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ALEPH 1.1.2 A Monte Carlo Burn-Up Code - SCK-CEN

1. 7 1 2 Improving Monte Carlo burn up o e 8 1 3 Iimerworkimes of ALEPH sussa Ed kon ala Caw fee won 9 1 3 1 Calculation flow and features of ALEPH 9 1 3 2 Nuclear data for burn up applications 12 2 ALEPH input options 13 2 1 The ALEPH inp tfile lA re DRE GS 13 2 2 End ALEPH input END keyword LL 000 13 2 3 Multi group cross sections and spectra 14 2 3 1 Group structure EGS keyword 14 2 3 2 Weight option GWS keyword 15 2 4 INTO datas ari SA n DALL VA PA 16 24 1 Data library DAT keyword e 16 2 4 2 Temperature TMP keyword on 20s aa aa o 17 2 4 3 Specifying nuclides MAT keyword 17 2 5 Calculating ORIGEN libraries 2 ico ade bia a ii 18 2 5 1 Running ORIGEN ORI keyword 18 2 5 2 Libraries LIB keyword q cus w te ei te se es e aye e e ares 18 2 6 Runing MENDIXJU sd etni ent aa ot dy elke dr die a Sas 18 2 6 1 MCNP material composition ABS keyword 18 2 6 2 MCNP tally specification TAL keyword 19 2 63 MCNP execution MCNP keyword 04 19 2 7 Specifying variable and burnable materials 20 2 7 1 Burnable materials BURN keyword 20 2 7 2 Volumes VOL keyword Mo 2 ee ak Y di dd n 20 2 73 Variable materials VAR keyword
2. 20 2 8 Specifying burn up History HIS keyword 21 2 8 1 Power irradiation IRP keyword reen 21 2 8 2 Flux irradiation IRF keyword 23 2 89 Decay lt DEC keyword us xar are te e sa E E a ena 24 2 8 4 Change temperature CHMT keyword 24 2 8 5 Change density CHMD keyword 24 2 8 6 Change material CHCM and CHBM keywords 24 2 8 7 Change TR card CHTR keyword 25 2 9 Optional KeyWords 4 pwa pit As e a 25 2 9 1 ORIGEN library title TIT keywords 25 2 9 2 Inputfile comment C and keywords 293 Output OUT Keyword vs cimas ko RRR a ao RE MCNP X input 2 1 The MCNP XJ input filee Ss ARES ER E ERIE E 3 2 Cell Specification a saca is a a R AA A ee 3 3 Surface Spe iicatloni 7 lt Cl A A AA o 3 4 Material specification Cass AAA A RS AAA ALEPH auxiliary files 4 1 The cross section directory file ooa aaa 4 2 Theisotopes lese ea 4 eg A E AE ea Error and Warning Messages 5 1 Reading the ALEPH input file LL 5 2 Reading the MCNP X input file 6 000000 5 3 Running ALEPH and MCNP X o e e ALEPH test problem and output 6 1 The NEA BUC single pin problem 004 6 2 Problem summary cs A A A RAS AS O 6 3 Burn p step Output sa Sat pt a ee ate 6 3 1 Calculation output Lilia a A ta RS A 6 3 2 Calculated cross section values iia e ir
3. 25 25 25 25 25 25 25 25 25 25 25 25 24 24 24 24 25 26 26 26 26 27 27 27 28 28 28 28 28 28 30 32 32 32 32 33 34 34 34 34 34 34 39 35 36 36 36 36 36 36 36 37 37 37 38 38 38 38 38 38 39 39 39 40 40 40 40 40 40 40 0500 0520 0530 0540 0550 0540 0560 0570 0580 0580 0581 0590 0580 0590 0600 0610 062 064 064 072 073 074 076 075 0740 0760 0770 0780 0800 0820 0790 0810 0780 0800 0820 0830 0840 0850 0860 0850 0860 0870 0840 0860 0870 0880 0890 0900 0890 0900 0910 0900 0910 0920 0930 0940 0950 0960 o o o zo o 5 ooo 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 79c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 54 410930 410940 410950 420920 420940 420950 420960 420970 420980 420990 421000 430990 440960 440980 440990 441000 441010 441020 441030 441040 441050 441060 451030 451050 461020 461040 461050 461060 461070 461080 461100 471070 471090 471110 481060 481100 481110 481120 481130 481140 481151 481160 491130 491150 501120 501140 501150 501160 501170 501180 501190 501200 501220 501230 501240 501250 501260 511210 09c 09c 09c 09c 09c 09c 09c
4. temperature specified for a previous cell ALEPH has found a cell containing a variable or burnable material with a certain density or temperature different from the temperature or density for a previous cell containing the same material Because ALEPH requires unique temperatures and densities the code cannot start the calculation Error density of material is different from the density previously specified in a cell When reading through the different materials ALEPH found a variable or burnable ma terial where the density specified in a style comment is different from the density found for cells containing the same material Error incomplete input for isotope in material The input for this isotope does not conform to the input specifications something must be missing check section 3 4 for the correct input Error nuclide specified for material has no libraries for temperature T eV When reading a material composition ALEPH didn t find a library for a certain isotope at this temperature Either the isotope wasn t included on the MAT keyword or the library number was set to 0 for this temperature Error material type does not exist use 1 for AP 2 for AC and 3 for FP There are only three different material types used by ORIGEN To identify these ALEPH uses the following three options activation products option 1 actinides and daughters option 2 and fission products option
5. 1 92357E 01 1 92442E 01 1 65372E 01 1 66953E 01 942410 7 00149E 02 7 14086E 02 9 50976E 02 7 47551E 02 942420 5 30620E 02 5 30168E 02 6 07010E 02 6 07028E 02 992540 0 00000E 00 0 00000E 00 1 20236E 14 1 21903E 16 992541 0 00000E 00 0 00000E 00 uh 3 07651E 16 0 00000E 00 992550 0 00000E 00 0 00000E 00 cut 3 75551E 17 3 01101E 31 6 4 4 Timing report The final part of the output file is the timing report It is a summary of the time that it took ALEPH to perform the calculation In this summary the MCNP X calculation time this is wall time the time required to read the spectra from the tally file the time to calculate the new libraries and the time to run ORIGEN is given for every point in the irradiation history All these times are given in seconds At the end of this summary ALEPH indicates the total calculation time in seconds minutes and hours ALEPH timing report Reading ALEPH and MCNP X input files 1 s ALEPH calculation time per point Calculate libraries 3 40000E 01 3 10000E 01 Running MCNPX 0 6 31600E 03 1 1 12870E 04 Reading spectra 49 50 2 49460E 04 3 10000E 01 Total calculation time 1 00824e 06 s Total calculation time 16804 min Total calculation time 280 066 h 47 Running ORIGEN 2 00000E 00 2 00000E 00 2 00000E 00 1 00000E 00 Total time 6 35200E 03 1 13200E 04 2 49790E 04 1 00000E 00 Appendix A Version
6. A temperature change to a temperature that was not specified on the TMP keyword has been requested using the CHT change keyword Since no library numbers are specified for this temperature ALEPH cannot perform the calculation Error on line density change requested for a burnable material Because the density of a burnable material is linked to the irradiation it is impossible to change the density of a burnable material Only densities of variable materials can be changed Error on line cell material change requested to a burnable material The CHCM keyword is used to change the material in a cell but only cells with variable materials can be changed and no burnable materials can be used Error on line non burnable material used in CHBM keyword The CHBM keyword is used to change burnable materials In this case ALEPH found a material number that is not a burnable material Error on line unknown IRF option Error on line unknown IRP option The constant flux irradiation IRF only allows two different options input of the abso lute number of source particles 1 or the absolute flux level of a material 2 The constant power irradiation IRP only allows four different options input of the power for every burnable material 1 the total power for all materials 2 the power of a particular material 3 or the total power of a subset of materials 4 Error on line illegal input
7. 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 79c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 55 511230 511240 511250 511260 521200 521220 521230 521240 521250 521260 521271 521280 521291 521300 521320 531270 531290 531300 531310 531350 541240 541260 541280 541290 541300 541310 541320 541330 541340 541350 541360 551330 551340 551350 551360 551370 561340 561350 561360 561370 561380 561400 571390 571400 581400 581410 581420 581430 581440 591410 591420 591430 601420 601430 601440 601450 601460 601470 4 4 4 4 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 79c 09c 79c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 56 601 601 611 611 611 611 611 621 621 621 621 621 621 621 621 621 631 631 631 631 631 631 631 641 641 641 641 641 641 651 651 661 661 661 661 661 671 681 681 1750 711 721 721 721 721 721 721 731 731 1820 741 741 1860 751 751 791 71 74 74 550 560 570 540 550 560 570 580 600 590 600 600 610 620 630 640 650 660 670 760 740 760 770 1780 790 800
8. 2 8 4 All of the power values used in these options have to be specified in MW When a single material is being burned all of the options described above are equivalent 2 8 2 Flux irradiation IRF keyword As was the case with the IRP keyword the IRF keyword is used for a substep of constant flux irradiation The syntax is similar to that of the IRP keyword IRF lt IRF option gt TU TIME There are two possibilities for the IRF renormalisation option lt IRF option gt e The absolute source strength S is specified 1 5 Because MCNP X provides us with flux values per source particle multiplied with the volume of the material the flux for every material j is calculated as Ng Y do l 1 Vj where o is the flux per source particle of group l for material j as calculated by MCNP X and where V is the total volume of material j present dj S 2 8 5 e The flux PHI of a specific material IK is given 2 IK PHI The flux of the other materials j is then calculated as 2 8 6 23 2 8 3 Decay DEC keyword Using the DEC keyword we can specify a sub step of natural decay The only input required is the time unit TU and the end time TIME of the sub step referenced from the start of the step DEC TU TIME 2 8 4 Change temperature CHMT keyword The CHMT keyword is used to assign a different temperature with value TEMP to the material with MAT as MCNP X material number CHMT MAT TEMP Thi
9. 810 820 830 840 850 870 970 832090 902300 09c 09c 09c 09c 79c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 57 902320 09c 912310 09 912330 09 922320 09 922330 09c 922340 09 922350 09 922360 09 922370 09 922380 09c 932370 09 932380 09c 932390 09c 942360 09c 942370 09c 942380 09c 942390 09c 942400 09 942410 09 942420 09 942430 09 942440 09c 952410 09 952420 09 952421 79c 952430 09 962410 09 962420 09c 962430 09c 962440 09c 962450 09c 962460 09c 962470 09c 962480 09c 972490 09 982490 09c 982500 09c 982510 09 982520 09 982530 09 992530 09c ESSAGE xsdir xsdir22 TITLE Benchmark BUC Phase IV B MOX fuels Case 5 first recycle MOX case A Pincell Pu U Pu 8 exterior worls 0 1 interior world 0 1 fuel pin fuel 900K 1 6 987923E 02 2 mannan nan on nn imp n 1 u 3 tmp 7 756E 08 58 c cladding 620K 2 2 3 885870E 02 2 3 imp n 1 u 3 tmp 5 343E 08 E moderator 575K 3 3 7 265121E 02 3 imp n 1 u 3 tmp 4 955E 08 a AA A ee a tele DIE RE aj a iwa em eta A tica i c surface cards c xl rpp 0 65635 0 65635 0 65635 0 65635 50 0 50 0 pitch 1 3127 2 cz 0 410 MOX pellet radi
10. input file will follow The details concerning this MCNP X input file are discussed in the following chapter chapter 3 Before starting we would like to make the following remarks concerning the input and the notations used in this manual e File and directory specification can be done using the rules under unix Specifying di rectories is done with the character and not the backslash character as is done under windows To specify a path from the root directory the directory must begin with a character For a subdirectory from the current working directory this not required It is also advisable to use directory names without white space e Anything appearing between lt gt is optional input This means that the user has either to choose between different options or that this is not required if certain conditions are met for instance the previous use of a certain keyword e Although all the keywords in this manual are in capital letters ALEPH will also accept keywords in small letters As such keywords like NSP or nsp are accepted but Nsp will not be accepted 2 2 End ALEPH input END keyword The END keyword is used to terminate the ALEPH input options and to indicate that the MCNP X input file will follow The MCNP X input starts either with MESSAGE to supply input directives that are not used in the command line or a problem title with or without the TITLE keyword If this keyword is not used ALEPH
11. n p and fission reaction rates can be used immediately but for the n y and n 2n reaction we also need branching ratios towards the ground state and the first metastable state And finally we also require accurate decay data for every nuclide considered The microscopic cross sections used in the Monte Carlo simulation must be the same as the ones used for the reaction rate calculation as is the case for traditional Monte Carlo burn up calculations This is why we have chosen to read the linearized microscopic cross sections used by ALEPH to calculate the multi group cross section from ENDF files generated by NJOY These ENDF files are the ones used by the ACER module of NJOY to prepare the MCNP X nuclear data files also known as ACE files To automatically generate these ENDF files and the corresponding ACE files for Monte Carlo transport calculations we are developing a utility called ALEPH DLG Data Library Genera tor 20 that wraps itself around NJOY This approach will allow us to quickly change our data when newer and better evaluations become available making ALEPH very flexible in its use of nuclear data ALEPH comes with its own data libraries for both ALEPH and MCNP X ALEPH LIB This library consists of 5 major evaluations JEF 2 2 JEFF 3 0 JEFF 3 1 JENDL 3 3 and ENDF B VI 8 at 6 temperatures 300 600 900 1200 1500 and 1800 K The atomic densities used by MCNP X are also updated for over 3000 nuclides by using t
12. 1 1 Except for some minor modifications to ORIGEN 2 2 to improve output accuracy the number of significant digits were increased from 3 to 5 and to increase the memory allocation no changes have been made whatsoever to the source code of the programs involved ALEPH itself has been written in C using a highly modular design to allow for great flexibility Replacing for instance MCNP or MCNPX by another Monte Carlo code would be quite easy because of this modular design we would only have to replace the object responsible for the MCNP X ALEPH interface by a similar object for the new Monte Carlo code And the same applies to ORIGEN as well A great effort has also been made to optimize ALEPH for speed the transition to fully object oriented code was essential for this Table 1 2 Monte Carlo burn up codes implementation details Code Name Language Reaction Rates Time Step Inside MC Code Optimisation MOCUP C Yes No MC REBUS Yes No OCTOPUS MONTEBURNS Fortran Perl Yes No EVOLCODE MCB Fortran Yes MCWO UNIX Script Yes No MVP BURN Yes No BURNCAL Fortran Yes No MCODE C Yes No ALEPH C Yes No Yes MCNPX Fortran Yes No ALEPH calculates the reaction rates outside the Monte Carlo code but the possi bility to calculate them within the MC code will be added in the future for comparison purposes b This will be implemented in a future version MCNPX uses a mixed method Principal reaction such
13. 3 37 5 3 Error metastable option does not exist Because MCNP X uses the ZAID identification and ORIGEN the ZAMID identification ALEPH still requires the input of the metastable state which is either 0 or 1 Error density of material not found in comment When reading through the different materials ALEPH found a variable or burnable ma terial where the density has not been specified in a style comment ALEPH requires this density to function Error f4 neutron tally with number not found While going through the MCNP X input file ALEPH will look for the f4 tally specified with the TAL keyword In this case ALEPH has not found the tally The user should check if he indicated the correct tally number Error burnable material not found When ALEPH is done reading the MCNP X input file 1t will also check if all burnable material from the BURN keyword were found in the MCNP X input file In this case ALEPH has not found such a material The user should check the materials on the BURN keyword and the materials in the MCNP X input file Running ALEPH and MCNP X Warning no transport library specified for nuclide nuclide skipped Warning no library specified for nuclide nuclide skipped When preparing a new MCNP X input file ALEPH found an isotope that doesn t have a transport library associated to it MCNPX would be able to continue because MCNPX would generate cross sections using m
14. 3 00 7 91 2 45 243 Am 3 16 0 93 3 65 0 90 3 97 0 94 242Cm 3 88 0 89 3 84 1 29 4 07 1 87 28Cm 5 13 1 46 4 81 1 84 4 90 2 55 2 Cm 0 01 0 20 0 64 0 30 1 17 0 44 25Cm 1 59 2 05 1 11 0 24 0 92 0 11 Mo 0 04 0 02 0 11 0 18 0 21 0 37 Te 0 97 0 58 0 37 0 78 0 12 0 98 Ru 0 64 2 88 0 64 2 70 0 66 2 52 IRh 0 20 3 00 0 39 3 45 0 50 3 83 128 Ag 0 41 3 05 0 38 3 73 0 66 4 46 133Cs 0 08 1 40 0 33 1 50 0 45 1 62 143Ng 0 37 1 35 0 18 1 29 0 23 1 30 Nd 0 36 0 48 0 47 0 25 0 53 0 03 Sm 0 50 0 62 0 79 0 39 1 02 0 30 Sm 18 12 11 34 1 51 6 52 2 47 7 44 150Sm 1 02 2 71 1 29 3 69 1 42 5 32 Bism 0 72 1 66 1 28 3 11 2 01 7 32 152Sm 1 43 1 39 1 68 3 07 1 93 5 69 I53Ey 0 87 0 21 1 22 0 80 1 59 22 22 B55Gd 3 58 16 75 0 20 10 30 5 88 3 42 To correctly take into account the rim effect we can also sub divide the fuel cell in several separate concentric fuel cells 41 BWIMS8A MALEPH 1 1 2 15 00 72 10 00960 L I A non Be rire nana ro ansiosa nina nani nono ini nede kenah L Mo IT TE T T F LI BLOOM e e A mm Relative difference to APOLLO2 E cid re f1 ___d0dau 000tte t i Am242m Figure 6 1 Relative differences given in of WIMS8A and ALEPH 1 1 2 to APOLLO2 using JEF 2 2 data after 1320 days of irradiation for the pin cell model The data used in the calculation
15. 7 3 VOL yes The volumes of the materials are being burned 2 7 2 no Comment after a keyword 2 9 2 This keyword must always precede the HIS keyword if IRP option 1 is used b If this keyword is not used the weight spectrum must be specified for every isotope in the MAT keyword If it is present it must always precede the MAT keyword Because the temperatures are linked to the MAT keyword it must always precede the MAT keyword 2 3 Multi group cross sections and spectra 2 3 1 Group structure EGS keyword The group structure to be used by MCNP X is specified using the EGS keyword EGS NG 1 EG 1 lt NG 2 EG 2 NG NGR EG NGR gt where NG i is the number of groups to be used between EG i 1 and EG i specified in MeV The energy corresponding to EG 0 has been set to le 11 MeV which is in most cases the lower energy boundary of cross sections Values for NG 1 and EG 1 are required To 14 specify complex group structures an unlimited number of values for NG i and EG i are allowed but the user should be aware of the fact that rounding problems can arise when using a group structure which is too fine The group structure itself is calculated using the constant lethargy approach the energy interval between EG i 1 and EG i is subdivided into NG i groups of constant lethargy The energy group values Ez between EG i 1 and EG i are therefore given by esta k EG i E EG i us exp NG 1 n
16. Development of ALEPH started roughly two years ago with Wim uttering the words Bernard je pourrais crire un petit program qui peut faire ca At the time we were formulating our approach to Monte Carlo burn up to solve the burn up problems that both of us were facing Little did we know that 20000 lines of C code later ALEPH would exceed all of our expectations Initially the code was just a simple post processor to MCNP X to calculate an ORIGEN library using cross section linearized by PREPRO 2002 The code didn t even have a name back then Very quickly we added the possibility to run ORIGEN from within this program so that it could calculate the resulting MCNP X material composition While this version which we had called ALEPH SPECTRUM was a big step forward it still required a lot of copy paste work to perform a full burn up calculation Encouraged by our initial success we set out to fully automate the entire process that is run MCNP X read the resulting spectra calculate the libraries run ORIGEN By the end of June 2004 the real first version of ALEPH was completed It was this version that was used in the VALMOX project and for the the burn up calculations in the MYRRHA DRAFT 2 pre design file Around that time we also decided to replace PREPRO 2002 by NJOY so that ALEPH and MCNP X would use exactly the same nuclear data This resulted in the creation of ALEPH DLG Data Library Generator to prepare the c
17. Either way ALEPH will not start the calculation Warning ENDF file for material specified in xsdiraleph does not exist No changes will be made to the ORIGEN library for this nuclide ALEPH checks the alephxsdir file for every isotope and library number other than zero on the MAT keyword to see if the ENDF file exists In this case ALEPH didn t find the ENDF file pointed to by the xsdiraleph file for this isotope and temperature This warning could indicate a badly specified data path in the DAT keyword a wrong entry in the xsdiraleph file or simply that the ENDF file is missing Because this is not an error per se ALEPH will continue as normal but no changes will be made to the ORIGEN library for this particular isotope and temperature Error irradiation history input before TMP and BURN keywords The irradiation history requires some input from the TMP and BURN keywords namely the temperature for use in the CHT keyword and the number of burnable materials that are being burned for IRP option 1 35 Error on line change requested without spectral recalculation This error message indicates that a change keyword CHD CHT CHCM CHBM and CHTR has been requested for a point where the spectrum is not being recalculated Because these change keywords can radically change the spectra in the system a spectrum calculation 1s required Error on line temperature has no cross sections associated to it
18. LI 2 3 1 where k 1 to NG 1 1 2 3 2 Weight option GWS keyword In the multi group ALEPH approach reaction rates 6 are calculated as follows L 070 Eh 8 2 3 2 in which Og and are the cross section and spectrum of group g with boundaries E 1 and Eg The group cross section 0 itself is calculated analytically by ALEPH using the following formula 0 2 3 3 with o E the energy dependent microscopic cross section and p E the spectrum used to weigh the cross section ALEPH forsees in two possible weight spectra The first being the constant weight spectrum the weight is constant over the entire energy range and the second one being the non self shielded PWR spectrum consisting of a Maxwellian spectrum a slowing down spectrum and a fission spectrum 22 Eexp lt E exp at AT Emax th max th kT 1 Q E E Emanx th EE Emax epi 2 3 4 VE exp BE Emax epi lt E 3 2 3Emax epi E marth exp 2E fis where max n is the upper energy boundary of the thermal region kT is the thermal energy Emax epi is the upper energy boundary of the epithermal region where E ji is the mean energy of a fission neutron The GWS keyword is used to specify this weight spectrum The syntax of this keyword is simply GWS lt weightoption gt 15 where lt weightoption gt is one of the following possibilities e For the constant weight spectrum 1 e For the n
19. NT the number of temperatures to be used At least one temperature value must be speci fied this will be the default temperature 2 4 3 Specifving nuclides MAT keyword One of the most important keywords in the input file is the MAT keyword used to specify the isotopes for which reaction rates have to be recalculated and which isotopes MCNP X can use for transport purposes Depending on the previous input of keywords the syntax for the MAT keyword is MAT ZAMID j LIB 1 lt LIB 2 LIB NT gt lt weightoption gt where ZAMID 3 similar to the ZAID identification used by MCNP X is the ORIGEN identi fication number of the isotope j using the element number Z the atomic mass number A and the metastable state M which is either 0 or 1 ZAMID 100002 10A M 10ZAID M 2 4 1 LIB i is the MCNP X and ALEPH library number between 1 and 99 corresponding with temperature T i specified in the TMP keyword In other words the number of library num bers given here must be the same as the number of temperatures NT specified using the TMP keyword and the order in which the libraries are specified must also be the same as the order of temperatures in the TMP keyword The library number LIB i can be specified as a simple integer for instance 15 or with a suffic c the MCNP X library notation for instance 15c The library numbers themselves can be positive negative or zero A positive library number indicates a library that
20. Prael H Lichtenstein User Guide of LCS The LAHET Code System MS B226 Los Alamos National Laboratory USA 1989 20 W Haeck B Verboomen ALEPH DLG 1 1 0 Creating Cross Section Libraries for MCNP X and ALEPH BLG 1002 Rev 0 Studiecentrum voor Kernenergie Centre d Etude de l Energie Nucl aire Mol Belgium 2005 21 G Audi O Bersillon J Blachot A H Wapstra The NUBASE evaluation of nuclear and decay properties Nuclear Physics A 729 p 3 128 2003 22 NEA JANIS User s Guide Nuclear Energy Agency Organisation for Economic Co operation and Development Paris France 2005 23 D E Cullen PREPRO 2004 ENDF B Pre processing Codes IAEA NDS 39 Rev 12 International Atomic Energy Agency Austria 2004 24 G J O Connor P H Liem Burn up Credit Criticality Benchmark Phase IV B Results and Analysis of MOX Fuel Depletion Calculations Nuclear Energy Agency Organisa tion for Economic Cooperation and Development Paris France 2003 25 WIMS The ANSWERS Software Package A General Purpose Neutronics Code Serco Assurance 1999 26 S Loubiere R Sanchez M Coste A Hebert Z Stankovski I Zmijarevic APOLLO2 Twelve Years After Proc Int Conf on Math and Comp M amp C99 Madrid Spain September 1999 27 W Haeck B Verboomen ALEPH An optimum approach to Monte Carlo burn up submitted for publication in Nuclear Science and Eng
21. but it is not entirely clear As can be seen in table 1 2 ALEPH is one of the first MC burn up codes to use a more efficient approach to reaction rate calculation It is also one of the first MC burn up codes that is capable of using MCNPX with all of its functionality For ADS systems such as MYRRHA the use of MCNPX is of crucial importance due to the necessity of high energy physics for the high energy reactions in the spallation target and because the coupled neutron proton transport in the system can then be treated in a single code without making any approximation 16 Some of the MC burn up codes mentioned in table 1 1 are being or have been used to calculate burn up in ADS systems notably MCB 17 and EVOLCODE 18 In these cases the external proton source has to be calculated with a separate code for instance LAHET 19 before MCNPX was created or even MCNPX itself Depending upon the size of the spallation target the primary external neutron source will also change due to burn up of the assemblies around the spallation target To take into account the change in this primary source the source would also have to be recalculated every burn up step This would be done by default if MCNPX is used in the burn up code 1 3 Inner workings of ALEPH 1 3 1 Calculation flow and features of ALEPH ALEPH is in essence an interface code between NJOY 99 90 12 ORIGEN 2 2 14 and any version of MCNP 13 or MCNPX 1 as can be seen in figure
22. from 300 to 400 nuclides while ALEPH requires atomic mass values for at least every nuclide possible in ORIGEN We have therefore decided to update those atomic mass values by using the Atomic Mass Evalua tion 2003 included into NUBASE 21 from the Atomic Mass Data Center These values are also specified with the isotopes file and they are also included in the xsdir files provided with ALEPH LIB Although the user should never need to worry about the isotopes file it is included in the libraries delivered with ALEPH the general structure of the file is detailed here should the 30 need arise For an element with an element name EL a proton number Z and NI different isotopes the entry looks like this EL Z ZAID 0 MASS 0 ZAID 1 MASS 1 ZAID NI MASS NI where ZAID 0 and MASS 0 are the ZAID identification of the element 2 1000 and the atomic mass of the element expressed in units of neutron mass as is the case in MCNP X ZAID i and MASS i for 1 lt i lt NI are the ZAID identification and the atomic mass in units of neutron mass of isotope i We do not distinguish the metastable state of nuclides in this approach because a nuclide is supposed to have the same atomic mass regardless of the metastable state 31 Chapter 5 Error and Warning Messages 5 1 Reading the ALEPH input file Error input file not found ALEPH tried to open the input file that was specified on the command line but cou
23. have reached a a final burn up of 48 GWd tHM As a test problem we consider the pin cell problem For the calculation in ALEPH we have subdivided every cycle into time steps of 26 25 days every step thus accounts for an average burn up of 1 GWd tHM for a total of 51 burn up steps 48 constant power steps and 3 decay 40 steps We have burned the fuel in the model as a single cell For every burn up step we have run a criticality calculation of 280 cycles 30 inactive and 250 active cycles with 20000 neutron histories per cycle The fractional absorption criterion has been set to 99 99 The average error on the total flux is 0 02 for The complete ALEPH input file for this problem can be found in appendix B Table 6 1 Relative differences given in of WIMS8A and ALEPH 1 1 2 to APOLLO2 using JEF 2 2 data after 420 870 and 1320 days of irradiation for the pin cell model Nuclide EOCI 420 days EOC2 870 days EOC3 1320 days WIMS ALEPH WIMS ALEPH WIMS ALEPH 2347 1 14 0 21 2 13 0 43 3 16 0 61 2351 0 22 0 20 0 48 0 26 0 86 0 21 2201 0 96 1 02 0 90 0 83 0 84 0 80 2380 0 02 0 01 0 04 0 02 0 10 0 02 238 Py 0 13 0 24 0 46 0 55 1 01 1 01 239 py 0 88 0 40 2 09 0 83 3 73 0 96 240pu 1 04 0 22 1 87 0 26 2 60 0 03 241 py 1 84 0 65 2 78 1 37 3 40 1 89 242 Du 0 55 0 44 1 13 0 70 1 71 0 66 23INp 8 83 14 38 8 94 1328 9 13 12 26 24lAm 1 06 0 31 2 12 0 81 3 16 1 17 242mAm 10 63 3 53 9 37
24. history ALEPH version 1 0 0 January 2005 e Initial release ALEPH version 1 0 1 May 2005 e Minor fixes some input options were rewritten and new ones were added to simplify the ALEPH input and especially the MCNP X input file The keywords MCN for the MCNP X executable and MMC for the fractional absorption criterion were replaced by MCNP and ABS This is just a cosmetic change the meaning of the keywords didn t change The keywords APL to specify the activation products library number ACL to specify the actinide library number FPL to specify the fission products library number and PHL to specify the photon library were merged into a single keyword LIB This keyword is also used to specify the name of the ORIGEN library file and the name of the decay library file The ORIGEN library name no longer needs to be specified on the ALEPH command line as before The keyword ORI is now used to specify the path and executable name of ORIGEN 2 2 similar to the MCNP keyword for the MCNP X executable but without the parallel options of course The DAT keyword is now used to specify the data path to the ENDF files used by ALEPH This allows for a greater flexibility for the nuclear data compared to the now obsolete option based keyword 1 for JEF 2 2 2 for JEFF 3 0 The xsdiraleph file has been introduced to specify the individual ENDF files for every isotope and library number on the MAT keyword The temperatures on the temper
25. in two by using the c comment If the user still wants to comment out entire lines in an entry he should use the style comment Whenever the entry ends c style comments are allowed ALEPH also requires the temperature of the cell For this purpose every cell containing a variable or burnable material must have the option tmp within its declaration If this temper ature specification is missing the default temperature will be used the first entry on the TMP keyword see section 2 4 2 27 3 3 Surface specification MCNP X distinguishes between three types of surfaces a normal surface designated by a simple integer a reflective boundary surface designated by a simple integer preceded by an and a white boundary surface designated by a simple integer preceded by an ALEPH will read this surface number along with the surface transform number that follows the surface number if a surface transform is present of course All the previous rules for cell specification still apply the surface number and the surface transformation number if any is given may not be interrupted by comment the surface entry may not be interrupted by c style comment and the line continuation card is not allowed the first 5 blank characters on a line indicate a continuation of the previous line 3 4 Material specification The initial material compositions of burnable materials and the composition of the variable materials to be used in the ALEPH
26. run are read from the material specification of MCNP X A variable or burnable material is specified according to the following rules First the first line of the material specification of a material lt matnr gt may not contain any isotope and it must be ended by a style comment containing the density lt density gt as it was given in the cell entries if the material is in use or the density foreseen by the user when it will be used m lt matnr gt lt density gt The different isotopes that compose the material have to be specified on the following lines using this syntax ZAID lt LIB gt FRAC lt origen type gt lt metastable gt where ZAID is the MCNP X material ID see equation 2 4 1 and FRAC is either the weight fraction in MCNP X this must be a negative number or the atom fraction this is a positive number of this isotope in the material The library number LIB is optional Library numbers that are specified here will be skipped as only the numbers given in the MAT keyword will be used To perform ORIGEN calculations some additional information on the isotope are still required the ORIGEN material type lt origen type gt and the metastable state lt metastable gt All this is specified after the weight fraction within a comment ORIGEN distinguishes be tween 3 different types of materials activation products actinides and fission products If the isotope is an activation product for instance 0
27. surfaces and material compositions and take whatever information it needs Apart from these points ALEPH will also look for the tally specified with the TAL keyword to see if it is present ALEPH will also determine if the calculation is a criticality calculation or a fixed source calculation In the case of a criticality calculation ALEPH will look for a source file called srctp the default MCNP source file name If that file is present the source specification using the ksrc or sdef keywords will be commented out to force MCNP X to use the source file generated by a previous run ALEPH will also read the value of the effective multiplication factor ke from a previous run to use it as a ke estimate for the next run This is done by default 3 2 Cell specification For the specification of a cell constructions using the like but scheme are not permitted This means that the cell number must be followed by the material number and that the third number must be the density provided that the material number was not 0 The density itself may be specified in g cm or in atoms barn cm as per normal MCNP X The input of these 3 cell parameters or 2 when it concerns a void cell may not be interrupted by comments either c or The line continuation card amp is also not allowed because ALEPH uses the first 5 blank characters on a line to identify entries on multiple lines Due to this it is also not allowed to cut an entry
28. then lt origen type gt 1 For actinides for instance 35U this is lt origen type gt 2 and for fission products it is lt origen type gt 3 ORIGEN also distinguishes between isotopes in the ground state and in a metastable state something which is not provided in MCNP X If the isotope is in the ground state then lt metastable gt 0 while it will be lt metastable gt 1 for metastable nuclides ALEPH can read both types of material specification possible in MCNP X weight fractions and atomic fractions The same counts for the densities specified on the cell entries it may be specified in g cm or in atoms barn cm as per normal MCNP X And it is also allowed for the input to be mixed density in g em and composition in atomic fractions and vice versa By default ALEPH will recalculate variable and burnable materials to densities given in atoms barn em and compositions given in atom fractions as MCNP X does internally 28 To recalculate compositions and densities from one formalism to another ALEPH requires precise values of the atomic mass of every nuclide And this is where the isotopes files comes in see section 4 2 29 Chapter 4 ALEPH auxiliary files 4 1 The cross section directory file The cross section directory file alephxsdir that has to specified in the DAT keyword is a simple text file that is used to specify the file names of the ENDF files for use with ALEPH With the
29. w s 992530 1 25686E 02 ai 4 10005E 01 6 4 Final output 6 4 1 Burn up history When the entire calculation is done ALEPH will print out the final output This starts with a detailed overview of the final irradiation history used by ALEPH This overview will indi cate material changes both to burnable and variable materials geometry changes This overview will also give the power or flux levels as calculated by ALEPH and used by ORIGEN 2 2 for every burnable material that was being burned 45 For our single pin problem this summary looks like this Burn up history overview Step 1 IRP 0 000350955 MW cm3 26 25 d Step 16 IRP 0 000350955 MW cm3 26 25 d Step 17 DEC 30 d Step 18 IRP 0 000350955 MW cm3 26 25 d Step 50 IRP 0 000350955 MW cm3 26 25 d Step 51 DEC 5 y 6 4 2 Accumulated burn up An important quantitv in depletion calculations is the burn up accumulated bv a material during the irradiation Burn up is usually expressed as GWd ton initial heavy metal or MWd kg initial heavv metal We prefer to use this last unit So in the case of an irradiation step with constant power P the burn up BU accumulated by material j will be given by 6 Pit Pa jVi BU 10 6 4 1 where pa is the density in g cm of the actinides initially present in the material and t is the irradiation period in days V is the volume of the cells containing the material j For a step with constant fl
30. will be used for reaction rates calculation and for transport calculations Negative library numbers are used when the library is only to be used for reaction rates calcu lation This way cross section files that are not suitable for transport calculations for instance the EAF 99 neutron activation files can still be used to calculate reaction rates Using zero as library number will cause the nuclide to be skipped in reaction rates calculations and transport 17 calculations for this temperature This can be used when a certain temperature is only required by a variable material and not a burnable material lt weightoption gt is optional input and has only to be included if the weight option keyword GWS 1s not used This option can be used to specify different spectral weight options for certain nuclides The input of this option itself is the same as the input for the GWS keyword The user should note that using this when the GWS keyword is present will lead to an error message On the other hand placing the GWS keyword after MAT will also lead to errors It is advised to make this isotope list as complete as possible Because ALEPH must ensure that MCNP X runs without any problem only materials with positive library numbers that are in this list will be used in MCNP X runs If somehow a nuclide is created that is not in this list 1t will be omitted and a warning will be printed Also only isotopes and no natural elements may be used f
31. will continue to read the MCNP X input file as if it was regular ALEPH input causing the code to terminate with errors 13 Table 2 1 Overview of the ALEPH input keywords Keyword Required Keyword Use Section ABS yes The fractional absorption criterion to be used 2 6 1 BURN yes The Burnable materials used in the problem 2 7 1 no Comment line 2 9 2 CHBM Change a burnable material 2 8 6 CHCM Change the material of a cell variable material only 2 8 6 CHMD Density change of a variable material 2 8 5 CHMT Temperature change of a burnable or variable material 2 8 4 CHTR Change the TR card on a surface 2 8 7 DAT yes The data path and the name of the xsdir file 2 4 1 DEC A decay sub step 2 8 3 EGS yes The group structure to be used 2 3 1 END yes Ends the ALEPH input options 22 GWS no The weight spectrum to be used for all isotopes 232 HIS yes The irradiation history 2 8 IRF An irradiation sub step of constant flux 2 8 2 IRP An irradiation sub step of constant power 2 8 1 LIB yes The ORIGEN libraries and library numbers 2 5 2 MAT yes The isotopes that have to be used 2 4 3 MCNP yes The MCNP X executable and the calculation type 2 6 3 ORI yes The ORIGEN 2 2 executable 2 5 1 OUT no ALEPH output options 2 9 3 STP An irradiation step 2 8 TAL yes The tally number 2 6 2 TET no A title for the new ORIGEN libraries 2 9 1 TMP yes The temperatures used in this ALEPH run 2 4 2 VAR no The variable materials used in the problem 2
32. 0 materials currently not present 0 materials currently undergoing decay Material 1 currently being burned Cells 1 Volume 52 810 cm3 Temperature 7 756E 08 eV Density 10 450 g cm3 Composition 8016 1 23735E 00 g cm 92234 1 00860E 04 g cm 92235 2 11885E 02 g cm 92238 8 45426E 00 g cm 94238 1 84248E 02 g cm 94239 4 03154E 01 g cm 94240 1 92357E 01 g cm 94241 7 00149E 02 g cm 94242 5 30620E 02 g cm WWWWW WWW Ww 6 3 Burn up step output 6 3 1 Calculation output For every burn up step ALEPH prints out a lot of information to the standard output screen If the user has set the appropriate option on the OUTkeyword ALEPH will also print all of that information to the output file First ALEPH will indicate for which points the spectra will be used in the example given below the spectrum is used for calculating the composition up to 43 point 1 After running MCNP X and reading the spectra ALEPH will start calculating the new ORIGEN libraries At this point ALEPH will write out data on every ENDF file used Because multiple temperatures are possible ALEPH will do this for every temperature used for burnable materials in this case 1t is only done for 900 K When the new libraries have been prepared ALEPH will run ORIGEN and accumulate the burn up of every material For the single pin model this part of the output file looks like this parts of the output were left out due to space restrictions Cal
33. L IRP 1 1 853398E 02 TEL IRP 1 1 853398E 02 TPL IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 TERA IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 TE 4 IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 JIP IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 TP 0 DEC 4 30 TP IRP 1 1 853398E 02 TE 1 IRP 1 1 853398E 02 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 52 DEC c the 10 10 10 20 30 30 40 501 501 701 701 801 801 901 110220 OMA WADA Di DD VI Kr 010 020 030 030 060 070 090 00 10 40 50 60 70 90 0230 0270 0310 0320 0330 0340 0360 0360 0380 0400 853398E 02 853398E 02 853398E 02 853398E 02 853398E 02 853398E 02 1 853398E 02 853398E 02 853398E 02 853398E 02 853398E 02 853398E 02 853398E 02 853398E 02 aterial list 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 09c 26 26 26 26 26 26 26 26 26 26 26 26 26 26 29 25
34. MAT keyword the user specified the library numbers for every isotope to be used in the calculation For every non zero library number lib both positive and negative in an entry of an isotope with a ZAMID identification on the MAT keyword the following line must be in the cross section directory file ZAMID lib endf file name where endf file name is the name along with subdirectories if required of the ENDF file containing the required information Please note that the common data path specified with the DAT keyword should not be included here As was the case for the MAT keyword the library number lib can be specified as a simple integer for instance 15 or with a suffic c the MCNP X library notation for instance 15c 4 2 The isotopes file Although ORIGEN uses the ZAMID identification for input of a nuclide it does not use this identification for the output There ORIGEN uses element names to identify nuclides The isotopes file is used to link element names with the appropriate Z number of the element To recalculate compositions both from the MCNP X input file and the ORIGEN output files to the standard atoms barn cm the correct atomic mass of every possible nuclide is re quired The atomic mass values used by ALEPH and by MCNP X itself must also be the same to ensure data consistency MCNP X uses the values that are specified in the first part of the xsdir file but the standard xsdir files only contain atomic mass data
35. October 7 10 2002 M Herman ENDF 102 ENDF 6 Data Formats and Procedures for the Evaluated Nu clear Data File ENDF VIT BNL NCS 44945 01 04 Rev Brookhaven National Labora tory USA 2005 R E MacFarlane D W Muir The NJOY Nuclear Data Processing System Version 91 LA 12470 M Los Alamos National Laboratory USA 1994 61 13 J F Briesmeister MCNP A General Monte Carlo N Particle Transport Code Version 4C LA 13709 M Los Alamos National Laboratory USA 2000 14 A G Croff A User s Manual for the ORIGEN2 Computer Code ORNL TM 7175 Oak Ridge National Laboratory Oak Ridge USA 1980 15 H Ait Abderrahim et al MYRRHA Pre Design File Draft 2 Report R 4234 Studiecentrum voor Kernenergie Centre d Etude de l Energie Nucl aire Belgium 2005 16 E Malambu T Aoust Strength and Weakness of MCNPX Experience Gained from MYRRHA ADS Calculations Monte Carlo 2005 Topical Meeting Chattanooga USA April 17 21 2005 17 K Tucek Neutronic and Burnup Studies of Accelerator Driven Systems Dedicated to Nuclear Waste Transmutation PhD Thesis Royal Institute of Technology Sweden 2004 18 E Gonzalez M Embid Segura A Perez Parra Transuranic Transmutation on Fertile and Inert Matrix Lead Bismuth Cooled ADS 6th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation Madrid Spain pp 207 218 2000 19 R E
36. RN and VAR keywords are used to specify lists of MCNP X material numbers to be used in the calcualtion These numbers cannot be zero or negative Error on line expected at least burnable materials but found only in the keyword BURN ALEPH tested the number of burnable materials found on the BURN keyword against the number of burnable materials that are being burned at any given time In this case there are less materials than indicated The user has probably forgotten to specify this number and that ALEPH has assumed that the first material number is the number of burnable materials that are being burned at any given time Error on line input of negative volumes with volume keyword VOL The VOL card is used to specify the volumes of the cells containing burnable materials It is obvious that these volumes cannot be zero or negative Error on line expected volumes but found volumes in the VOL keyword The number of entries on the VOL keyword must be the same as the number of burnable materials that are being burned at any given time the first entry on the BURN keyword 34 Either the first entry on the BURN keyword is wrong or the user has not specified all volumes Error on line no BURN keyword specified before VOL ALEPH has detected the use of the VOL keyword before the BURN keyword Because the number of volume entries on the VOL keyword depends upon the input on the BURN keyword the VOL keywor
37. STUDIECENTRUM VOOR KERNENERGIE CENTRE D TUDE DE L NERGIE NUCL AIRE OPEN REPORT SCK e CEN BLG 1003 Rev 0 ALEPH 1 1 2 A Monte Carlo Burn Up Code aleph Wim Haeck Bernard Verboomen January 2006 SCK e CEN Boeretang 200 2400 Mol Belgium OPEN REPORT OF THE BELGIAN NUCLEAR RESEARCH CENTRE SCK e CEN BLG 1003 Rev 0 ALEPH 1 1 2 A Monte Carlo Burn Up Code aleph Wim Haeck Bernard Verboomen January 2006 Status Unclassified ISSN 1379 2407 SCK e CEN Boeretang 200 2400 Mol Belgium SCK e CEN Belgian Nuclear Research Centre Boeretang 200 2400 Mol Belgium Phone 32 14332111 Fax 321431 5021 http www sckcen be Contact Knowledge Centre library Osckcen be RESTRICTED All property rights and copyright are reserved Any communication or reproduction of this document and any communication or use of its content without explicit authorization is prohibited Any infringement to this rule is illegal and entitles to claim damages from the infringer without prejudice to any other right in case of granting a patent or registration in the field of intellectual property SCK e CEN Studiecentrum voor Kernenergie Centre d Etude de 1 Energie Nucl aire Stichting van Openbaar Nut Fondation d Utilit Publique Foundation of Public Utility Registered Office Avenue Herrmann Debroux 40 B 1160 Brussel Operational Office Boeretang 200 2400 Mol Belgium Foreword
38. a 6 4 A A 6 4 1 Burn up history cc ie OR eee wes 6 4 2 Accumulated burn up e noon 6 4 3 Material composition LL GAA Timing report css LELE A aa Version history NEA BUC MOX pin cell input Chapter 1 Introduction to ALEPH 1 1 The Monte Carlo method and burn up applications In the last 40 years Monte Carlo particle transport has been applied to a multitude of prob lems such as shielding and medical applications to various types of nuclear reactors The success of the Monte Carlo method is mainly based on its broad application area on its ability to handle nuclear data not only in its most basic but also most complex form namely con tinuous energy cross sections complex interaction laws detailed energy angle correlations multi particle physics on its capability of modeling geometries from simple 1D to com plex 3D There is also a current trend in Monte Carlo applications toward high detail 3D calculations for instance voxel based medical applications something for which deterministic codes are neither suited nor performant as to computational time and precision Apart from all these fields where Monte Carlo particle transport has been applied successfully there is at least one area where Monte Carlo has had limited success namely burn up and activation calculations where the time parameter is added to the problem The concept of Monte Carlo burn up consists of coupling a Monte Carl
39. able names volumes densities Other essential information like the initial material composition temperatures are read from the initial MCNP X input file itself When the input file has been processed ALEPH will build a new MCNP X input file based upon the input options of the user and start an MCNP X calculation The neutron spectra for the different burn up zones considered are then passed on for building the new ORIGEN libraries and ORIGEN input files one for every zone with burnable materials The MCNP X output file is also processed to determine if MCNP X hasn t encountered any problems In the case of criticality calculations ALEPH also reads the value of the effective multiplication factor which will be used as an initial estimate for the following calculation In that case the fission 10 ALEPH Build MCNP input file MCNP X ACE file ALEPH DLG Figure 1 1 Calculation flow inside ALEPH and ALEPH DLG source calculated in the previous calculation will also be used in the following calculation ALEPH is capable of using all irradiation features of ORIGEN It provides for constant power irradiation constant flux irradiation and simple decay ALEPH will use a relative power dis tribution to determine the absolute levels of flux or power in every zone After the ORIGEN calculation ALEPH reads the results and cleans up all the temporary fi
40. as n y n 2n n 3n n p n and fission are calculated by MCNPX using multiplier bins A 63 group flux is used to take into account the other reactions using 63 group cross section An argument against Monte Carlo burn up codes that is often used is their complexity because most codes use a script or link approach to the problem As a result the user would have to understand and manage a large number of input and output files while the conversion of data from one form into another would introduce approximate results due to successive round off ALEPH strives to solve this problem either true or conceived as well ALEPH is indeed an interface code but it actually wraps itself around the codes involved and automates the entire process The ORIGEN input files for instance are created by ALEPH itself without any inter vention of the user so that the use of ORIGEN is actually hidden from the user By providing an easy to understand user interface we also take away the burden from the user For the user it is as if he is running a simple MCNP X problem with some extra options A typical ALEPH calculation starts by the processing of the input file the black arrow at the top of figure 1 1 The input itself consists of the ALEPH code options along with an initial MCNP X input file These ALEPH code options are the irradiation history the group structure to be used the materials and libraries to be used the ORIGEN and MCNP X execut
41. ature keyword TMP have now to be specified in eV and not in K as before This was changed because temperatures are specified in eV in MCNP X The NGR keyword used to specify the number of groups in the group structure has been replaced by the EGS keyword for specifying the entire group structure using a con stant lethargy approach The group structure does no longer have to be included in the MCNP X input file which shortens the input file considerably The ERG keyword to specify the begin energy of the group structure is now obsolete this has been set to 1 1071 MeV by default 48 The NSP keyword that was used to specify the number of burnable material is now re placed by the BURN keyword The BURN keyword is used to specify the number of mate rials that are burned at any given time along with all the burnable material numbers A similar keyword called VAR has to be used to declare the variable material numbers The new keyword VOL has now to be used to specify the volumes of the cells containing burnable materials using the same order as on the BURN keyword These volumes no longer have to be specified in comment lines in the MCNP X input file Due to the previous two changes the comment labels VARCELL BURNCELL VARMAT and BURNMAT along with the required paramemters volumes index no longer have to used Only the density of burnable and variable materials has still to be specified using a style comment in the MCNP X inp
42. be non negative and specified in eV 32 Warning temperature above 2500 K detected in temperature keyword TMP Temperature input on the TMP keyword must be specified in eV To be sure that the user uses realistic temperature values ALEPH will recalculate the temperature to K and test 1f the temperature is smaller than 2500 K which is a reasonable upper limit for realistic temperatures This is just a warning message for the user to detect erroneous input Error on line illegal input negative or zero for the library In the ORI keyword the user has to specify among others the original ORIGEN library numbers for the activation products actinides and fission products These numbers must be non negative and non zero integers Error on line ORIGEN library found in file Expected library or While reading the ORIGEN library specified on the ORI keyword ALEPH found an ORIGEN library number different from the three numbers specified on the ORI keyword Either a wrong ORIGEN library file was used or there is a formatting error in the file Error ORIGEN library file not found ORIGEN requires three library files to function the cross section libraries the decay library and the photon library If ALEPH cannot find one of these files the code will issue this error Error on line parallel option on keyword MCNP is either 1 or 2 ALEPH foresees two modes for running MCNP X serial and pa
43. cell in which the new variable material with number NEWMAT will be used The temperature of the new material will be set to that of the old material unless the new material is already in use Because every material can only be associated with a single temperature the temperature that is already in use will be assigned to the cell It is up to the user to make sure that variable materials are not assigned to cells with different temperatures For burnable materials it is only allowed to exchange on old burnable material OLDMAT with a new burnable material NEWMAT with the CHBM keyword CHBM OLDMAT NEWMAT 24 It is not possible to change specific cells because that would involve changing the tally used to calculate the spectra This might be added in a future version but for now it is not possible If one of both materials is not being burned the material that is being taken out will undergo decay for the rest of the calculation or until 1t is used again in a CHBM keyword The tempera ture of the new material will be set to that of the old one If both materials are being burned the materials will simply swap positions Their respective temperatures will be swapped as well Again these keyword can only be used in an STP block where the spectrum is being recalcu lated 2 8 7 Change TR card CHTR keyword The previous change keywords were used to change things on the material level The CHTR keyword can be used to perform changes on the geom
44. cted by using the DICTIN code from the PREPRO package 23 Error no ENDF file found ALEPH tried to open the ENDF file but couldn t find it Normally this error should not occur because this is checked before running ALEPH but it is still possible that something else happened Warning material has no atomic mass in the isotopes file atomic mass has been set to When reading the ORIGEN output an isotope was found that does not appear in the isotopes file The isotopes file does contain data for over 3000 nuclides but this does not mean that it is complete This error message might also indicate a problem with the isotopes file itself the element name field might be incorrect Error no ORIGEN output files found check executable ALEPH failed to open the ORIGEN output file Either there is something wrong with the executable compiler related problems or the name and path of the executable were Wrong 39 Chapter 6 ALEPH test problem and output 6 1 The NEA BUC single pin problem The NEA OECD BUC IV B benchmark 24 has been proposed by the expert group on burn up Credit BUC of the Working Party on Nuclear Criticality Safety NEA WPNCS The investigation of burn up credit for different types of fuel is an ongoing objective of the NEA OECD Burn Up Credit BUC expert group For this specific benchmark three geometrical 2D models were considered e A single pin in a moderator cell with reflective bou
45. culating spectra for point 1 Preparing MCNPX input file Running MCNPX Performing file clean up Reading spectra Calculating ORIGEN libraries for materials with temperature T 7 756e 08 Processing ENDF cross section file for nuclide 10010 E Calculating multigroup cross section for ENDF MT number 102 Processing ENDF cross section file for nuclide 10020 Calculating multigroup cross section for ENDF MT number 16 Calculating multigroup cross section for ENDF MT number 102 Processing ENDF cross section file for nuclide 922350 Calculating multigroup cross section for ENDF MT number 16 Calculating multigroup cross section for ENDF MT number 17 Calculating multigroup cross section for ENDF MT number 18 Calculating multigroup cross section for ENDF MT number 102 Processing ENDF cross section file for nuclide 982530 Calculating multigroup cross section for ENDF MT number 18 Calculating multigroup cross section for ENDF MT number 102 Processing ENDF cross section file for nuclide 992530 Calculating multigroup cross section for ENDF MT number 102 Performing evolution calculation for point 1 Processing irradiation history Calculating material composition for material 1 Accumulating burn up 44 6 3 2 Calculated cross section values Whenever asked by the user ALEPH will print out the one group cross section values and branching ratios used by ORIGEN to perform the depletion calculation Th
46. d must always be used after the BURN keyword Error on line invalid library number for temperature T ev The library numbers specified in the MAT keyword can be negative to indicate a library for reaction rate calculation only positive to indicate a transport library or zero for no library needed These library numbers must be integers between and including 99 and 99 because the positive ones have to be valid MCNP X library numbers Error on line no TMP keyword specified before MAT ALEPH has detected the use of the MAT keyword before the TMP keyword Because the number of library entries on the MAT keyword equals the number of temperatures on the TMP keyword the TMP keyword must always be used after the MAT keyword Error on line non existant material Isotopes on the MAT keyword are identified using their ORIGEN identifaction ZAMID The smallest possible nuclide is hydrogen with 10010 as identification Numbers smaller than this value are not legal identification numbers for isotopes Error no ENDF file for T eV was specified in xsdiraleph for wan ALEPH checks the xsdiraleph file for every isotope and library number other than zero on the MAT keyword to see if the ENDF file exists In this case ALEPH found no entry in the xsdiraleph file for this isotope and temperature The user should either fix the xsdiraleph by adding the isotope or he should remove the nuclide from the MAT keyword
47. dd more steps after a calculation was already finished or to restart a calculation with a slightly different irradiation history for which previously calculated spectra can be used 19 2 7 Specifying variable and burnable materials 2 7 1 Burnable materials BURN keyword The BURN keyword has to be used to specify the number NB of materials that are burned at any given time which is the same as the number of spectra to be calculated and the number of cells or collection of cells that are found in the tally specification in the MCNP X input file and to provide the material numbers of every burnable material used ALEPH allows for material reshuffling so the total amount of materials used in the problem may be greater then NB The syntax for this keyword is BURN NB BURNMAT 1 BURNMAT NB lt BURNMAT NB 1 BURNMAT NB k where BURNMAT 1 is a MCNP X material number The materials with i between 1 and NS are the materials that are to be burned at the start of the calculation The order in which these material numbers are given must be the same as the cell order on the tally used to calculate the spectra Additional material numbers for burnable materials that will be used later on in the calculation for instance through reshuffling are given after the first NB library numbers The temperature of a burnable material can be changed using the CHT keyword see section 2 8 4 during an irradiation step specified with the HIS keyword see
48. e cell c nuclear data DAT xs_aleph aleph xsdiraleph22_900 data is JEF2 2 at 900 K TMP 7 756E 08 temperature set to 900 K 7 756E 08 eV c spectra information BURN 1 1 burn material 1 VOL 52 8101725 EGS 1000 le 10 1000 le 9 1000 le 8 1000 le 7 1000 le 6 4000 le 5 4000 le 4 10000 le 3 10000 le 2 4000 le 1 4000 le 0 1000 le 1 1000 2e 1 the group structure c ORIGEN information ORI 02_THERM LIB buc lib 701 702 703 GXUO2BRM LIB DECAY LIB c MCNPX information ABS 0 9999 produce material composition responsible for 99 99 absorption TAL 4 the tally number that contains all the spectra MCNP menpx250_1fc 1 use MCNPX 2 5 0 c weigh the multigroup cross sections GWS 1 use constant flux weighing c output OUT 1 1 1 IRP 1 1 853398E 02 4 26 25 STP 1 IRP 1 1 853398E 02 4 26 25 IRP 1 1 853398E 02 4 26 25 IRP 1 1 853398E 02 4 26 25 IRP 1 1 853398E 02 4 26 25 STP 1 IRP 1 1 853398E 02 4 26 25 STP 1 IRP 1 1 853398E 02 4 26 25 51 rp IRP 1 1 853398E 02 IP 1 IRP 1 1 853398E 02 S T IRP 1 1 853398E 02 IP 1 IRP 1 1 853398E 02 Pd IRP 1 1 853398E 02 TPL IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 TE IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 PO DEC 4 30 TP 1 IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 TP 1 IRP 1 1 853398E 02 IP IRP 1 1 853398E 02 EE a IRP 1 1 853398E 02 BA
49. e overview consists of 8 columns total n y total n 2n n 3n fission n n p and the n y and n 2n branching ratio for every nuclide considered To provide an easy overview one group cross section values that are zero have simply been left blank in the table All cross section values are expressed in barn For the first point of the single pin calculation the following values were used by ORIGEN the n 3n fission and n columns were left out due to lack of space Cross section values used in evolution calculation for material 1 for point 1 n gamma n 2n atak n p n g BR n 2n BR 10010 7 32792E 03 10020 1 45761E 05 2 35605E 03 10030 8 32168E 05 a 20030 La 1 17128E 02 30060 8 54820E 04 1 61288E 03 30070 8 10438E 04 2 31914E 05 sord 40090 2 44718E 04 ves 3 36593E 08 50100 1 10028E 02 3 27360E 03 50110 1 35484E 04 2 40301E 06 7 48990E 07 110230 1 46639E 02 3 08755E 06 7 10060E 04 4 04840E 01 130270 6 45261E 03 2 14205E 06 1 72824E 03 150310 4 36780E 03 2 58941E 06 1 41911E 02 160320 1 25646E 02 1 07506E 08 2 70882E 02 160330 8 10124E 03 1 28321E 04 2 42587E 02 922340 1 81934E 01 8 59481E 04 922350 4 57342E 00 5 71103E 03 922360 9 09639E 00 4 19806E 03 922370 1 35042E 01 1 14287E 02 922380 8 13424E 01 6 15924E 03 ond 932370 2 14749E 01 1 31837E 03 pra 7 41493E 01 932380 4 66131E 00 7 30583E 03 ime 932390 1 25471E 0 1 85385E 03 para 1 15044E 01 982520 1 69042E 00 4 21989E 03 982530 1 52449E 01
50. etry level It allows a user to change the tr card number of a surface entry SURF to another tr card number NEWTR CHTR SURF NEWTR The user should test 1f the surface transformation works properly before using this option As was the case with the previous change keywords this keyword can only be used in an STP block where the spectrum is being recalculated 2 9 Optional keywords 2 9 1 ORIGEN library title TIT keywords The TIT keyword can be used to specify a title for the calculation It will be used as the title of the ORIGEN libraries TIT title 2 9 2 Inputfile comment C and keywords To put comments in the input file two different keywords have been foreseen C and The C keyword can be used to comment out entire lines while the keyword is used to comment out parts of lines The keyword can be used everywhere when the regular input is present This means that comment after every isotope in the MAT keyword like MAT 922350 03c 15c U235 from JEF 2 2 942380 03c 15c U238 from JEFF 3 0 or comment in the HIS keyword like this HIS STP 1 point 1 calculate spectrum IRP 2 0 25 4 30 constant power irradiation for 30 days STP 0 IRP 2 0 25 4 30 constant power irradiation for 30 days is permitted Using the C keyword in those keywords is however not permitted 23 2 9 3 Output OUT keyword The OUT keyword is used to set certain output options of ALEPH For now there are four opt
51. f 24M Am we can even speak of a major improvement We also observe a global agreement between WIMS and APOLLO for the fission products in the three cases especially for the Sm isotopes The fact that ALEPH doesn t perform as well on the fission products the compositions show differences of 0 5 to 10 with APOLLO as with the actinides is probably caused by the fact that the version of ALEPH used for these calculations still uses the original fission yield values from the ORIGEN library A newer version of ALEPH will also update these yield values 42 6 2 Problem summary The first part of the ALEPH output file consists of the ALEPH input instructions and the MCNP X input file This is followed by the problem summary This is an overview of some important calculation parameters such as the number of groups used and the upper and lower energy considered in the calculation The problem summary also gives an overview of the burnable and variable materials that are being used in the problem For every material the composition in g cm will be given For burnable materials that are being burned or for variable materials that are being used in the beginning an overview of the cells volumes and temperature will also be given The problem summary for the single pin problem looks like this Problem summary Group Structure 43000 groups E_ min le 05 eV E_max 2e 07 eV Burnable materials 1 materials being burned in 1 different cells
52. for in keyword OUT Input on the OUT keyword is with 0 and 1 only In most cases this simply means that not all 4 output options were specified Error on line unknown keyword The ALEPH input parser has detected an unknown keyword in the input file Error on line incorrect input detected ALEPH tried to read an input parameter but found something else or nothing at all In this case ALEPH will print the line with all the input parameters that the code expects followed by the original line that it found Check the input option for the correct syntax to solve the problem 36 5 2 Error on line unexpected input ALEPH has detected input beyond the regular input of a keyword Only the style comment is allowed on the same line when the regular input of a keyword is finished Error no input of keyword Certain keywords are required by ALEPH while others are optional see table 2 1 In this case ALEPH has detected that one of those keywords has not been used Reading the MCNP X input file Error temperature specified for cell number not specified in TMP keyword While going through the cell entries ALEPH has found a cell with a burnable or variable material that uses a temperature that has not been declared on the TMP keyword Error density specified for cell number differs from the density specified for a previous cell Error temperature specified for cell number differs from the
53. he Atomic Mass Evaluation 2003 included into NUBASE 21 from the Atomic Mass Data Center This has been done to ensure data consistency between ALEPH and MCNP X both codes now use the same values for atomic masses For now we have limited ourselves to microscopic cross section data All of the other data required by ORIGEN being the branching ratios the direct fission yield can also be be found in the different nuclear data evaluations that use the ENDF format By using this data from the ENDF files we would have no need of other third party data or models We are currently implementing this into ALEPH and ALEPH DLG By only using data from a single source being an ENDF file we can ensure full nuclear data consistency within the entire code system 12 Chapter 2 ALEPH input options 2 1 The ALEPH input file The ALEPH input file consists of two distinct parts This chapter deals with the different input options for ALEPH being the first part of the input file We have chosen to use a keyword approach similar to that used in ORIGEN Some keywords are even copied from ORIGEN such as the IRP IRF and DEC keywords As we pointed out before we strived to make ALEPH as simple to use as possible and this reflects in these input options The number of keywords that are actually required is kept as small as possible see table 2 1 This part of the input file is ended with the END keyword see section 2 2 after which the MCNP X
54. ime units in the irradiation history Time Unit seconds minutes hours days years 103 years 10 years 10 years Hg a Ooo J Ln L LA a in which e is the elementary electron charge Na is the number of Avogadro Of is the one group fission cross section for nuclide i this is taken from the modified ORIGEN libraries for the material in question o is a measure of the total flux in the material j in fact MCNP X will provide us with o jV per source particle for every material that we are burning and where O fi is the total fission Q value both the prompt and delayed of the nuclide i expressed in MeV by the following formula this is the same formula as the one used inside ORIGEN 14 Of 1 29927 103 22495 433 12 2 8 2 f L ALEPH will use the specific normalisation power Po j corresponding to the last spectrum calcu lated In other words the power is calculated by using the composition of the material when its spectrum was recalculated So 1f the user has specified points without a spectral recalculation the specific normalisation power Po will not be updated with the new composition This is currently under investigation It should be noted that for now the specific normalisation power Pp will be zero for a material that doesn t contain actinides It is therefore impossible to calculate the evolution of such materials by using constant power irradiation Constant flux irradiation is
55. ineering 62
56. ions full output OUTFULL library changes LIBCHANGES full library output LIBFULL and ORIGEN output ORIGENOUTPUT OUT OUTFULL LIBCHANGES LIBFULL ORIGENOUTPUT If OUTFULL 1 everything that is written to the screen will be writtten to the output file see section 6 3 1 When performing large amounts of steps with library recalculation this can lead to large output files For LIBCHANGES 1 the changes made to the original ORIGEN file will be reported see 6 3 2 If LIBFULL 1 ALEPH will print the libraries generated to separate files This last option can be used when preparing updated libraries for normal ORIGEN calculations Again in the case of large calculations with many materials and steps this may lead to a large amount of files If ORIGENOUTPUT 1 ALEPH will print the ORIGEN output files to separate files with the following naming convention This keyword is not required By default all output options are assumed to be 0 26 Chapter 3 MCNP X input 3 1 The MCNP X input file This chapter deals with the second part of the ALEPH input file the MCNP X input file For the ALEPH interface to work ALEPH must be capable of easily going through the input file For this some basic rules have to be established the ALEPH MCNP X parser is not as complete as the one used by MCNP X itself But apart from those rules anything goes When ALEPH reads the MCNP X input instructions it will read in all cells
57. is JEF 2 2 as the results given in the BUC benchmark are obtained with JEF 2 2 data The fuel cell has a temperature of 900 K the JEF 2 2 data at 900 K that we used can be found in ALEPH LIB 20 The cladding and moderator cell have a temperature of 620 and 575 K respectively We used ALEPH DLG to prepare the required data at these temperatures The calculation time was 280 hours on a single Dual Xeon 3 GHz machine Table 6 1 gives the relative difference of ALEPH 1 1 2 and WIMS8A 25 to APOLLO 26 after an irradiation of 420 870 and 1320 days the end of the three cycles EOC1 EOC2 and EOC3 Figure 6 1 shows the results for the end of cycle 3 Figure 6 1 and table 6 1 shows that ALEPH performs admirably well compared to APOLLO2 for the actinides All U and Pu isotopes except Pu have final compositions within 1 5 of those of APOLLO The 2 Np content is computed lower and higher respectively by WIMS and ALEPH as compared to APOLLO It should be noted that the 227Np build up is mainly linked to the n 2n reaction of 238U for MOX fuel As we showed before the value for this reaction is in perfect agreement with MCNPX 2 5 f see table tab reactionratesu238comparisonmcnpx25e so this is probably a data problem in the three codes The Am and Cm isotopes are predicted within 1 5 to 4 from the values given by APOLLO In general ALEPH performs better compared to WIMS for the actinide compositions except for 37Np 28Pu and Cm In the case o
58. ither the constant spectrum option 1 or the PWR weight spectrum option 2 In this case ALEPH did not find any of these options Error on line no input or illegal negative input for in keyword GWS This error occurs when no input or illegal input negative values are used for any of the 4 input parameters for PWR spectrum in the GWS keyword Error on line keyword GWS used after the MAT keyword ALEPH has detected the use of the GWS keyword after the MAT keyword A global weight option can be set for the weight spectrum using the GWS keyword but this keyword must appear before the MAT keyword This error will most likely be accompanied by errors in the MAT keyword on the individual isotopes weight spectra See the input instructions for the GWS and MAT keyword for more details see sections 2 3 2 and 2 4 3 Error on line illegal input negative or zero for the number of materials burned in the keyword BURN Because ALEPH allows for variable materials both those that are being burned and those that are not not all burnable materials must be burned from the beginning In order to avoid confusion the number of burnable materials that are being burned at any given time is the first input parameter on the BURN keyword This must be a non zero and non negative integer Error on line illegal input in the keyword the identification number of a variable material cannot be zero or negative The BU
59. l s density cannot be 11 changed by the user Burnable materials that are being taken out of the model will by default undergo decay At the beginning of every new burn up step the compositions of the materials that are being burned are updated along with the other material and geometry changes requested by the user For the purpose of transport calculations we truncate the material composition calculated by ORIGEN using a fractional absorption criterion specified by the user Only those nuclides responsible for e g 99 or 99 9 of all absorptions are included nuclides that were originally present are added by default and do not necessarily contribute to this fractional absorption criterion This entire process continues until the end of the calculation 1 3 2 Nuclear data for burn up applications The Monte Carlo code and the burn up code have different demands on nuclear data The Monte Carlo code requires specific data being microscopic cross sections angular distribu tions energy spectra for every nuclide used in the transport simulation ORIGEN on the other hand needs microscopic cross sections for n y n 2n n 3n n n p and fission reactions and this for every nuclide in the transmutation chains To correctly calculate the dis tribution of fission products ORIGEN also requires direct fission yield data associated with 8 primary actinides 32Th 2331 250 2380 239 24 Pu 25Cm and 2 2Cf The n 3n 1 0
60. ldn t find it Error in line path specified with DAT keyword does not exist or the isotopes file is missing To check if the data path specified with the DAT keyword is correct ALEPH will attempt to open the isotopes file that should be located there If ALEPH fails to open the isotopes file either the data path is incorrect or the isotopes file is missing Either way ALEPH cannot run the calculation Error in line the ALEPH xsdir file is missing ALEPH tried to open the xsdiraleph file that was specified using the DAT keyword Be cause ALEPH can t find it the code cannot find the required cross section data required to calculate the ORIGEN libraries Warning material has no atomic mass in the isotopes file atomic mass has been set to When reading the xsdiraleph file ALEPH will read the value for the atomic mass from the isotopes file for every nuclide entry in the xsdiraleph file Although the isotopes file is as complete as possible it contains data for over 3000 nuclides it might be possible that the xsdiraleph file points to nuclides that are not included in the isotopes file In that case ALEPH will use the atomic mass number A of the nuclide as an estimate for the atomic mass Whenever this warning is issued the user should check the isotopes file and add the missing data Error on line input of negative temperature with temperature keyword TMP Temperature input on the TMP keyword must
61. les from ORIGEN The number of significant digits used by ORIGEN was increased from 3 to 5 to address the perceived round off problem mentioned above although tests showed little to no influence on the final result The new compositions are then stored and passed on either to build a new MCNP X input file for a new burn up step or new ORIGEN input files to obtain composi tions at intermediate points within the burn up step We have also added the possibility to change materials and geometry in the model during the irradiation ALEPH distinguishes between 2 different types of materials variable materials and burnable materials Variable materials are materials that can be changed by the user but that can not be burned A user can therefore change the density and or temperature of such a material to for instance simulate heating effects of water or even replace the material by another one to take into account changes in the boron concentration in the coolant of a PWR Geometry changes by using surface transformations are also possible for instance for the simulation of control rod movement Burnable materials are obviously being burned As was the case with variable materials a user can change the temperature of such a material or even replace it with another burnable material which doesn t necessarily have to be one that is being burned this is to simulate core reshuffling and reloading For obvious reasons a burnable materia
62. nd DEC four material control keywords CHT CHD CHCM and CHBM and one geometry control keyword CHTR 2 8 1 Power irradiation IRP keyword The IRP keyword is used to specify a sub step of constant power irradiation along with the irradiation time IRP lt IRP option gt TU TIME where lt IRP option gt is the power normalisation option TU the time unit to be used see table 2 3 and TIME the total time elapsed at the end of this sub step since the beginning of the step referenced from the beginning of the STP block in which this keyword appears The power renormalisation option lt IRP option gt is used to determine the power for every burnable material so that every material is burned relatively to the other materials in the system according to the power distribution The power in a system and in the different materials that compose the system is determined by the power distribution which is in turn determined by the flux distribution The power produced in a material j consisting of N different nuclides i characterised by a proton number Z mass number A with density p and fission cross section Of is proportional to what we call the specific normalisation power P j This power Po represent the relative power distribution in all materials In essence Po is the total power produced through direct fission and delayed energy expressed in MW 24 a PiViNa Po e 10 Y EGE 57 00 j 2 8 1 i 1 i 21 Table 2 3 T
63. ndary conditions This model is representative for an infinite medium and provides information about reactivity and fuel inventory as a function of burn up The pitch of the cell must be modified to take into account the assembly global moderation ratio that the pin cell simulates e An assembly model with reflective boundary conditions This model provides more in formation than the pin cell model since the pin power distribution and pin material in ventory can be computed The assembly calculation is a standard calculation step in a complete design core calculation e A supercell model that aims to calculate a fuel assembly while taking into account its environment It is especially recommended when dealing with MOX fuel as this type of fuel is loaded in cores filled with a larger number of UO fuel assemblies the MOX fuel assembly behavior depends on the neutron spectrum established in the neighboring assemblies In particular power peaking at the MOX fuel assembly border needs to be carefully assessed Two types of MOX fuels were considered in the benchmark specifications MOX made with reactor grade plutonium RG MOX and MOX made with weapon grade plutonium WG MOX The depletion was assumed to be at constant power and the irradiation history consists of three cycles of 420 days with 30 days of downtime in between the cycles The third cycle is then followed by a cooling time of 5 years At the end of the irradiation the spent fuel must
64. o code to a burn up module to improve the accuracy of depletion and activation calculations For every time step the Monte Carlo code will provide reaction rates to the burn up module which will return new material compositions to the Monte Carlo code So if static Monte Carlo particle transport is slow then Monte Carlo particle transport with burn up will be even slower as calculations have to be performed for every time step in the problem The computational issues to perform accurate Monte Carlo calculations are however contin uously reduced due to improvements made in the basic Monte Carlo algorithms due to the development of variance reduction techniques and due to developments in computer architec ture more powerful processors the so called brute force approach through parallel processors and networked systems This evolution of computer architecture is going to continue in the future Moore s law on computer processor development clearly states that the speed of processors doubles every year So within 10 years we will see computers that are 1000 times faster compared to our high end computers of today although it is possible that constraints such as processor cooling will limit the validity of this law in the future new technologies might however resolve this issue In recent years these developments have created a renewed interest in Monte Carlo burn up As a matter of fact work is now under way at LANL to finally include a transmu
65. odels but MCNP would not This is why we decided to omit such nuclides from transport calculations Fatal error occurred in MCNP X run terminated A fatal error has occurred in MCNP X The best solution is to try the MCNP X input file in a pure MCNP X run to see what is wrong and to correct the problem Again it 1s up to the user to provide a working MCNP X input file for use with ALEPH Bad trouble occurred in MCNP X run terminated This error is similar to the one above The best solution is to try the MCNP X input file in a pure MCNP X run to see what is wrong and to correct the problem Again it is up to the user to provide a working MCNP X input file for use with ALEPH Wrong executable for MCNP X or unknown trouble in MCNP X run terminated When the operating system returned control to the ALEPH executable the code didn t find the MCNP X output file It is possible that the MCNP X executable is wrong or that there was another unexpected problem with MCNP X 38 Error this is not a linearized ENDF file Fault detected in file 3 mt When calculating the multi group cross sections ALEPH has found an ENDF file with a cross section that it requires that has not been linearized or that has multiple interpo lation zones This can happen when an unprocessed ENDF file is used or when there is something wrong with the index of the ENDF file in file 1 mt451 In this last case the index can be corre
66. of transport calculations ALEPH will truncate the material composition cal culated by ORIGEN using a fractional absorption criterion ALEPH will calculate the total 18 absorption of every nuclide being the sum of all reactions given in table 2 2 and use it to de termine to what amount this isotope contributes to the total absorption The nuclides are sorted in decreasing absorption importance and they will be added to the list as long as the cumulative fractional absorption is not equal to or greater than the fractional absorption criterion specified by the user After that only nuclides that were originally present are still added As such the initial nuclides are used by default The user needs to use the ABS keyword to specify this fractional absorption criterion FRAC ABS FRAC Because it is a fractional criterion FRAC must be a number between O and 1 If this number is 0 the original nuclides with their new composition will be used Using 1 is not advised because 1t will add all nuclides some of which will probably not have a library associated with them Acceptable values are for instance 99 and 99 9 The density used by MCNP X will be the density of this truncated nuclide list in order to conserve the absolute number of atoms 2 6 2 MCNP tally specification TAL keyword The TAL keyword has to be used to specify the tally number TNR this must be a type 4 tally used to calculate the spectra for the purpose of an ORIGEN libra
67. ome of the error and warning messages to be more clear to the user e New feature a fourth output option has been added to the OUT keyword write the ORI GEN output file for every ORIGEN calculation to a separate output file This has been added to allow the user access to ORIGEN data other than compositions such as toxici ties e New feature in the ENDF format the n 2n n p and n reactions can be represented using a summation cross section discrete levels and a continuum in the same way that inelastic scattering is represented For n 2n this is mt16 and mt875 891 for n p this is mt103 and mt600 649 and for n this is mt107 and mt800 849 The fission reaction is also defined as a summation cross section mt18 with partials mt19 21 and mt38 for first second third and fourth chance fission The ENDF format states clearly that the summation cross section should always be given if any partials are present for the fission n p and n reaction in other words mt18 mt103 and mt107 should always be present This is not the case for n 2n For Be from JEFF 3 1 only the partials of the n 2n reaction are given In order to correct problems like this ALEPH will now first check if the summation cross sections are present and if they are not ALEPH will look for the partials and use those should they exist 50 Appendix B NEA BUC MOX pin cell input TIT NEA Burn up Credit Criticality Benchmark singl
68. ometimes users also consider very few burn up steps over long periods of irradiation or provide a small number of burn up zones in which reaction rates are to be calcu lated This can have profound consequences on the accuracy of the calculation By adopting a new approach to reaction rate calculation where we perform the calculation outside the Monte Carlo code one can reduce the calculation time significantly while ensuring maximum accu racy The nuclear data itself should also get some attention Apart from the reaction rates there is still the need for accurate branching ratio and direct fission yield data By using nuclear data evaluations in the ENDF format 11 we have access to all this information along with the cross sections themselves at once This also allows for great flexibility because we can start using new data whenever it becomes available An automatic coupling with NJOY 12 or a similar code will also facilitate library generation Another weak point either true or conceived of Monte Carlo burn up codes is their complexity These codes often use a script or link approach so that the user would have to understand and manage a large number of input and output files while the conversion of data from one form into another would introduce approximate results due to successive round off 2 This can be solved by providing an easy to use interface that actually hides the burn up code which is something the new transmutation o
69. on self shielded PWR spectrum 2 ETH T EEPI EFIS where ETH is the upper energy boundary of the thermal region T is the nuclear tempera ture EEPI is the upper energy boundary of the epithermal region and where EFIS is the mean energy of a fission neutron All the energies must be specified in MeV and the temperature must be specified in K The GWS keyword is used to select this weight spectrum globally This means that all group cross sections will be calculated using the same weight spectrum If the user wants to mix different weight spectra then he may not use this keyword Instead he will need to specify the weight spectrum for every isotope in the MAT keyword see section 2 4 3 Also if this keyword 1s used it must be used before the MAT keyword 2 4 Nuclear data 2 4 1 Data library DAT keyword ALEPH requires microscopic cross sections which can be linearly interpolated for every isotope that the user wants to change in the ORIGEN input file These cross sections have to be supplied in the ENDF 6 format 11 We have chosen the ENDF format and not some other home made format because it is universally accepted and because there exist numerous cross section processing codes that are capable of linearizing cross sections in the ENDF format Examples of such data processing codes are PREPRO 2002 23 and NJOY the ALEPH DLG utility to generate MCNPX and ALEPH cross section files uses NJOY 99 112 Furthermore it is not re
70. or the moment Adding a natural element to this list is possible but 1t will never be used 2 5 Calculating ORIGEN libraries 2 5 1 Running ORIGEN ORI keyword The ORI keyword is used to specify the ORIGEN 2 2 executable origenexe ORI origenexe This is required because there exist different versions of ORIGEN one for thermal reactors and one for fast reactors 2 5 2 Libraries LIB keyword ORIGEN 2 2 needs three types of libraries to run the cross section libraries ALEPH uses an original library and will adapt it where necessary the decay library and the photon libraries The LIB keyword is used to specify these libraries LIB xsfile APL ACL FPL photonfile decayfile where xsfile photonfile and decayfile are the names and path of the files in question The integers APL ACL and FPL are used to specify the identification numbers of the activation product actinide and fission product libraries in the original ORIGEN library that the user wants to change If the user has specified an isotope in the MAT keyword that is not present in this library it will not be added Nuclides present in the original file will not be touched if they do not appear in the MAT keyword The new libraries will receive new identification numbers to distinguish them from the original library 701 for activation products 702 for actinides and 703 for fission products 2 6 Running MCNP X 2 6 1 MCNP material composition ABS keyword For the purpose
71. ption in MCNPX will achieve These insights and developments have resulted in the creation of ALEPH a Monte Carlo burn up interface code that is capable of using any version of MCNP 13 or MCNPX 1 with ORIGEN 2 2 14 for the evolution calculation and NJOY 99 112 12 for the nuclear data ALEPH is currently under development at SCKeCEN as part of a PhD work in collaboration with Ghent University in the framework of the MYRRHA project 15 The main idea behind ALEPH was to create a general purpose Monte Carlo burn up code that is efficient flexible and easy to use Another point that could be seen as a minor problem of burn up codes in general this includes deterministic codes and that ALEPH will try to address is that the user has to specify himself when the reaction rates need to be recalculated Because of this a user has the inclination to recalculate the reaction rates more often than actually would be required a standard crite rion is after every GWd ton of accumulated burn up A time step optimisation routine would therefore be interesting It would give the user an initial estimate of the time steps required during the irradiation history Such a time optimisation routine is currently being developed for ALEPH again using a hybrid technique to reduce calculation time From table 1 2 we can again conclude that there are no other MC burn up codes that provide such an optimisation It is however possible that MCB might have such a thing
72. quired to use the complete ENDF file Only the reactions given in table 2 2 have to be present in file 3 the cross section file of the ENDF file If one of those entries is missing ALEPH will assume that the cross section is zero Table 2 2 Cross sections required by ORIGEN 2 2 and their ENDF reaction numbers MT number MT Reaction Material Types 16 or 875 891 n 2n all materials 17 n 3n actinides 18 or 19 21 38 n fission actinides 102 n y all materials 103 or 600 649 n p activation and fission products 107 or 800 849 n a activation and fission products 16 The DAT keyword has to be used to specify the path where ALEPH can find the ENDF files that are specified in the ALEPH cross section directory file This path is the common path to the cross section files separate subdirectories can still be specified using the cross section directory file The structure of the cross section directory file is given in section 4 1 The syntax of this keyword is as follows DAT datapath alephxsdir 2 4 2 Temperature TMP keyword ALEPH allows the use of multiple temperatures both for variable materials and burnable ma terials The TMP keyword has to be used to specify all the temperatures T i the unit used is K that the user will require The order of the temperatures is also the order in which libraries are specified in the MAT keyword The syntax for this keyword is TMP T 1 T 2 T NT gt with
73. rallel calculations which have to specified after the MCNP X executable Option 1 has to be used for serial calculations and option 2 for parallel calculation In this case there probably is incorrect input in this line Error on line number of parallel slaves must be non zero in keyword MCNP For parallel calculations the MCNP X command line needs an extra argument the number of parallel tasks The amount of parallel slaves must be a non negative and non zero integer Please note that it is up to the user to allocate the proper amount of resources to run the calculation Error on line illegal input negative or larger than 1 for the fractional absorption limit in keyword ABS The fractional absorption criterion to be used by ALEPH is specified as an integer be tween 0 and the values of 0 and 1 are allowed Error on line illegal input negative value with keyword TAL The tally number must be a positive integer Error on line the tally specified with the TAL keyword is not a type 4 tally ALEPH requires a neutron flux tally in all cells with burnable materials This is a type 4 tally so that the tally number must end with 4 for instance 4 14 24 In this case the tally number specified is not a type 4 tally 33 Error on line weight option in the GWS keyword does not exist ALEPH foresees in the use of two possible weight options for calculating the multi group cross sections e
74. ransformation card number on a surface card during an irradiation step Previously geometry changes were only possible by using the CHCM card where a cell s material was replaced by another A disadvantage of this method was that every cell that took part in the geometry change had to be modeled separately something which is not the case with the surface transformation ALEPH version 1 1 1 August 2005 e Minor bug fix due to an indexing error in the calculation of multi group cross sections it was possible to have negative cross sections when the upper energy limit of the linearized cross section is smaller than the upper energy of the group structure This has been corrected 49 ALEPH version 1 1 2 November 2005 e Minor bug fix when adding tallies to the MCNP X input file other than the tally for the calculation of the spectra for the reaction rate calculation ALEPH stopped with a segmentation fault when starting to calculate the ORIGEN libraries This was due to a problem when reading the tally output file that contained multiple tallies This has been corrected e Minor bug fix under some circumstances the total composition of a burnable material that was placed into a cell again when it had been taken out in a previous step produced some anomalies impossible nuclide identification numbers This was due to a test that was missing when assigning the correct library numbers This has been corrected e Minor fix rewrote s
75. ross section data for ALEPH and MCNP X by automating the entire NJOY process By December 2004 we decided to add features to allow for core reshuffling multiple tem peratures Because these features dictated drastic changes in the original source code we decided to rewrite the entire code Along the way we also added the possibility to change the geometrical specification of the MCNP X model to allow for variable geometry to sim ulate for instance moving control rods This reports deals with this fourth incarnation of ALEPH the first official release of ALEPH version 1 1 2 Before we conclude we would like to express my gratitude to a number of people First of all John Hendrickx LANL for some fruitful discussions both by e mail and in person on the internal workings of MCNPX Luc Borms for encouraging me to try C and for his technical help Gert Van den Eynde for his help concerning numerical techniques Ben Vanhaeren for his help concerning compilers and the Linux operating system Thierry Aoust Edouard Malambu Nadia Messaoudi Vitali Sobolev and Andr Beeckmans de West Meerbeeck for their interest in ALEPH from the very beginning Thanks also to Dirk Maes and everybody else for enduring my incessive ranting about this subject and Martine Vos for making the ALEPH logo Wim Haeck and Bernard Verboomen Mol January 2006 Contents 1 Introduction to ALEPH 7 1 1 The Monte Carlo method and burn up applications
76. ry calculation TAL TNR 2 6 3 MCNP execution MCNP keyword The MCNP keyword is used to specify the MCNP X executable mcnpexe to be used within the program followed by the parallel option lt parallel gt MCNP menpexe lt parallel gt where lt parallel gt is one of the following e for serial runs single processor calculation 1 e for parallel runs multi processor calculation 2 NSL where NSL is the number of parallel processors It is up to the user to ensure that enough processors are reserved and that mcnpexe points to an executable capable of parallel processing For every burn up step with library calculation ALEPH will create an MCNP X input file and MCNP X will create output files and tally files The names for these files consists of the step number starting with O for the initial library calculation with an extension 1 for an input file o for an output file and m for a tally file The so called runtape files with the standard name runtpe are deleted by ALEPH to reduce disk space because these files can take up quite a lot of space The user should ensure that enough disk space is available to store all these files a single pin calculation of 50 steps takes already 500 MB of disk space ALEPH will always check for the existence of tally files before running MCNP X If a tally file exists ALEPH will skip the MCNP X run for that step This was added to allow for continue runs to a
77. s keyword can be applied to both variable and burnable materials This could for instance be useful to change the temperature of a burnable material during a step where it will produce a lot more or less power compared to the previous step The temperature TEMP must also have been declared using the TMP keyword If the temperature has not been specified or if the TMP keyword is used after the HIS keyword ALEPH will issue an error This keyword can appear anywhere within an STP block but only if the libraries are recalcu lated The changes will be performed at the beginning of the step If this keyword is used in an STP block where the spectrum is not recalculated ICAL 0 for the block ALEPH will issue an error This also applies to the CHMD CHCM CHBM and CHTR keywords 2 8 5 Change density CHMD keyword The CHMD keyword is used to change the density of a variable material MAT to DENSITY CHMD MAT DENSITY This can only be applied to variable materials If a burnable material is chosen an error will be issued As was the case with the CHMT keyword this can only be used in an STP block where the spectrum is recalculated 2 8 6 Change material CHCM and CHBM keywords Changing materials in a cell is done through the CHCM and CHBM keywords depending upon the type of material variable or burnable For a variable material for instance you change the material of a single cell with the CHCM keyword CHCM CELL NEWMAT where CELL is the
78. section 2 8 and a burnable material can be changed to another by using the CHBM keyword see section 2 8 6 You may use any number of materials both to be burned and to be used later on there are no limitations except maybe for the memory available on the system used to run the program in our present time that shouldn t be much of a problem 2 7 2 Volumes VOL keyword ALEPH requires the volumes of the cells containing the burnable materials to recalculate the total fluxes power distributions Because these cells can consist of multiple parts and be cause they can be repeated in lattices 1t is not possible to rely upon MCNP X to normalize the flux tallies The user has to use the VOL keyword to specify the total modeled volume of the cells containing burnable materials VOL V 1 VINB where NB is the total number of materials that are burned as any given time see the BURN keyword section 2 7 1 The order in which these volumes are given must be the same as the order of materials on the BURN keyword 2 7 3 Variable materials VAR keyword The VAR keyword is used to specify the variable materials in the problem Variable materials can help in better modeling the irradiation conditions in certain problems because they allow the change of parameters such as composition temperature This is for instance useful for simulating the boron concentration in water for a PWR If the user has no need to change materials this key
79. tation module as a standard option into MCNPX 1 2 3 one of the standard Monte Carlo codes available today with a development history of over 30 years An overview of different MC burn up codes can be found in table 1 1 4 10 Table 1 1 Overview of Monte Carlo burn up codes in existence Code Name MC Code Burn Up Code Year Institution MOCUP 4 MCNP4A ORIGEN 2 1 1995 INEL MC REBUS 5 MCNP REBUS 1998 ANL OCTOPUS MCNP FISPACT 1998 NRG MONTEBURNS 6 MCNP4B ORIGEN2 1999 LANL EVOLCODE 7 MCNP4B ORIGEN 2 1 1999 CIEMAT MCB 8 MCNP4C Custom 1999 KTH MCWO 9 MCNP ORIGEN2 2000 INEEL MVP BURN MVP Custom 2000 JAERI BURNCAL MCNP4B Custom 2002 SNL MCODE 10 MCNP4C ORIGEN 2 1 2002 MIT ALEPH MCNP MCNPX ORIGEN 2 2 2004 SCKeCEN MCNPX 1 MCNPX CINDER 2005 LANL 1 2 Improving Monte Carlo burn up By examining and reflecting upon the basic Monte Carlo transport algorithms on the tally track length estimator which is used to calculate the reaction rates and on the basic functions of a Monte Carlo burn up code we have identified some points of possible improvement in Monte Carlo burn up calculations First of all Monte Carlo burn up codes are quite time consuming and thus rather inefficient due to the sheer number of reaction rates that have to be calculated Because of this most users decide to limit the number of actinides and fission products to be considered in the trans mutation chains S
80. the only possibility to do this This will be fixed in a future version of ALEPH There are four possibilities for this power normalisation option e The power for every material i is specified This option must be followed by a number of power values P i equal to the number NS of materials to be burned 1 P I PINS In this case no further normalisation of the power will occur This option should be avoided because it doesn t take into account the relative power distribution in the system e The total power PTOT for all materials is given 2 PTOT The power of every material j is now determined by using the relative power level Po of the material to that of all materials together Po N Y Pol l 1 P PTOT 2 8 3 22 e The power P of a specific material IK is given 89 EK P The relative power level of every other material j to material IK is now used to determine the power values of those materials Po j P P 2 2 8 4 Po IK e The total power PCTOT of a collection of materials with index I i i 1 NC is given 4 NC 1 1 I NC PCTOT This option is somewhat similar to the second and third option For materials in the collection the relative power level of the material to that of all materials in the collection 1s used see equation 2 8 3 For materials that aren t in the collection the relative power of that material to the power of any material from the collection is used see equation
81. us 3 Ez 0 475 cladding outer radius IA E ESE A E E A c data cards c average MOX E Nat 6 987923E 02 ml nlib 01c 6 987923E 02 92234 2 5952E 7 72 0 92235 5 4287E 5 2 0 92238 2 1387E 2 5 20 94238 4 6610E 5 2 0 94239 1 0156E 3 Sen 94240 4 8255E 4 Spe 94241 1 7491E 4 250 94242 1 3201E 4 2200 8016 4 6586E 2 SO A cosi Dl lattici ini cali A DRE RENE c Zr 2 Nat 3 885870E 02 m2 nlib 03c 40090 1 9889E 02 40091 4 3373E 03 40092 6 6297E 03 40094 6 7186E 03 40096 1 0824E 03 26054 7 8068E 06 26056 1 2244E 04 26057 2 8291E 06 26058 3 7366E 07 24050 2 9690E 06 24052 5 7190E 05 24053 6 4841E 06 24054 1 6108E 06 I AA T rada A a e c water G Nat 7 265121E 02 m3 nlib 05c 001001 4 8414E 2 008016 2 4213E 2 005010 4 7896E 6 005011 1 9424E 5 mt3 lwtr 62t dpto o e e a al oto tin LE gl o AN Oe am c tallies c fc4 flux f4 sd4 1 0 39 fal E F fm4 1 0 Sflux c e print 85 128 PRDMP 0 0 100 kcode 20000 1 16 30 280 totnu sdef erg dl axs 0 0 1 rad d2 ext d3 spl 3 0 988 2 249 si2 0 0 410 si3 50 0 50 0 OK 60 Bibliography 1 2 3 4 9 10 11 12 J S Hendricks et al MCNPX VERSION 2 6 a LA UR 05 8225 Los Alamos Na tional Laboratory USA 2005 G W McKinney Transmutation Feature Within MCNPX 10th UK Monte Carlo User Group Meeting Teddington UK March 15 16 2004 H R Trellue Use of MCNPX CINDER90 for Residual Nuclide Production Deca
82. ut file It is no longer necessary to end a burn up step using the END keyword and it is no longer necessary to specify the total number of steps on the HIS keyword and the step number on the STP keyword e New feature a third output option has been added to the OUT keyword write the screen output to the output file ALEPH version 1 1 0 June 2005 e New feature material compositions can now be entered in atom fractions and densities can be entered as atoms barn cm as well Previously the compositions had to be entered as weight fractions and densities in g cm 77 As is the case in MCNP X both representation can also be interchanged compositions in weight fraction and density in atoms barn cm and vice versa are allowed Material compositions are now stored internally as atoms barn cm and no longer in g cm 3 In order to recalculate the compositions to this representation ALEPH requires the isotopes file that should be located in the directory given by the data path specified using the DAT keyword We have decided to update the atomic mass values from MCNP X by using the Atomic Mass Evaluation 2003 included into NUBASE 21 from the Atomic Mass Data Center These values can be found in the isotopes file and they are also included in the xsdir files provided with ALEPH LIB e New feature to allow for easier geometry changes the CHTR change keyword has been added This keyword allows a user to change the t
83. ux irradiation of material j with flux this becomes Post di Pa Vi P0 BU 10 6 4 2 where Po is the specific normalisation power of material j and o is a measure of the total flux in the material j in fact MCNP X will provide us with dp V per source particle for every material that we are burning see equation 2 8 1 For the single pin model this burn up summary looks like this Accumulated burn up summary MWd kg initial BM point 1 point 2 eee IN point sl total burn up 1 1 00000E 00 1 00000E 00 0 00000E 00 4 80000E 01 46 6 4 3 Material composition The most important part of the output consists of the material compositions For every burnable material an overview of the evolution of every nuclide is given For the single pin problem this look like Evolution of individual nuclides for material 1 g cm3 point 0 point 1 sile point 50 point 51 10010 0 00000E 00 7 98349E 10 setti 4 22820E 08 4 22820E 08 10020 0 00000E 00 3 95773E 15 v 1 21228E 11 1 21228E 11 10030 0 00000E 00 2 08759E 08 ua 9 12884E 07 6 89493E 07 922350 2 11885E 02 2 09231E 02 sisi 1 05881E 02 1 06173E 02 922360 0 00000E 00 6 69858E 05 sare 2 29681E 03 2 38345E 03 922380 8 45426E 00 8 44889E 00 sal 8 16957E 00 8 16957E 00 942380 1 84248E 02 1 82784E 02 1 57764E 02 1 66780E 02 942390 4 03154E 01 3 96673E 01 2 05135E 01 2 05846E 01 942400
84. word can be omitted The use of this keyword is similar to that of the BURN keyword VAR VARMAT 1 VARMAT NV 20 where VARMAT i is a MCNP X material number and NV is the total number of variable mate rials to be used The order in which these materials are specified is not important The temperature of a variable material can be changed using the CHT keyword see section 2 8 4 during an irradiation step specified with the HIS keyword see section 2 8 The density of a variable material is changed by the CHMD keyword see section 2 8 5 and the content of a cell is replaced by another variable material by using the CHCM keyword see section 2 8 6 2 8 Specifying burn up History HIS keyword The burn up history has to be specified with the HIS keyword This keyword must be followed by at least one burn up step A burn up step is simply a point in the irradiation history where a user can ask to output the material compositions or where the library needs to be recalculated Such a burn up step is specified using a block starting with the keyword STP and an integer ICAL to specify if the library has to be recalculated for this point ICAL 1 to recalculate and ICAL 0 to use the previously calculated libraries The STP block ends when another STP or when any other keyword is encountered HIS STP ICAL STP ICAL Within every STP block there are three different types of burn up keywords the usual ORIGEN options IRP IRF a
85. y LA UR 04 8433 SPIRAL 2 Workshop Caen France December 13 14 2004 R L Moore B G Schnitzler C A Wemple R S Babcock D E Wessol MOCUP MCNP ORIGEN2 Coupled Utility Program INEL 95 0523 Idaho National Engineer ing Laboratory Idaho Falls USA 1995 N A Hanan A P Olson R B Pond J E Matos A Monte Carlo Burnup Code Linking MCNP and REBUS 1998 International Meeting on Reduced Enrichment for Research and Test Reactors Sao Paulo Brazil October 18 23 1998 D L Poston H R Truelle User s Manual Version 2 0 for MONTEBURNS Version 1 0 LA UR 99 4999 Los Alamos National Laboratory USA 1999 E Gonzalez M Embid Segura A Perez Parra EVOLCODE ADS Combined Neu tronics and Isotopic Evolution Simulation System Mathematics and Computation 1999 Reactor Physics and Enviromental Analysis in Nuclear Applications Madrid Spain September 1999 pp 963 974 J Cetnar W Gudowski J Wallenius User Manual for Monte Carlo Continuous Energy Burnup MCB Code Version 1C Royal Institute of Technology Stockholm Sweden 2002 G S Chang J M Ryskamp Depletion Analysis of Mixed Oxide Fuel Pins in Light Water Reactors and the Advanced Test Reactor Nuclear Technology 129 pp 326 337 2000 Z Xu P Hejzlar M J Driscoll M S Kazimi An Improved MCNP ORIGEN Depletion Program MCODE and its Verification for High Burnup Applications PHYSOR 2002 Seoul Korea

ALEPH 1.1.2 A Monte Carlo Burn-Up Code - SCK-CEN

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