of pathways all actinides are considered during the calculation of inventories unless the use of other keywords indicates otherwise An example of the use of this keyword is FISCHOOSE 4 U238 Pu239 Pu240 Pu242 In this case any pathways containing a fission reaction can have o
218. ored to the requirements of LSODES Also the user program needs to provide subprograms that give LSODES values of the driv ing function F y t and its Jacobian J OF Oy The subroutine argument lists of these user supplied routines are defined by the internal details of LSODES and cannot be changed CCFE Page 165 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual LSODES was written at a time when computers had much smaller main memories than present machines and the code improves memory efficiency by overlaying floating point and integer workspace Correct operation of the code requires knowledge of the ratio of the sizes of the storage units for floating point numbers and integers The authors of LSODES expected their users to be aware of this ratio on their computing platforms and this number is hard wired into the code as supplied in two places Users had to edit the source before using the code FisPACT II improves the portability of the solver by automatically determining the required ratio of floating point to integer storage sizes using standard Fortran 95 intrinsic functions The argument list of the driving routine for LSODES has been extended to pass this ratio to the relevant points in the solver and the hard wired data statements have been removed The precision of the floating point computations in FISPACT II is controlled by using a specific real kind in all declarations to achieve a floating poi
219. oup 22 Group Group Energy range Group Energy range number MeV number MeV 1 0 00 0 01 1 0 00 0 01 2 0 01 0 02 2 0 01 0 10 3 0 02 0 05 3 0 10 0 20 4 0 05 0 10 4 0 20 0 40 5 0 10 0 20 5 0 40 1 00 6 0 20 0 30 6 1 00 1 50 7 0 30 0 40 7 1 50 2 00 8 0 40 0 60 8 2 00 2 50 9 0 60 0 80 9 2 50 3 00 10 0 80 1 00 10 3 00 3 50 11 1 00 1 22 11 3 50 4 00 12 1 22 1 44 12 4 00 4 50 13 1 44 1 66 13 4 50 5 00 14 1 66 2 00 14 5 00 5 50 15 2 00 2 50 15 5 50 6 00 16 2 50 3 00 16 6 00 6 50 17 3 00 4 00 17 6 50 7 00 18 4 00 5 00 18 7 00 7 50 19 5 00 6 50 19 7 50 8 00 20 6 50 8 00 20 8 00 10 00 21 8 00 10 00 21 10 00 12 00 22 10 00 12 00 22 12 00 14 00 23 12 00 14 00 24 14 00 gt 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual 5 2 22 HAZARDS This keyword causes data on potential ingestion and inhalation doses to be read and the dose due to individual nuclides to be printed in the output at all timesteps 5 2 23 INDEXPATH This keyword causes the index of nuclides that lie on the significant pathways to be written to the ind_nuco channel if a pathways calculation is selected see also Section 5 2 59 on the UNCERTAINTY keyword 5 2 24 IRON This keyword should be used only for calculations where small quantities of impurities in an iron matrix are to be irradiated In a run without this keyword the activity of the impuriti
220. output O Blocks 1 5 1 Block 1 only 2 Blocks 2 3 4 and 5 3 Block 5 only 4 Block 3 to extra significant figures in two column format with the collapsed dpa and kerma cross sections added 5 Block 6 6 Block 7 Note that if no uncertainty data exist in the library then the keyword NOERROR must be used before PRIN TLIB Note that it is recommended that a separate FISPACT II run giving a library output and no inventory be done for each decay data library and kept for reference An example of the use of this keyword is PRINTLIB 1 The library data for decays half lives average energies y spectra and fission yields are output CCFE Page 77 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual 5 2 42 PROBTABLE multxs 0 usepar 1 The primary use of this keyword is in the library data preparation section of the input file see Section 5 1 14 Its use here is in conjunction with a subsequent GETXS keyword 5 2 43 ROUTES par dau nmaz pmin iprpa As an alternative to specifying a particular pathway with the keyword PATH the keyword ROUTES can be used This will search for all pathways from the parent nuclide par to the daughter nuclide dau with a maximum of nmaz links reactions or decays The contribution of each pathway is calculated and if the number of daughter atoms is greater than pmin the pathway and the contribution will be printed in the output The parameter iprpa must be
221. pathways but use the pathways in uncertainty estimates 1 display pathways for all dominant nuclides at each pathways reset If the PATHRESET keyword is included in the initial conditions section of the input file then pathways are recalculated at each step where there are new target nuclides and all occurrences of the PATHRESET keyword in the inventory calculation phase are ignored The recommended usage of this keyword is to use it where required in the cooling phase of the inventory calculation c f Section 5 3 17 on page CCFE Page 76 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual 5 2 41 PRINTLIB print This keyword causes the printing of the data libraries in a readable form The output consists of seven blocks of data the contents of which are 1 decay data including fission yields if appropriate for each nuclide 2 the branching ratios of decays for each radionuclide 3 the cross section data including uncertainties for each reaction in the specified projectile spectrum 4 nuclides which will give a bremsstrahlung contribution to the y dose rate 5 the projectile spectrum used to collapse the cross section library 6 the photon and particle decay spectral lines energy and intensity for unstable nuclides 7 a list giving the library source of the ENDF cross section data file for each nuclide ENDF library input only The value of the parameter print determines which blocks are
222. pplication of the code is to neutron activation calculations In these the transmutation of a nuclide j to another nuclide i and in some cases additional secon daries may result from 1 one of the decay processes listed in Table 10 2 one of the neutron induced reactions listed in Table 12 The output for the induced reaction produced by FISPACT II uses the code n for the neutron projectile and g n p d t h a respectively for products y N T h a when printing output Decay processes are described in Section above There are three special cases in the list of neutron induced reactions CCFE Page 139 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual Elastic scattering MT 2 in Table 12 This special case is where the projectile z elastically scatters from the target nuclide and is designated z E Other reactions MT 5 in Table 12 The set of reactions labelled as other reac tions is a special case and is designated z O for projectile z Fission MT 18 in Table 12 The neutron induced fission reaction n F is a special case and is treated below in Section A 5 The total effective cross section cl used in Equation 7 is obtained by summing the contributions from the different reactions 0 gt st prod me X sie n sii Do 24 mt kAt where cl n prod mt is the cross section for the production of nuclide from nuclide j through the neutron induced react
223. projection operator SF maps cross section energy bins to covariance energy bins as illustrated in F igure 7 1 biniin bin k is 5i 0 otherwise 18 k k 1 k covariance A a E cross section io o iH Figure 7 Projection operator SE maps cross section energy bins to covariance energy bins The shaded energy bins have S 1 and all others have S 0 Using SF the formula used to construct estimates of the covariance matrix from the library data are as follows M LB 1 Cow X X 5 SESE Fp XiX 19 k 1 M M LB 5 Co Xi Yj M SESP Fig XY 20 k 1k 1 M M LB 6 Cal C o S Er 21 k 1 k 1 M LB 8 Cov Xi X Dd SPS 1000F Koning 22 k 1 M or Y Sf61000F 23 k 1 The LB 1 case Equation 19 is the one that applies to the computation of A for the EAF data Covariances are described by a fraction for each k bin and the different k bins are assumed to be uncorrelated CCFE Page 136 of 200 A 3 Decay Modes CCFE R 11 11 Issue 6 FisPACT II User Manual The LB 5 6 and 8 cases appear in the TENDL 2013 libraries The LB 5 data for X and Y referring to the same reaction are used to compute A and are assumed to have LS 0 The LB 6 data give cross correlations between collapsed cross sections These are read but not used in the present version of the code The LB 8 data are produced from the same source as the LB 5 data for X Y with some of the cross correlations discarded and use de
224. ption coefficient Men p of air m kg Um mass energy attenuation coefficient p p of the material m kg B build up factor 2 S rate of y emission MeV kg s C 3 6 x 10 e converts MeV kg s71 to Sv h The EAF library file absorp see Section B 7 contains y p cm g for all elements in increasing Z order u m and pen p cm g for air and the mean energies of the 24 group structure The value of um for the material is calculated from the elemental values mj provided by the absorp data file using Hm D film 59 j CCFE Page 153 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual where f mass of element 7 total mass The value of the emission rate S is calculated using Sy Ei LA t 60 where J is the intensity of energy group MeV and A t is the specific activity of material at time t Bq kg If discrete spectral line data are available then J is obtained by summing the contributions from spectral lines in energy group 7 read from the decay data files If data are not available then an approximate value may be computed as described below in Section A 10 3 A 10 2 Gamma dose rate from point source Equation shows the standard formula taken from Reference 38 for calculation of the dose rate from a point source in air Ny Ha u E r D ex acd MED S E 61 where C Ny ua Sy are as defined above for Equation 58 and ms
225. r Manual 97 58 59 60 61 62 63 65 66 67 68 69 J H Hubble and S M Seltzer Tables of X ray mass attenuation coefficients and mass energy absorption coefficients 1 keV to 20 MeV for elements Z 1 to 92 and 48 additional substances of dosimetric interest Technical Report NISTIR 5632 NIST U S Department of Commerce 1995 J H Hubble Photon Mass Attenuation and Energy absorption Coefficients from 1 keV to 20 MeV Int J Appli Radiat Isot 33 1269 1982 J Ch Sublet A J Koning and D A Rochman Toward a unified ENDF 6 for matted file frame Technical Report CCFE R 11 16 CCFE 2012 A J Koning D A Rochman S Van der Marck J Kopecky J Ch Sublet S Pomp H Sjostrand R A Forrest E Bauge and H Henriksson TENDL 2012 TALYS based Evaluated Nuclear Data Library http www talys eu tendl 2012 2012 R A Forrest et al Validation of EASY 2007 using integral measurements Tech nical Report UKAEA FUS 547 UKAEA 2008 D E Cullen PREPRO 2012 2012 ENDF 6 Preprocessing codes Tech nical Report IAEA NDS 39 Rev 15 IAEA 2012 http www nds iaea org ndspub endf prepro R W Mills UKFY4 1 A set of prototype fission product yield library for neu tron proton deuteron alpha particle photon and spontaneous fission Technical Report NDA1648 3 06 15 NNL 2007 D A Rochman and A J Koning Pb and Bi neutron data libraries with full covariance evaluati
226. r decay constants is calculated CCFE Page 79 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT 1 User Manual If insens4 is set to zero then the merged list of dominant nuclides i e all nuclides that appear on any of the dominant lists is used as the nuclide list See Appendix on page for further details of the sensitivity method and Section on page for a description of the output produced Example SENSITIVITY SIGMA 0 8 2 1 Ti48 Sc48 Ti49 Sc48 Sc48 Parameters for the Monte Carlo calculation may be reset using the MCSAMPLE and MCSEED keywords If insens4 0 then the number of nuclides displayed may be controlled by the SORTDOMINANT keyword 5 2 45 SORTDOMINANT topzz 20 topz 20 This keyword controls the uncertainty calculations and their display in the output file toprr nuclides are included in the dominant list used for uncertainty calculations and topz of them are displayed in the output file topr must be less than or equal to topzz 5 2 46 SPECTRUM This keyword is an alternative to ATOMS It suppresses the inventory output so that only the y spectrum and total values are printed When it is used in the initial conditions section of the input file this summary applies to the initial inventory This keyword may also be used in the inventory calculation section of the input file see Section 5 2 47 SPLIT split 0 This keyword allows the display of an additional summary table at the
227. r energy decade equally spaced in the logarithm of the energy between 107 eV and 10 MeV and thereafter bins with appropriately chosen equally spaced boundaries in energy up to 1 GeV Table 20 Energy group boundaries for the five low energy standard structures TRIPOLI 315 VITAMIN J 175 GAMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 1 1 9640E 7 1 1 9640E 7 1 1 9640E 7 2 1 7330E 7 2 1 7333E 7 2 1 7333E4 7 3 1 6910E 7 3 1 6905E 7 4 1 6490E 7 4 1 6487E 7 5 1 5680E 7 5 1 5683E 7 6 1 4920E 7 6 1 4918E 7 1 1 4918E 7 3 1 4918E 7 continued on next page CCFE Page 170 of 200 B 1 Cross section Group Structure CCFE R 11 11 Issue 6 FisPACT II User Manual continued from previous page TRIPOLI 315 VITAMIN J 175 GAMM II 100 XMAS 172 WIMS 69 grp energy eV grp energy eV grp energy eV grp energy eV grp energy eV 7 1 4550E 7 7 1 4550E 7 8 1 4190E 7 8 1 4191E 7 9 1 3840E 7 9 1 3840E 7 4 1 3840E 7 10 1 3500E 7 10 1 3499E 7 2 1 3498E 7 11 1 2840E 7 11 1 2840E 7 12 1 2523E 7 12 1 2210E 7 13 1 2214E 7 3 1 2214E 7 13 1 1620E4 7 14 1 1618E 7 5 1 1618E 7 14 1 1050E 7 15 1 1052E 7 4 1 1052E 7 15 1 0510E 7 16 1 0513E 7 16 1 0000E 7 17 1 0000E 7 5 9 9998E 6 6 1 0000E 7 1 1 0000E 7 17 9 5120E 6 18 9 5123E4 6 18 9 0480E 6 19 9 0484E4 6 6 9 0482E 6 19 8 6070E 6 20 8
228. r short lived nuclides and includes the evolution of actinide sources in the rate equations The core engine of the FIsPACT IT stiff ode solver is the LSODES package 17 If the inventory calculation includes irradiation then the first step must have a non zero irradiating flux amplitude The rate equation coefficients in subsequent steps may be changed in one or more of the following ways e changing the flux amplitude CCFE Page 19 of 200 CCFE R 11 11 Issue 6 2 WHAT FISPACT II DOES FISPACT 1 User Manual e changing the library cross section data e g to take account of temperature effects e changing the flux spectrum The end of the irradiation heating phase is signalled by the ZERO keyword in the user input file and it is this keyword that triggers the subsidiary pathways and sensitivity calculations as well as resetting the elapsed time to zero The cooling phase is a sequence of steps the same as the irradiation phase although the projectile flux amplitude is usually but not necessarily set to zero and must be zero for the first cooling step The purpose of cooling steps with irradiation is to provide flexibility in the range of applicability of pathways analysis and graphical output The principal output of an inventory calculation step is the inventory of nuclides at the end of the step Secondary outputs are derived from the inventory the choice of which is controlled by a number of the keywords describ
229. r this example is as follows PROBABILITY TABLE CHANGES TO CROSS SECTIONS The EAF TENDL infinitely dilute values old sigma are replaced by the probability table effective cross section new sigma The effective self shielding factor is the ratio of new to old value parent daughter mt cal mt old sigma new sigma self shielding nuclide nuclide barns barns factor 82 83 102 10385E 00 99375E 01 23 74 82 83m 02 14997E 01 52384E 02 23 89 82 79 07 26039E 05 25547E 05 99 93 82 07 78139E 06 76192E 06 99 75 82 07 85923E 07 85918E 07 100 00 83 02 41600E 00 06712E 00 31 24 83 07 04463E 05 73309E 05 96 13 83 107 92752E 06 92574E 06 99 95 84 02 42228E 01 35932E 01 30 74 84 02 81381E 04 06302E 04 61 78 84 07 28951E 05 28901E 05 99 99 86 02 12173E 00 68884E 00 54 10 86 07 14859E 05 14858E 05 100 00 W W W W W W W W W W W W W cx o ooo C2 oda C3 Co Co Cn iO 4 CO CO CO N CO DY CO Lr 010 n uw se NNN e For fission MI 18 and other MT 5 the daughter nuclide names are respectively replaced by fission and other 7 5 Universal Curve Self Shielding Collapse Run The cross section collapse using the universal sigmoid curve approximation to com pute the self shielding factor and the effective collapsed cross sections differs from the standard collapse c f Section 4 2 in that 1 The files file must specify an ENDF format cross section library that includes
230. rance data hazards 14 Biological hazards data fispact example would generate the output file example out the log file example log and so forth When FISPACT IT is run with only the single argument lt fileroot gt then the program looks in the order given for files in the present working directory with the name files Files or FILI ES The first one found is used and if none of them is found then the program will flag a fatal error to the log file and close down If the filename of the files will be used iles file is given as second argument then a file of name In order to work through the following examples copy the FISPACT II test tree into your own work space In the test directory is a sub directory getting started with 1Users of Mac OS and Wi ndows systems should avoid having more than one files file in a directory because of problems with the lack of case consciousness CCFE Page 27 of 200 CCFE R 11 11 Issue 6 FISPACT II User Manual 4 GETTING STARTED Table 3 Mapping of internal unit names to external CALENDF and ENDF directories of library files unit unit Library directory name number CALENDF or ENDF prob_tab 51 Probability table data xs endf 52 Resonance cross section and covariance data dk endf 53 Decay and its uncertainty data fy endf 54 Fission yield data xs endfb 55 Binary compressed cross section and covar
231. rd Both TAB1 and NOT1 may be used several times during a run to restrict the output as required 5 2 35 NOT2 This keyword switches off the output to the external file that was switched on by the TAB2 keyword Both TAB2 and NOT2 may be used several times during a run to restrict the output as required 5 2 36 NOT3 This keyword switches off the output to the external file that was switched on by the TAB3 keyword Both TAB3 and NOT3 may be used several times during a run to restrict the output as required 5 2 37 NOTA This keyword switches off the output to the external file that was switched on by the TABA keyword Both TAB4 and NOTA may be used several times during a run to restrict the output as required 5 2 38 OVER ja This keyword enables library data to be modified for a particular case It can be called several times during an irradiation if required ja specifies the nuclide that is to have data changed The identifier can be specified using the format Te129mf The OVER keyword is followed by one of four keyword options Note that the nuclide name specification is not case conscious so Tel29m or TE129M or tel29m or tE129M etc could equally well be used CCFE Page 73 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FisPACT Il User Manual ACROSS jb sig n n 1 ngr jb is the daughter of the reaction and sig n is the new cross section barns for the n th energy group For all existing EAF and END
232. rences X X and AX AX give a measure as to how well the sample set matches the distribution and as S increases these two differences should tend to zerd Similarly Y should tend to the value of the base calculation and AY gives the uncertainty in the dependent variable resulting from uncertainties in the independent variables The sensitivity of dependent quantity Y on independent variable X is assessed using the Pearson product moment correlation coefficient Xu X YF 7 SXiY DARAY 71 The magnitude of r is less than one and a magnitude close to one indicates strong linear correlation Values of r close to 1 will be found for reactions or decays on principal pathways leading to nuclide j and values close to 1 are expected for reactions or decays acting as sinks on pathways The best fit line relating Y to X is given by Y Y Xi Xi fri 2 This is not strictly true the sample standard deviation will be systematically smaller than the input value because of the truncation of the tails of the distributions for normal and log normal distributions CCFE Page 157 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual FISPACT II writes tables of means standard deviations and correlation coefficients to output and writes the raw data X Y i l hj l J4 s 1 S to file sens for post processing by the user A 12 Pathways The reaction network illu
233. roduce meaningful results only if the atol parameter of the TOLERANCE keyword is also set to a suitable small value less than the value of mind An example of the use of this keyword is In this case all nuclides with numbers of atoms lt 10 are omitted from the inventory output 5 2 31 NOCOMP This keyword causes the table of elemental compositions to be omitted from the in ventory printout 5 2 32 NOSORT The default output includes a sorted list of the dominant nuclides where a maximum of topz 10 nuclides is shown The nuclides are sorted by activity heat y dose rate ingestion dose inhalation dose 6 heat y heat and clearance index The list can be removed by the use of this keyword to reduce running time although including the list typically increases the running time by only a few percent Note that removing the dominant nuclide list also disables the output of pathways and uncertainty estimates that might have been requested by the UNCERTAINTY keyword CCFE Page 72 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual 5 2 33 NOSTABLE Use of this keyword inhibits the printing of any stable nuclides in the inventory and is useful when the inventory is large and it is required to save space This keyword may also be used in the inventory calculation section of the input file 5 2 34 NOT1 This keyword switches off the output to the external file that was switched on by the TAB1 keywo
234. ror logging and code timing objects Consequently the user will observe some differences in the output particularly the improved re porting of errors and warnings Dynamical memory allocation is used throughout so the same code works for all the data sets irrespective of their sizes files file changes Filename mnemonics can be used as an alternative to unit num bers to link external filenames to FISPACT II input and output streams and com ments can be included in the files file Repeated entries for the same stream name are read into a queue of external filenames which are used in sequence See the GETXS keyword below CCFE Page 22 of 200 3 2 Obsolete Features CCFE R 11 11 Issue 6 FisPACT II User Manual A summary of the physical models and algorithms is given in Appendix A more complete treatment of the model algorithms architectural design and software speci fication is given in References 3 4 20 3 2 Obsolete Features There is very little reliable data for the treatment of sequential charged particle re actions and so this feature has been disabled in the new code to simplify the user interface The new algorithms for integrating the rate equations and for computing pathways have led to a number of related keywords e g LEVEL CONV DOMINANT LOOPS becoming redundant See the following section for details 3 3 Keyword Changes All the FISPACT 2007 keywords are recognised by the new program but w
235. rradiation is described by CCFE Page 142 of 200 A 4 Neutron Activation CCFE R 11 11 Issue 6 FISPACT II User Manual Table 13 Additional MT numbers for Gas production Dpa and Kerma assessment MT Description 201 202 203 204 205 206 207 301 302 303 304 318 401 403 407 442 443 444 445 446 447 z Xn Total neutron production z Xy Total gamma production z Xp Total proton production z Xd Total deuteron production z Xt Total triton production z Xh Total helion He production z Xa Total alpha particle production Kerma total eV barns Kerma elastic Kerma non elastic all but MT 2 Kerma inelastic MT 51 91 Kerma fission MT 18 or MT 19 20 21 38 Kerma disappearance MT 102 120 Kerma for protons Kerma for alphas Total photon eV barns Total kinematic kerma high limit Dpa total eV barns Dpa elastic MT 2 Dpa inelastic MT 51 91 Dpa disappearance MT 102 120 Table 14 Additional MT numbers for reactions that are silently ignored MT Description 19 n f First chance fission reaction 20 n nf Second chance fission reaction 21 n 2nf Third chance fission reaction 38 n 3nf Fourth chance fission reaction 46 101 z n Neutron production with residuals in excited states 110 Unassigned 118 150 Various p d t o reactions 221 unassigned 251 253 scattering of neutron 402 energy release parameter
236. ry of the cross section libraries For more information see References 8 9 11 B 2 1 Groupwise neutron induced eaf n gxs Many group cross section libraries in EAF format are available for the neutron induced library that can be used as input to FISPACT II The group boundaries of the LANL 66 WIMS 69 GAM II 100 CCFE 142 XMAS 172 VITAMIN J 175 VITAMIN J 211 TRIPOLI 315 TRIPOLI 351 LLNL 616 and CCFE 709 formats are listed in Appendix B 1 where details of the micro flux weighting spectra are also given Note that three choices of weighting spectra are available for the most general formats This is necessary because of the very different neutron spectra found in pure CCFE Page 187 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT II User Manual fission or fusion applications in addition a flat weighting library is available for the other applications The group boundaries of the LANL 66 WIMS 69 XMAS 172 and TRIPOLI 315 structures are appropriate for fission applications The group boundaries of the GAM II 100 VITAMIN J 175 and TRIPOLI 315 structures are appropriate for fusion applications The VITAMIN J 211 and TRIPOLI 351 group structures cater for applications where the neutron flux may extend to 55 MeV The LLNL 616 up to 20 MeV and CCFE 709 up to 1 GeV group structures cover all applications and energy ranges The CCFE 162 group structure caters for all
237. s calculated then the density of the material must be specified using DENSITY An example of the use of this keyword is FUEL 2 Li6 8 5E24 Li7 1 5E24 In this case lithium highly enriched in the Li isotope is to be irradiated 5 2 15 FULLXS This keyword may also be used in the library data preparation section of the input file see Section For it to be effective it must be specified before the cross section libraries are collapsed i e before the GETXS keyword with arguments 1 ebins CCFE Page 64 of 200 5 2 Initial Conditions CCFE R 11 11 Issue 6 FisPACT II User Manual 5 2 16 GENERIC gener 1 In addition to the normal output of pathway data there is a section showing generic pathway data A generic pathway is one in which all instances of a link of type Nuclide isomer state m or n IT Nuclide state g is replaced by Nuclide state g All pathways that when simplified in this fashion have the same form belong to the same generic pathway and the contributions of all the pathways are added to give the contribution of the generic pathway The default is always to print the generic information but it can be switched off by setting igener to 0 5 2 17 GETXS libzs lt ebins gt This keyword may also appear in the library data preparation section of the input file see Section 5 1 7 When this keyword is used in the initial conditions section of the input file its actions are performed immediately so
238. s four single character flags that are printed immediately following each nuclide identifier Note that the that was present in FISPACT 2007 output has been dropped as the equilibrium approximation is not used in the FISPACT II solver is the convergence flag whose presence indicates a nuclide with larger uncertainty in its inventory It is set if the error for the nuclide is greater that 1 5 times the rms norm error set by the rtol and atol flags see Appendix A 14 amp indicates that no y spectral data were present in the decay data library and that the keyword SPEK was used to calculate a spectrum approximately see Ap pendix A 10 3 If most of the y dose rate is produced from nuclides with this flag then the result should be treated with great caution jii indicates that the nuclide is stable CCFE Page 101 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FISPACT II User Manual indicates that this nuclide was present in the material input specified by the MASS or FUEL keyword 7 1 3 Time line and nuclide inventory The time line is printed at the start of the output produced at the end of an integration step initiated by the ATOMS or SPECTRUM keyword It displays the time interval number the step length and the total elapsed time The ZERO keyword causes the elapsed time counter to be reset to zero and the word COOLING to be added to the time line TIME INTERVAL 2 TIME IS 7 88
239. s into histograms for use in the y dose computations and PRINTLIB output The same bins are used for the approximate y spectra generated when the SPEK keyword is used in the condense phase of the library data processing The default igamgp 0 means that the y spectrum data are output in a 24 energy group structure However if igamgp 1 then the output is in the 22 group Steiner energy structure Note that the structure determined by igamgp is also used when TABA is specified to produce a file of the y spectrum data An example of the use of this keyword is GROUP 1 In this case data will be output in 22 energy groups Table 6 summarises the energy group structures for the 24 and 22 group formats 5 2 20 GRPCONVERT nestrc ndstrc This keyword may also be used in the library data preparation section of the input file see Section 5 2 21 HALF This keyword causes the half life of each nuclide to be printed in the output at all timesteps The units are seconds but if the nuclide is stable then the word Stable is printed If this keyword is not used then an indication of the stable nuclides in the output can be seen in the flags section to the right of the nuclide identifier CCFE Page 67 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual Table 6 The Gamma spectrum energy group structures for the 24 and 22 group formats CCFE Page 68 of 200 24 Gr
240. s not used then only the cross section uncertainties are used in the calculation of uncertainties If iuncty 2 then only the half life uncertainties taken from the decay data library are used in the calculation of uncertainties If twncty 3 then both cross section and half life uncertainties are used Examples of the use of this keyword are UNCERT 2 UNCTYPE 2 Uncertainty calculations will be done but only using the half life uncertainties Cross sections are assumed to have no uncertainties Such a calculation is useful to isolate the contribution generally small of half life uncertainties CCFE Page 85 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FisPACT Il User Manual UNCERT 2 UNCTYPE 3 Uncertainty calculations will be done but using both the cross section and half life uncertainties 5 2 01 USEFISSION This keyword causes fission reactions for which fission yield data are stored in the fis sion yield library to be self consistently included in the matrix describing the inventory equations It should be used in conjunction with FISYTELD whenever actinides or other heavy elements that are transmuted to actinides are specified in the target ma terial When it is absent all fission reactions are omitted from the inventory equations leading to much faster calculations which remain accurate when there are no actinides in the initial inventory and none is produced If there are actinides in the initi
241. s of the solution vector to its calling program This information is used in FISPACT II to flag the 5 This is usually double precision on byte oriented platforms Exceptionally if a specific platform cannot achieve 15 decimal places of precision FISPACT II will not compile CCFE Page 166 of 200 A 14 Method of Solution of Rate Equations CCFE R 11 11 Issue 6 FisPACT II User Manual nuclides with larger than usual error estimates The criterion for flagging outlying nuclides is that the estimate should exceed the specified tolerance by a factor of more than the dimensionless parameter err factor which is set in the code to 1 5 The solver sets an array of logical convergence flags which are used by the output module to place markers in the inventory output This provides the equivalent of the markers used in the output from FISPACT 2007 Specifically LSODES computes a vector w rtol y atol 86 from the solution vector where the parameters rtol and atol are provided by the user through the TOLERANCE keyword The weights w are used with the local estimates of the component wise errors e to compute 1 2 1 gt p D is used as a single measure of acceptability if D gt 1 then LSODES refines its internal timesteps until a satisfactory D is obtained 87 Users should be aware that LSODES works with local errors Estimation of the global error is much harder and in common with many numerical methods LSO
242. se Tate ees A 10 3 Approximate gamma spectrum llle A 10 4 Bremsstrahlung corrections e 0002 eae 155 A 10 5 Bremsstrahlung candidates cles 156 A 11 Monte Carlo Sensitivity Estimation les 156 A T2 Pathways ek ex e wok bon e uU OE a a ed 158 Ac 191 Algorithmi oS 3 ace Eee ea denk UE RR e ee ERE A a 160 A 13 Uncertainty Estimates A 161 A 14 Method of Solution of Rate Equations 162 A 14 1 Properties of the equations a 163 A 14 2 The choice of solver llle A 14 3 The interface to the solver o o 165 A 14 4 Error estimation and step control les 166 A 14 5 Runtime error reporting o B EAF Library Data 169 B 1 Cross section Group Structure 2e 169 B 1 1 Weighting spectra ee 186 B 2 Cross section Data llle B 2 1 Groupwise neutron induced eafn_gxs 137 B 2 2 Probability tables ls 188 B 2 3 Groupwise deuteron induced eaf d gxs 188 CCFE Page 11 of 200 CCFE R 11 11 Issue 6 CONTENTS FISPACT II User Manual B 2 4 Groupwise proton induced eaf_p_gxs 188 B 2 5 Uncertainty eafun sns 188 B 3 Neutron Flux Sample Data ln 189 B 4 Decay Data eafdec 222A 190 B 5 Fission Yield Data cs 190 B 5 1 Neutron eafn fis and eaf n asscfy
243. sed dpa reaction cross section in eV cm The constant eg is the DPA efficiency factor and is set to 8096 p 2757 A list of the dpa cross sections recognised by FISPACT II is given in Table 13 The mean atomic displacement energy Ey is given by 7 Nn Nn Ea M NiEa Zi Y Ni 2 i l i 1 Zi is the atomic number of nuclide i and Eq are atomic displacement energies in eV taken from Table II of Reference 26 with the exception of the value 55eV used for tungsten see Table 9 Alternatively the displacement rate may be estimated using the mean of the displace ment rates of the constituents Nn Diot ead Nid 2Eq Zi 3 l Both options have been evaluated and have been shown to give similar results Equa tion is used in the present version of FISPACT II The displacements per atom is given by dividing this by the total number of atoms Nn DPA RATE Dios YN 4 i l CCFE Page 106 of 200 7 1 The Inventory Run output File CCFE R 11 11 Issue 6 FISPACT 1 User Manual Table 9 Atomic displacement energies used to compute DPA Eq is 25 eV for all other elements Element Eq in eV Element Ej in eV Be 31 Co 40 C 31 Ni 40 Mg 25 Cu 40 Al 27 Zr 40 Si 25 Nb 40 Ca 40 Mo 60 Ti 40 Ag 60 V 40 Ta 90 Cr 40 W 53 Mn 40 Au 30 Fe 40 Pb 25 The kinetic energy released in materials rates are given by Nn KERMA RATE o Y Ri 5 i 1 where k is the collapsed kerma cross se
244. ses are now included 5 The total cross section per atom 0 0 which is related to p can be written as the sum over contributions from the principal photon interactions Otot Ope Ocoh Cincoh Opair trip Span 88 where Ope is the atomic photoeffect cross section Ocoh and Gincon are the coherent Rayleigh and incoherent Compton scattering cross sections respectively Opair and Otrip are the cross sections for electron positron production in the fields of the nucleus and the atomic electrons respectively and aphan is the photonuclear cross section However the latter contribution has been neglected as well as other less probable photon atom interactions The eaf_abs file contains the photon mass energy attenuation coefficient j1 p for all the elements Z 1 100 in increasing Z order The attenuation coefficient u and energy absorption coefficient Men p for air are also listed All data are stored in the same 24 group energy structure as shown in Table 6 on page 68 C TENDL Library Data FISPACT II requires connection to several nuclear data libraries and forms before it can be used to calculate inventories While any libraries in the correct ENDF 6 format could be used c f Appendices DIF below the development of FISPACT II over the last few years has run in parallel with the development of the TALYS based Eval uated Nuclear Data Library TENDL project and those European libraries are the recommended source of
245. ssing codes a combination of the three is needed to extract the data forms that are the most useful in all applications A schematic of the processing sequences is shown in Figure Further details of the data assimilation processes and its history can be found in Reference 59 TENDL 2011 and TENDL 2012 processed data forms differ in some respects 35 36 This is due to enhancements made in the original ENDF 6 compliant TENDL data format and the way the files are processed This is particularly noticeable in the partials kerma and dpa outputted from TENDL 2012 and the more complete usage made of the variance covariance information contained in this library C 1 Cross section Data The principal sources of cross section data are the different generations of the TALYS based Evaluated Nuclear Data Libraries The TENDL 2013 is the latest recom mended evaluated data source for use in any type of nuclear technology applications The principal advances of this new library are in the unique target coverage 2434 nuclides the upper energy range 200 MeV variance covariance information for all nu clides and the extension to cover all important projectiles neutron proton deuteron alpha and gamma and last but not least the proven capacity of this type of library CCFE Page 196 of 200 C 1 Cross section Data CCFE R 11 11 Issue 6 FisPACT II User Manual E e NJOY12 021 PREPRO 2013 CALENDF 2010 reconr linear calendf 2
246. ssion on Radiological Protection editor Age dependent Doses to Members of the Public from Intake of Radionuclides Part 5 Compilation of ingestion and inhalation dose coefficients Number 72 in ICRP Publications Pergamon Press Oxford 1996 A W Phipps G M Kendall J W Stather and T P Fell Committed Equiva lent Organ Doses and Committed Effective Doses from Intakes of Radionuclides Technical Report NRPB R245 NRPB 1991 A W Phipps and T J Silk Dosimetric Data for Fusion Applications Technical Report NRPB M589 NRPB 1991 R A Forrest Dosimetric data for FISPACT2 Technical Report AEA FUS 182 UKAEA 1992 IAEA Regulations for the safe transport of radioactive material 1985 edition and supplement 1988 Technical Report Safety Series No 6 IAEA Vienna 1988 IAEA Application of the Concepts of Exclusion Exemption and Clearance Tech nical Report Safety Standards Series No RS G 1 7 IAEA Vienna 2004 IAEA Clearance levels for radionuclides in solid materials application of exemp tion principles 1994 Draft Safety Guide Technical Report IAEA Safety Series No 111 G 1 5 IAEA Vienna 1994 NIST NIST X ray and gamma ray attenuation coefficients and cross section database Technical report U S Department of Commerce National Institute of Standards and Technology Standard Reference Data Program Gaithersburg Maryland 20899 1995 CCFE Page 130 of 200 REFERENCES CCFE R 11 11 Issue 6 FISPACT I Use
247. ssue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual PATH 3 Ti46 R Ti45 D Sc45 R Sc44m This generated the output Target nuclide Sc 44m 5 060 of inventory given by 1 path 5 060 Ti 46 R Ti 45 b Sc 45 R Sc 44m S 100 00 n 2n 100 00 b 100 00 n 2n 0 00 n p i e 5 06 of the daughter nuclide Sc was formed from Ti along the path Tia 2 TC 3 9 Se n 20 9 Sa A very small percentage lt 0 00596 of Ti was transmuted to Sc by the np reaction MT 103 5 2 40 PATHRESET showpathways For inventory calculations with long cooling times the dominant nuclides at late times may not be significant at the end of the irradiation phase and this leads to poor estimates for the uncertainties One remedy for this is to use the LOOKAHEAD keyword In some instances particularly where there are actinides in the source mate rial the look ahead approach may lead to excessively large numbers of target nuclides in the pathways calculations The PATHRESET keyword provides an alternative means of including late time dominant nuclides Its inclusion leads to the pathways calculation being repeated in the cooling phase and this causes the late time dominant nuclides to be included in the uncertainty calculations There are three values for the showpathways argument 1 display pathways for a target nuclide for which pathways have not been displayed at earlier times O do not display
248. strated in Figure 6 may be described either by the rate equations Eq 7 or as the sum of paths and loops which we refer to as pathways The inventory of a given nuclide computed using the rate equations can equivalently be found by a linear superposition of contributions of flows along the pathways to that nuclide Pathways are used in FISPACT II to aid interpretation and to estimate uncertainties If we know the inventory at the start and end of an irradiation or cooling period then pathways analysis may be used to identify the most significant chains of reactions and decays in transmuting the initial inventory to the dominant nuclides in the final inventory of the step Key aspects of pathways analysis are methods for searching directed graphs or di graphs of the form illustrated in Figure 6 to identify routes from a parent to a chosen descendant and the assembly and solution of rate equations for chosen subsets of nuclides on the pathway to get the flow along the pathway In the directed graph nuclides correspond to the vertices of the graph A parent nuclide is connected to a daughter nuclide by a graph edge Associated with the edge is a flow rate given by the sum of the rates of all reactions and the decay that take the parent to the daughter This flow rate is given by the off diagonal elements of the rate equation matrix The flow rate from parent j to daughter i is given by the element A in row i and column j of matrix A of
249. t care as it can easily lead to misleading and inappropriate numerical results Illustrations of typical spectral profiles are given in Figures 12 which show plots of the neutron fluxes for the following assemblies 1 Magnetic confinement fusion EEF study Figure 10 a 2 Light water reactor Paluel Figure 10 b 3 Fast breeder reactor Ph nix Figure 10 c 4 Fast breeder reactor Superph nix Figure 10 d 5 Inertial confinement fusion NIF ignited Figure 10 e 6 Californium 252 fission Figure 10 7 International Criticality Safety Benchmark Experiment Bigten Figure 11 a 8 JAEA Fusion Neutron Source D T Figure 11 b 9 ENEA Frascati Neutron Generator D T Figure 11 c 10 TU Dresden D T Figure 11 d 11 IFMIF D Li Figure 11 e 12 CERN H4IRRAD Figure 11 f 13 Magnetic confinement fusion ITER D T Figure 12 a 14 Magnetic confinement fusion ITER D D Figure 12 b CCFE Page 189 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FisPACT Il User Manual It is clear that in each of these typical spectra the integral responses are most influenced by the energy region where the profile peaks However it is important not to overlook the upper or lower tails For instance there is more neutron flux above 15 MeV in a fission environment due to the high energy tail of the fission spectrum than in a pure MCF D T fusion only environment B 4 Decay Data eaf dec In addition to cross sections
250. ta TENDL2013data tal2013 n gxs 709 Only those nuclides listed in the ind_nuc file are included in the compressed library The input files for collapse calculations using the compressed ENDF libraries differ from those using the full ASCII libraries only in that the GETXS has first argument 1 rather than 1 and the files file contains a mapping for the compressed library Compressed library cross section data xs_endfb tal2013 n bin CCFE Page 43 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual Typically the compressed library is about one quarter of the size of the full ASCII library and collapse calculations are typically a factor four faster Further reductions in file size and execution times can be realised using a reduced nuclide index 4 9 Reduced Nuclide Index TENDL 2013 contains 3875 nuclides In most applications only a small number of these nuclides are significant Those that are not can be left out of the activation or transmutation calculation without affecting the quality of the physics predictions Omitting the unimportant nuclides leads to much faster FISPACT II calculations and smaller data files FrsPACT II has a simple mechanism for excluding unwanted nu clides only those nuclide listed in the ind nuc master index file are included in the calculation Keeping only significant nuclides can reduce the computation time for collapse calculations and inventory by one or two orders
251. tainty in some radiological quantity Q where Q 5 qt 73 tE St is given by AQ where ANM AQy gt Ft e 74 t t St and N is the number of atoms of target nuclide t formed from the initial inventory and AN is the error in N AN will be computed from the pathways inventories and the fractional squared error Af in the number of atoms of target nuclide t formed along pathway p to that target If we let the set of pathways to target t be Sp then we may write Sp Wee Sa Ss 75 where Sp is the set of pathways leading to target t Sa is the subset of these pathways where the pathway starts from the fission of actinide source nuclide a and 5 is the set of other pathways Ssa is the subset of set Ss of source nuclides that are actinides The reason for the split in Eq is that the pathways arising from the fission of source actinide a are treated as correlated and other pathways are treated as uncorrelated The formula used to compute A is AM Y ARNG DS NL Ap Nip 76 pe Se aESsa PESa CCFE Page 161 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual where Nip is the number of atoms of target t formed along path p to that target Aj is given by E y A SRM Y Ea 77 ecSe TEST ec De where Se is the set of edges on pathway p Sp is the set of reactions on edge e R is the pulse averaged reaction rate of reaction r Re is the total pulse averaged reaction rate on edge e
252. terial that is irradiated by a time dependent projectile flux where the projectiles may be neutrons protons deuterons o particles or y rays The material is homogeneous infinite and infinitely dilute and the description of the evolution of the nuclide numbers is reduced to the stiff ode set of rate equations 27 x M olo N 7 where N number of nuclide i at time t pt projectile flux cm s7 for j At E n ESI A decay constant of nuclide j producing i s o reaction cross section for reactions on j producing i cm for j i total decay constant of nuclide j s 0 total cross section for reactions on j cm The processes described by Equation may be interpreted in terms of a directed graph with vertices corresponding to nuclides and edges giving the flow from parent to daughter nuclides via a decay process or an induced reaction Figure 6 schematically presents a fragment of this graph Graph theoretic methods are used to construct pathways see Section on page 158 The total flow out from vertex j by decay is equal to the total flow into other vertices a M A 5 X 8 izj Similarly the balances of the flows by projectile induced reactions give og of Yo 9 izj CCFE Page 133 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual d decay reaction j j A O i d d Figure 6 Directed graph representation of reactions and d
253. the charged particle deuteron proton alpha and gamma libraries up to 1 GeV B 2 2 Probability tables The LLNL 616 and CCFE 709 neutron induced cross section libraries are also pro vided with a set of probability tables that cover the resolved and unresolved resonance ranges of any evaluations that contain a resonance parameters file The CALENDF 2010 code provides those probability tables in the energy range from 0 1 eV up to the end of the unresolved resonance range B 2 3 Groupwise deuteron induced eaf d gxs The deuteron induced cross section library is available in two group structures that can be used as input to FisPACT II These are the VITAMIN J 211 and CCFE 142 formats the group boundaries of which are listed in Tables and Only flat weighting libraries are available which are suitable for most applications B 2 4 Groupwise proton induced eaf p gxs The proton induced cross section library is available in two group structures that can be used as input to FisPACT II These are the VITAMIN J 211 and CCFE 142 formats the group boundaries of which are listed in Tables and Only flat weighting libraries are available which are suitable for most applications B 2 5 Uncertainty eaf un A unique feature among activation libraries is the inclusion of an uncertainty file eaf un containing data for all neutron induced cross sections Reference 44 describes the uncertainty data for EAF 3 1 while reference describ
254. the US latest recom mended evaluated nuclear data file for use in nuclear science and technology applica tions and incorporates advances made in the five years since the release of ENDF B VILO including many new evaluation in the neutron sublibrary 423 in all and over 190 of these contain covariances new fission product yields for 31 isotopes and a greatly expanded decay data sublibrary for 3817 radionuclides For more details visit http www nndc bnl gov endf b7 1 E JENDL 4 0 Library Data The purpose of JENDL 4 0 is to provide a Japanese standard library for fast breeder reactors thermal reactors fusion neutronics and shielding calculations and other applications The data libraries used have been updated to the JENDL 4 0u level of August 2013 for both the neutron reaction and fission yields sublibrary JENDL FP Decay Data File 2011 70 contains decay data of 1284 FP nuclides of which 142 nuclides are stable that includes recent TAGS Total Absorption Gamma ray Spectroscopy information For more details visit http wwwndc jaea go jp jendl j40 j40 html F JEFF 3 2 Library Data The Joint Evaluated Fission and Fusion File is an evaluated library produced via an international collaboration of Data Bank member countries co ordinated by the JEFF Scientific Co ordination Group under the auspices of the NEA Data Bank The new JEFF 3 2 general purpose library has been released on March 5 2014 in ENDF 6 format and contains incid
255. the file eaf p asscfy 20100 connected to the stream asscfy contains these associations B 6 Radiological Data B 6 1 Biological hazard index eaf_haz Activity is one quantity used to judge the potential hazard of an irradiated material However activity takes no account of the biological impact on human beings To enable FISPACT II to give some indication of the potential biological hazard of irradiated materials a library of dose coefficients has been assembled which determine the dose received by a man over his lifetime 50 years following the ingestion or inhalation of 1 Bq of activity of a particular radionuclide The basic sources for these data are reports published by the ICRP 48 49 and the NRPB 51 However these sources primarily cover radionuclides generated by the CCFE Page 193 of 200 CCFE R 11 11 Issue 6 B EAF LIBRARY DATA FISPACT II User Manual fission power producing community and consequently only cover some of the nuclides that can arise in fusion applications In order to extend the range of nuclides to all those in the EAF decay library it has been necessary to use an approximate method Reference describes how available data for an element are used with decay data for a nuclide to derive Committed Effective Doses per unit uptake for ingestion and inhalation for the nuclides with no data In total 1209 nuclides have had data calculated approximately References document the eaf_haz library B 6 2 Legal tr
256. the files are read directly by FisPACT II without any further processing C 4 Probability Tables The CALENDF nuclear data processing system is used to convert the evaluation defin ing the cross sections in ENDF 6 format i e the resonance parameters both resolved and unresolved into forms useful for applications Those forms used to describe neu tron cross section fluctuations correspond to cross section probability tables based on Gauss quadratures and effective cross sections The CALENDF 2010 code pro vides those probability tables in the energy range from 0 1 eV up to the end of the resolved or the unresolved resonance range Probability table data in 709 or 616 group formats are provided for 2143 isotopes of the TENDL 2013 library These data are used to model dilution effects from channel isotopic or elemental interferences To account for Doppler broadening effects the tables are given at three temperatures 293 6 600 and 900 degree Kelvin C 5 Decay Data In addition to cross sections the other basic quantities required by an inventory code are information on the decay properties such as half life of all the nuclides considered These data are available in a handful of evaluated decay data libraries FISPACT II is able to read the data directly in ENDF 6 format it requires no pre processing to be done The now well verified and validated eaf_dec_2010 library based primarily on the JEFF 3 1 1 and JEF 2 2 radioactive deca
257. they cause to be printed in columns 8 11 of the inventory output table keyword description value units HAZARDS ingestion dose Aje Y Sv inhalation dose A ei Sv CLEAR clearance index A MtotL ATWO transport index A A2 C 2 HALF half life Ar loge 2 or Stable s See Appendix B 6 and References 13 25 for more details on the hazards clearance and transport A2 data 7 1 4 Inventory step summary The step summary appears after the table of values for individual nuclides The first line contains the number of nuclides N printed in the preceding table and the remaining lines give sums over nuclides of various diagnostic quantities The first three of these lines contain 1 the total activity in curies Nn TOTAL CURIES C3 A j l where C3 1 3 7 x 101 is the conversion factor from Bq to Ci CCFE Page 103 of 200 CCFE R 11 11 Issue 6 7 INTERPRETATION OF OUTPUT FisPACT Il User Manual 2 the total alpha power in Ci MeV Nn TOTAL ALPHA 10 9C3 M A Ea q 1 where the 10 is the conversion factor from eV to MeV 3 the total beta power in Ci MeV Nn TOTAL BETA 10 9C5 Y A Eg i 1 4 the total gamma power in Ci MeV Nn TOTAL GAMMA 10 9C3 V A E i 1 TOTAL NUMBER OF NUCLIDES PRINTED IN INVENTORY 87 TOTAL CURIES TOTAL ALPHA TOTAL BETA TOTAL GAMMA CURIE MeV CURIE MeV CURIE MeV 3 38027E 03 3 72537E 13 5 92195E 02 5 48191E 03 The next line splits the total ac
258. this keyword causes the bremsstrahlung dose rate of each specified nuclide to be printed at the end of each time interval An example of the use of this keyword is BREM 4 CL36 AR39 AR42 K42 In this case the bremsstrahlung contributions of C1 39 Ar Ar and K are calculated and output at the end of each time interval 5 2 4 CLEAR This keyword causes information on the clearance data of radionuclides to be input the calculations to include these data to be performed and the results for individual nuclides and summed clearance indices to be output at all timesteps CCFE Page 59 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual 5 2 5 CULTAB This keyword inserts additional lines at the beginning and end of the tab files so that the files can be processed more easily by other computer programs The data written are unchanged by the use of this keyword which is retained for consistency with earlier FISPACT versions 5 2 6 DENSITY densty This keyword enables the input of the density of the material undergoing irradiation The parameter densty should be given in units of gem If this keyword is used then the total activity will also be output in units of Cicm in addition to the standard output in Bqkg If FUEL is used to specify the input material for a run in which an inventory is calculated then the density must be specified An example of the use of this keyword is DENSITY 8 96
259. tial material being irradiated source nuclides and the target nuclides The target nuclides are those on the merged dominant nuclides list at the end of the irradiation If the LOOKAHEAD keyword is used then nuclides that appear on the merged dominant nuclide list at later steps in the cooling phase are added to the list The number of target nuclides included in the calculation may be altered by changing the value of topxx using the SORTDOMINANT keyword Source Nuclides Ti 46 Ti 47 Ti 48 Ti 50 Target Nuclides Sc 48 Sc 46 Sc 47 Ca 45 Ti 45 Ti 51 Ca 47 Sc 46m Ar 42 Sc 45m Sc 44 Sc 50m Ar 39 K 42 Sc 44m K 43 H 3 The pathways calculation prints lists of all significant paths and loops ordered by target nuclide The first line for each target nuclide gives the nuclide name and the percentage of the total number of atoms given by the number of significant paths shown The first line for each pathway identifies a path or loop gives its number and its respective percentage contribution to the target nuclide inventory The remainder of the line gives the nuclides on the path or loop from source to target and the type of graph edge joining them Edge types r R d D and b B respectively denote reaction decay and combined reaction and decay edges from short lower case and long lived upper case parents L and S denote short and long lived target nuclides Short lived nuclides have half lives less than the time interval and long lived hav
260. tivity into parts associated with o and y decays according to their decay type c f Table 10 on page 138 Activity from decays with type IRT 4 is assigned to the ALPHA BECQUERELS total activity from those with IRT 1 11 16 17 20 2 14 19 is assigned to the BETA BECQUERELS total and from those with IRT 3 to GAMMA BECQUERELS Activity from decays with IRT 12 or 13 is split between the o and totals and activity from decays with IRT 15 is split between the o and y totals Note that this definition of the split is different from that used in FISPACT 2007 I ALPHA BECQUERELS 5 153452E 22 BETA BECQUERELS 1 133988E 14 GAMMA BECQUERELS 1 167103E 13 TOTAL ACTIVITY FOR ALL MATERIALS 1 25070E 14 Bq 6 50837E 01 Ci cc DENSITY 1 93E 01 gm cc TOTAL ACTIVITY EXCLUDING TRITIUM 1 25070E 14 Bq 6 50835E 01 Ci cc TOTAL ALPHA HEAT PRODUCTION 20867E 18 kW TOTAL BETA HEAT PRODUCTION 51057E 03 kW TOTAL GAMMA HEAT PRODUCTION 24971E 02 kW TOTAL HEAT PRODUCTION 3 60076E 02 kW INITIAL TOTAL MASS OF MATERIAL 00000E 00 kg TOTAL HEAT EX TRITIUM 3 60076E 02 kW TOTAL MASS OF MATERIAL 00006E 00 kg NEUTRON FLUX DURING INTERVAL 27701E 14 n cm 2 s NUMBER OF FISSIONS 00000E 00 BURN UP OF ACTINIDES 0 00000E 00 INGESTION HAZARD FOR ALL MATERIALS 38528E 05 Sv kg INHALATION HAZARD FOR ALL MATERIALS 06441E 05 Sv kg INGESTION HAZARD EXCLUDING TRITIUM 38528E 05 Sv kg INHALATION HAZARD EXCLUDING TRITIU
261. to be applied CCFE Page 54 of 200 5 1 Library Data Preparation CCFE R 11 11 Issue 6 FISPACT II User Manual nprint is by default 0 in which case it prints the list of probability table data files and the nuclide mixture If it is set to 1 then it additionally prints total and partial cross sections and dilutions versus energy bin for all the nuclides to which self shielding is being applied The following is an example of a collapse run where probability table corrections are included for all the naturally occurring isotopes of titanium and tungsten in a mixture of titanium tungsten and iron GETXS 1 616 PROBTAB 1 0 SSFCHOOSE 2 0 Ti W FISPACT COLLAPSE EAF 616 FLT WITH PROBABILITY TABLE CORRECTIONS ASS 1 0 3 TI 85 0 W 10 0 Fe 5 0 END END OF RUN 5 1 19 SSFDILUTION nnuc nucname j num j grp i j dilution i j i 1 num j j 1 nnuc This keyword adds further user control to that provided by the SSFCHOOSE key word If SSFCHOOSE is used with argument nprint set to 1 then the computed dilutions versus energy bin are printed for each nuclide These dilutions are computed using the formulae given in Appendix A 4 3 If the user wishes to override these values then he may do so using the SSFDILUTION keyword The first argument nnuc lists the number of nuclides for which dilution values are to be specified For each nuclide j the nuclide name nucname j and the number of table entries num j are gi
262. ually absorbed by it In this sense the wall loading represents a convenient but not fundamental parameter The power carried by the neutron flux CCFE Page 86 of 200 5 3 Inventory Calculation Phase CCFE R 11 11 Issue 6 FisPACT II User Manual impinging upon the first wall is related to the 14 MeV neutron current not flux If one works out the heating power of 14 MeV neutrons it is found that a current C of 4 44 x 101 ncm s is equivalent to 1 MW m The relationship between 14 MeV neutron current and flux depends upon the source and first wall geometry and will vary from plant to plant 5 3 Inventory Calculation Phase This section of the input file is introduced by the first occurrence of the TIME key word that triggers the start of the solution of the inventory equations The inventory calculation has one or more irradiation steps optionally interleaved with cooling steps and terminated by the occurrence of the keyword ZERO optionally followed by one or more cooling steps 5 3 1 ATOMS This keyword starts the solution of the inventory equations over the time interval specified and causes the results isotopic elemental spectral to be output After the ZERO keyword it also causes pathways and uncertainty results to be output It is the standard method of producing output other options are SPECTRUM and RESULT The time step is set to zero after the completion of the output and so must be reset in subsequent steps usi
263. ucts when they undergo fission 5 2 13 FLUX fluz2 This keyword enables the total energy integrated projectile flux in cm s 1 to be specified for a particular time interval Note if several consecutive time intervals require the same flux value then it need be entered only once for these intervals Setting the total flux to zero gives a decay time step The flux must be set to a strictly positive value before the first irradiation step CCFE Page 63 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual The flux must be set to zero before using the keyword ZERO An example of the use of this keyword is FLUX 1 5E15 For the next time interval a total flux of 1 5 x 1015 ncm s will be used and this will also be used for subsequent time intervals until countermanded by a further FLUX keyword 5 2 14 FUEL n1 is j atoms j j 1 n1 This keyword allows the input of the number n1 of nuclides and the name is j and the number of atoms atoms j for each nuclide The name is specified using the format Te129m The specification of nuclides is essential if the materials to be irradiated do not have the natural isotopic abundance If different values are required then FUEL should be used The total mass of input material is calculated from the amounts of the nuclides input Note that FUEL and MASS must not both be used in a particular case If FUEL is used for a run in which an inventory i
264. ue mop abe RO ARCA IU S 94 5 3 25 SSFGEOMETRY es B320 SSEMASS y s adma a k a a e E a D a E R E a E 5 9 21 LABI 421 2x9 ox on e eek eee me ee Re R oec we Roos 5 39 28 LAB a risa a xw esp IS kc a peter dea od 0 9 20 TABS dida ara a S Rebus ue deer ua eode 3 5 39 90 TABA se ka morra C RUE SUE T S RR Ee ES A dal TIME e 5 39 92 WALL ai d eee uad m RL RB da a 5 9 99 CERO usc uso ee aa ie e e RR ERR s 5 4 Miscellaneous 97 5 4 1 comment gt gt 2 00 00 ee 97 6 Test Cases 97 7 Interpretation of Output 7 1 The Inventory Run output File 0 e 7 1 1 Header and run information 00 7 1 2 Table key ers 101 7 1 3 Time line and nuclide inventory o 102 7 1 4 Inventory stepsummary en 103 7 1 5 Elemental inventory o 108 7 16 Gamma spectrum ls 108 7 1 7 Dominant nuclides een 109 7 1 8 Bremsstrahlung correction 2222 109 7 1 9 Sensitivity output oe OR RR o ee Eon Ro C Ron 109 7 1 10 Uncertainty estimates e 111 7 1 11 Pathways a 50 404 al ad Ra He e m 7 1 12 Generic pathways uk ow ox XE e ae eae A 113 7 1 13 Run SUEDE dorus d AAA A o x RIA TR RE OU UA 113 7 2 The Inventory Run runlog File 2 208 114 7 3 The Printlib Run output File 0 000002 eae 118 T31 2 crc 118 7 3 2 Branching ratios 5 29 cx dom o RR oo RR RARE Ec eo 119 7 3 3
265. uns above CCFE Page 32 of 200 4 4 Library Summary Printing CCFE R 11 11 Issue 6 FISPACT II User Manual Figure 2 The files used by FISPACT II in the decay and fission data condense run example lt lt print library data summary gt gt NOHEADER lt lt read condensed cross section and collapsed decay data gt gt GETXS 0 GETDECAY 0 FISPACT PRINTLIB OF FW EEF lt lt print library data summary selection 0 gt gt PRINTLIB 0 END END OF PRINTLIB The keywords GETXS and GETDECAY with 0 arguments respectively instruct FISPACT II to read cross section and decay data from the binary files mapped to collapxi and arrayx See the definition of PRINTLIB on page 77 for details of the tables printed To generate print lib out type fispact print lib To convert the Fortran carriage control characters 0 and 1 to line feed and form feed in the output file and print the results to the default printer set for the a2ps command type fisprint print lib out CCFE Page 33 of 200 CCFE R 11 11 Issue 6 4 GETTING STARTED FisPACT Il User Manual 4 5 Inventory Calculation The example file for the inventory calculation is inventory i The first part reads in the collapsed and condensed data as for the print lib example NOHEAD MONITOR 1 GETXS 0 GETDECAY 0 FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 The MONITOR 1 keyword causes the input keywords to be echoed to the terminal for t
266. ure target and the target geometry could be a foil wire sphere or cylinder of finite height This model has been generalised further and applied to the mixture of nuclides required for a FISPACT II calculation The FIspact II user invokes this model of self shielding by using the SSFGEOME TRY keyword to define the type and dimensions of the target as detailed in Table 16 Table 16 The types of target geometry recognised by FISPACT II Identifier Type Dimension s Effective length y 1 foil thickness t y 1 5t 2 wire radius r y 2r 3 sphere radius r je 4 cylinder radius r height h y 1 65rh r h In more detail the initial form of the model 21 that accounts for the effect of a single resonance in a pure target containing a single nuclide defines a dimensionless parameter ry a tot Eres Y T 42 that depends on the physical length y the macroscopic cross section rot Eres at the energy Eres of the resonance peak the resonance width I for radiative capture and the total resonance width I Then the self shielding factor is A Az Gres 2 1 2 2 Ag 43 where the parameters defining this universal sigmoid curve are A 1 000 0 005 44 A 0 060 0 011 45 zo 2 70 0 09 46 47 p 0 82 0 02 These parameters were determined empirically by Martinho et al by fitting to a set of points generated by performing Monte Carlo simulations with the MCNP code
267. used in conjunction with the 616 energy group cross section data in the EAF library or the 709 group data in the TENDL 2013 library In the following discussion we use the term library or LIB to refer to either the EAF of TENDL 2013 cross section data as appropriate The dilution computed using the CALENDF data is applied either as scaling factors to the library cross section data or as replacements over the energy ranges for which the probability table data are available This is selected using the multzs argument to the PROBTABLE keyword If the CALENDF and library data were fully self consistent then the same self shielding would be obtained for both choices of multzs but the absence of elastic scattering cross section in the EAF data lead to some differences For either choice of multrs either partial or total scaling may be applied Scaling applied to LIB data Scaling is applied to the library data in one of two ways depending on the usepar argument to the PROBTABLE keyword see Section on page 52 If the partial self shielding scaling factor option is chosen then the cross section for nuclide p in energy group g and for MT value y belonging to the macro partial group x is scaled according to aya 0 ES 31 CCFE Page 145 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual and for the total scaling factor a y d a P y CES 32 god e The dilution d p g for a given nuclide p a
268. ut file An example of the use of this keyword is URS SPECTRUM URS SPECTRUM DPULSE LUX 0 0 E 1 0 HOURS SPECTRUM LUX 1 0E15 E 1 0 HOURS ATOMS At the end of the irradiation it is wished to include six hour long pulses Five of these are specified in the loop using SPECTRUM so that no detailed inventory is produced The final pulse the end of the irradiation has a detailed inventory since A TOMS is used 5 3 20 RESULT nresu sym i x i i 1 nresu This keyword is used when calculating pathways The pathway output includes the percentage of the total amount of the daughter nuclide produced by a particular path way One way to obtain this total amount is to perform an inventory run prior to the pathway calculation However it is much easier to be able to get the inventory from a separate run and then manually to use results from that inventory and input them into the pathway calculation nresu nuclides are specified and for each the identifier sym i e g Tel29m and the number of atoms z i are specified If ATOMS or SPECTRUM is not present then RESULT is necessary to start the pathway calculation and so must follow the keyword PATH or ROUTES An example of the use of this keyword is CCFE Page 93 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual RESULT 3 C14 1 356E19 N14 8 560E17 N15 7 568E12 The numbers of atoms of 14C 14N and N obtained from a standard
269. ven followed by a list of num j pairs of energy group indices grp and dilution cross section values dilution in barns In this example the dilution for 82W is set to 100 and 80 barns respectively in energy groups 300 and 301 and the dilution for 184W is set to 2 5 and 10 barns in groups 194 and 200 SSFDILUTION 2 W182 2 300 100 0 301 80 0 CCFE Page 55 of 200 CCFE R 11 11 Issue 6 5 CONTROL FILE KEYWORDS FISPACT II User Manual W184 2 194 2 5 200 10 0 5 1 20 SSFFUEL ni is j atoms j j 1 n1 This keyword allows the input of the number n1 of nuclides and the identifier s 7 and the number of atoms atoms j for each nuclide that is to be used in the self shielding calculation The identifier should be a nuclide name with the format of a chemical symbol followed by an atomic mass number e g W184 The specification of nuclides is essential if the materials specified do not have the natural isotopic abundance If different values are required then SSFFUEL should be used Note that SSFFUEL and SSFMASS must not both be used in a particular case An example of the use of this keyword is SSFFUEL 4 W182 1 34834187E 22 W183 7 27597094E 21 W184 1 55899050E 22 W186 1 44654079E 22 In this case tungsten with the 99 W isotope removed is to be used in the self shielding calculation The SSFFUEL keyword in this section applies to the collapse calculation initiated by the FISPACT keyword The keyword m
270. wed by the number of spontaneous fission neutrons per second and the number of neutrons from a n reactions The average energies for a 8 and y decays shown as ALPHA BETA and GAMMA in MeV and the y energy MeV in each of the 24 groups follow The independent fission yield 96 from each of the fissionable nuclides is then given At the end of this section details of the neutron spectrum used to weight the fission yields are given showing the fraction of the neutrons in different energy ranges see Appendix A 5 MAT NUMBER 1 2 3 4 5 6 ISOTOPE H 1 H 2 H 3 He 3 He 4 He 6 LAMBDA 0 000E 00 0 000E 00 1 781E 09 0 000E 00 0 000E 00 8 577 01 HALF LIFE KKKKKKKKK KEKKKKKKK 12 330 y KKKKKKKKK KKKKKKKKK 808 100ms SP FISS n s 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 a n n 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 lt ALPHA gt 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 lt BETA gt 0 000E 00 0 000E 00 5 707E 03 0 000E 00 0 000E 00 1 561E 00 lt GAMMA gt 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 5 644E 03 GAMMA GROUP 1 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 5 547E 06 GAMMA GROUP 2 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 1 591E 05 GAMMA GROUP 23 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 GAMMA GROUP 24 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 0 000E 00 Th232 FIS YIELD 1 751E 03 5 350E 04 7 633E 03 0 000E 00 1 107E 01
271. wing the final irradiation step The first section is terminated by the FISPACT keyword and this triggers the ex ecution of the library data preparation actions which have been queued prior to the occurrence of the FISPACT keyword The actions are queued in the correct order to ensure that any dependences between them are respected The initial conditions section of the input file is terminated by the first occurrence of the TIME keyword or exceptionally the END keyword for a run that does not involve any inventory calculations FISPACT II requires that all initial condition settings are declared before the inventory calculation is started and so there are more restrictions on the placing of keywords in the input file than in FISPACT 2007 Consequently some older files may need minor editing before they can be reused The final inventory calculation section of the input file is terminated by the END keyword any further content in the file is ignored FISPACT II attaches more signifi cance to the ZERO keyword than did FIsSPACT 2007 ZERO may now occur at most once and it triggers the calculation of pathways routes sensitivities and uncertainties The relevant keywords for each section of the control input file are presented in al phabetical order in the following three sub sections A further sub section describes miscellaneous input constructs Table 5 gives a complete list of the keywords together with the pages on which they are defined
272. y data libraries with additional data from the latest UK evaluations UKPADD6 10 contain 2233 nuclides However to handle the extension in incident particle type energy range and number of targets many more are needed A new 3873 nuclide decay library dec 2012 has been assembled from eaf dec 2010 complemented with all of JEFF 3 1 1 and a handful of ENDF B VII 1 decay files See Reference for more details There remain compatibility issues between the isomer definitions arising from the cross section library through the RIPL 3 database and the newly assembled decay library Historical incompatibilities in isomeric state number g m n o and energy levels between radionuclide daughter products of reactions and the associated decay data files will need to be addressed in a future release C 6 Radiological Data The radiological data for the increased number of nuclides present in the TENDL 2013 data are computed in the same manner as described for the EAF data see CCFE Page 199 of 200 CCFE R 11 11 Issue 6 F JEFF 3 2 LIBRARY DATA FISPACT II User Manual Appendix on page 193 The new hazards clearance and transport data are respectively for 3647 3873 and 3872 nuclides compared to 2006 2233 and 2233 for the EAF data For further details see Reference 67 D ENDF B VII 1 Library Data The Cross Section Evaluation Working Group CSEWG released the ENDF B VII 1 library on 22 December 2011 The ENDF B VII 1 library is
273. y sufficiently for them not to overlap significantly Note that TENDL 2012 uses a unique approach to create parameters for resolved sta tistical resonances for a large number of isotopes that did not have any This method invokes global average parameters from the different systematics and from the TALYS reaction code 35 These parameters are then used by either the CALENDF code or by the R matrix code AVEFTT Statistical resonance parameters are then obtained from zero up to the first excited level reflecting the average resonance parameters coming from compound model calculations Above the first inelastic level grouped inelastic cross sections with local fluctuations are obtained This method complements the measured resonance parameters or provides a resolved resonance range when mea surements do not existing In between these two cases statistical resonance parameters are adjusted to integral measurements when available This method which has been successfully applied to all isotopes living longer than one second has been used to populate resonance range of the TENDL 2012 libraries The cross section at a resonance peak is not supplied in the ENDF data The simple expression provided by Fr hner Eq 186 is used to supply this information CCFE Page 149 of 200 CCFE R 11 11 Issue 6 A THE MODEL FISPACT II User Manual Secondly Gres is made energy dependent by taking averages separately for each en ergy bin used for t

FISPACT-II User Manual - Culham Centre for Fusion Energy

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