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Determination of criticality safety MCNP5 calculation bias by using different libraries of cross section data
Progress in Nuclear Energy 59 (2012) 96e99
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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Determination of criticality safety MCNP5 calculation bias by using different libraries of cross section data Jakub Lülely, Branislav Vrban*, Gabriel Farkas, Ján Has cík, Martin Petriska Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovicova 3, 812 19 Bratislava, Slovakia
a r t i c l e i n f o
a b s t r a c t
Article history: Received 12 October 2011 Accepted 23 April 2012
Application of different cross section libraries and different versions of Monte Carlo code MCNP has an influence on the calculation results and therefore determination of criticality safety calculation bias forms part of improving accuracy of simulations using computational systems and codes. In this paper, criticality calculations results are presented for 21 problems coming from the International Handbook of Evaluated Criticality Benchmark Experiments (International Handbook, 2007). All of these problems are related to VVER-440 reactors because of their extensive use in Slovakia. Three libraries of cross section data were investigated:
Keywords: Bias VVER-440 MCNP Cross section Benchmark
JEFF-3.1 General purpose library, ENDF/B-VII library, JENDL 4.0.
Calculations were provided with MCNP5-1.40 and MCNP5-1.60 transport codes. Two cluster systems situated at our Institute were used. Main purpose of this analyses was the determination of the bias which should be used in further simulations. Ó 2012 Elsevier Ltd. All rights reserved.
1. Introduction
2. Material and methods
Determination of criticality safety calculation bias caused by use of different libraries of cross section data and different versions of stochastic Monte Carlo N-Particle Transport Code (X-5 Monte Carlo Team, 2003) plays an essential role in process of tuning used calculation systems. These systems should be needed in analyses of diverse nuclear problems. In this paper we are presenting obtained criticality calculations results for 21 problems coming from the International Handbook of Evaluated Criticality Benchmark Experiments (International Handbook, 2007). Three sets of calculations were performed using JEFF-3.1 General purpose library (JEFF, 2005), ENDF libraries (ENDF, 2011) and Japanese Evaluated Nuclear Data Library version 4.0 (JENDL, 2010). Versions MCNP5-1.40 and MCNP5-1.60 were investigated. All these calculations have been provided on two computer clusters used at our Institute. Main purpose of this analysis was the determination and specification of the best bias which should be used in further calculations at Institute of Nuclear and Physical Engineering.
Problems coming from the International Handbook were chosen in regard to analyses performed at our Institute. Only VVER-440 reactors are situated and operated in Slovakia, so all of these problems are connected to this type of reactor and should be useful for our purposes. We tried to specify benchmarks problems at different temperatures focused on use of low enrichment uranium oxide as a fuel, light water or light water with boric acid as a moderator, gadolinium or cadmium as absorber, stainless steel or zirconium as a clad of fuel rods and use of hexagonal pitched lattices. Four main problems are presented. Each of them consists of another cases. Enrichment of 235U is at the level of 17.4% in IEU-COMP-THERM-002, 5.4% in LEU-COMP-THERM-019 and in LEU-COMP-THERM-021, and finally 4.924% in problem LEU-COMPTHERM-026. Further subdivision of problems can be found in Table 1. All of different libraries of cross section data in ACE format needed for MCNP5 were calculated by cross section processing program NJOY 99.364. (MacFarlane and Muir) At our Institute, two computer clusters for MCNP calculations are in operation. Cluster versions of MCNP5-1.40 and MCNP5-1.60
* Corresponding author. Tel.: þ421 260291504. E-mail address: [email protected] (B. Vrban). 0149-1970/$ e see front matter Ó 2012 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2012.04.007
J. Lülely et al. / Progress in Nuclear Energy 59 (2012) 96e99 Table 1 Specified problems coming from the International Handbook of Evaluated Criticality Benchmark Experiments (International Handbook, 2007). Spectrum
Thermal
Moderator
Pure light water IEU-COMP-THERM-002 LEU-COMP-THERM-019 LEU-COMP-THERM-026 Gadolinium IEU-COMP-THERM -002 (3) IEU-COMP-THERM -002 (4) Stainless steel IEU-COMP-THERM-002 LEU-COMP-THERM-019 LEU-COMP-THERM-026
Boric acid water LEU-COMP-THERM-021
400 K
500 K
IEU-COMP-THERM -002 (4) IEU-COMP-THERM -002 (6)
IEU-COMP-THERM -002 (2) LEU-COMP-THERM -026 (2) LEU-COMP-THERM -026 (4) LEU-COMP-THERM -026 (6)
Absorber
Clad
Temperatures 300 K IEU-COMP-THERM -002 (1) IEU-COMP-THERM -002 (3) IEU-COMP-THERM -002 (5) LEU-COMP-THERM -019 LEU-COMP-THERM -021 LEU-COMP-THERM -026 (1) LEU-COMP-THERM -026 (3) LEU-COMP-THERM -026 (5)
97
were compiled and prepared on these systems. In next Table 2 main parameters of computer clusters are given. 3. Calculation The NJOY code consists of a set of modules, each performing a well-defined processing task. Each of this modules is essentially a separate computer program. In our calculation modules Moder, Broadr, Unresr, Leapr, Therm, Heatr, Purr, Gaspr, Viewr and Acer were used. Moder module converts ENDF tapes back and forth between formatted and blocked binary modes. Unresr computes effective self-shielded pointwise cross sections in the unresolved energy range. Thermr produces cross sections and energy-to-energy matrices for free or bound scatters in the thermal energy range. Heatr generates pointwise heat production cross sections and radiation-damage-production cross sections. Purr is used to prepare unresolved region probability tables for the MCNP continuous-energy Monte Carlo code. Gaspr is used to add gas production to PENDF fie. Viewr makes plots in postscript format. Acer prepares libraries in ACE format for the MCNP. (MacFarlane & Muir) All of cross section data were calculated for desired temperatures specified in the benchmark problems. The Leapr module is used to prepare the scattering law S(a,b) and related quantities, which describe thermal scattering from bound moderators, in the ENDF-6 format used by Thermr module. (MacFarlane, 1994) This module was used in combination with modules mentioned above to produce material cards for hydrogen bound in light water (H2O) for desired temperatures. Leapr requires a uniform grid for the continuous frequency distribution r(u) for
Cadmium IEU-COMP-THERM -002 (5) IEU-COMP-THERM -002 (6) Zirconium LEU-COMP-THERM-021
Table 2 Main clusters parameters. Name
Hardware
Cluster solution
Used MCNP5 compilers
CLUSTER1
12x Intel Core Duo E6850 2 3.00GHz (node) 1x Intel Core Quad Q700 2 2.40GHz (frontend)
gfortran gcc version 4.1.2 20080704
CLUSTER2
5x Intel Core Quad Q700 2 2.40GHz (frontend þ node)
Rocksþ 5.4 Maverick Open MPI 1.4.3 Red Hat 4.1.2-48 Rocks 5.1(V.I) Open MPI 1.2.7 Red Hat 4.1.2-42
gfortran gcc version 4.1.2 20071124
Table 3 Criticality benchmark results for MCNP5-1.4 and MCNP5-1.6 using ENDF/B-VII, JEFF-3.1 and JENDL-4.0 nuclear data. Case
Benchmark
mcnp5-1.4 ENDF/B-VII
IEU-COMP-THERM-002 (1) IEU-COMP-THERM-002 (2) IEU-COMP-THERM-002 (3) IEU-COMP-THERM-002 (4) IEU-COMP-THERM-002 (5) IEU-COMP-THERM-002 (6) LEU-COMP-THERM-019 (1) LEU-COMP-THERM-019 (2) LEU-COMP-THERM-019 (3) LEU-COMP-THERM-021 (1) LEU-COMP-THERM-021 (2) LEU-COMP-THERM-021 (3) LEU-COMP-THERM-021 (4) LEU-COMP-THERM-021 (5) LEU-COMP-THERM-021 (6) LEU-COMP-THERM-026 (1) LEU-COMP-THERM-026 (2) LEU-COMP-THERM-026 (3) LEU-COMP-THERM-026 (4) LEU-COMP-THERM-026 (5) LEU-COMP-THERM-026 (6)
mcnp5-1.6 JEFF-3.1
JENDL-4.0
ENDF/B-VII
JEFF-3.1
JENDL-4.0
keff
sbenchmark keff
sMCNP
keff
sMCNP
keff
sMCNP
keff
sMCNP
keff
sMCNP
keff
sMCNP
1.00140 1.00190 1.00170 1.00190 1.00140 1.00160 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 0.99960 1.00180 0.99780 1.00070 0.99830
0.00390 0.00400 0.00440 0.00440 0.00430 0.00440 0.00630 0.00580 0.00610 0.00720 0.00720 0.00720 0.00500 0.00500 0.00500 0.00340 0.00340 0.00620 0.00620 0.00410 0.00410
0.00076 0.00079 0.00070 0.00075 0.00072 0.00077 0.00120 0.00109 0.00088 0.00118 0.00117 0.00124 0.00105 0.00107 0.00107 0.00025 0.00025 0.00024 0.00024 0.00024 0.00025
0.99630 1.00394 1.00192 1.00067 0.99517 0.99475 1.01712 1.01257 1.00856 1.01525 1.01457 1.01410 1.01445 1.01268 1.01318 1.00284 1.00429 1.00446 1.00221 1.00270 1.00352
0.00073 0.00077 0.00074 0.00073 0.00082 0.00077 0.00113 0.00107 0.00091 0.00116 0.00119 0.00120 0.00105 0.00099 0.00106 0.00024 0.00026 0.00024 0.00024 0.00025 0.00025
0.99800 1.00067 1.00289 1.00118 0.99637 0.99464 1.01427 1.00812 1.00602 1.01329 1.01296 1.01283 1.01494 1.01588 1.01496 1.00196 1.00179 1.00528 1.00117 1.00312 1.00190
0.00073 0.00082 0.00074 0.00080 0.00077 0.00077 0.00111 0.00121 0.00089 0.00114 0.00121 0.00122 0.00107 0.00102 0.00105 0.00025 0.00025 0.00025 0.00024 0.00024 0.00025
0.99902 1.00188 1.00537 1.00167 0.99751 0.99689 1.01560 1.01034 1.00715 1.01300 1.01237 1.01224 1.01178 1.01356 1.01352 1.00310 1.00314 1.00575 1.00173 1.00414 1.00380
0.00074 0.00078 0.00073 0.00076 0.00073 0.00073 0.00111 0.00111 0.00085 0.00122 0.00123 0.00121 0.00105 0.00106 0.00099 0.00025 0.00025 0.00025 0.00024 0.00024 0.00025
0.99754 1.00207 1.00198 1.00153 0.99580 0.99445 1.01532 1.01219 1.00770 1.01262 1.01212 1.01410 1.01455 1.01442 1.0117 1.00216 1.00423 1.00455 1.00209 1.00302 1.00381
0.00077 0.00075 0.00075 0.00075 0.00072 0.00074 0.00111 0.00112 0.00092 0.00117 0.00116 0.00120 0.00093 0.00106 0.00108 0.00026 0.00025 0.00024 0.00023 0.00025 0.00025
0.99761 0.99907 1.00233 1.00009 0.99673 0.99514 1.01441 1.01058 1.00746 1.01039 1.01289 1.01367 1.01338 1.01559 1.01496 1.00196 1.00207 1.00479 1.00053 1.00316 1.00180
0.00077 0.00082 0.00073 0.00074 0.00069 0.00075 0.00110 0.00113 0.00089 0.00126 0.00127 0.00121 0.00107 0.00099 0.00105 0.00026 0.00026 0.00025 0.00024 0.00024 0.00025
1.00016 1.00092 1.00360 1.00385 0.99953 0.99600 1.01554 1.00977 1.00799 1.01121 1.01509 1.01431 1.01253 1.01371 1.01406 1.00289 1.00376 1.00631 1.00201 1.00446 1.00360
98
J. Lülely et al. / Progress in Nuclear Energy 59 (2012) 96e99
Table 4 Criticality benchmark results for MCNP5-1.6 using material cards prepared by different approaches. mcnp5-1.6
ENDF/B-VII
Jeff-3.1
Approach 1
Approach 3
Case
keff
s
keff
s
IEU-COMP-THERM-002 (1) IEU-COMP-THERM-002 (2) IEU-COMP-THERM-002 (3) IEU-COMP-THERM-002 (4) IEU-COMP-THERM-002 (5) IEU-COMP-THERM-002 (6) LEU-COMP-THERM-019 (1) LEU-COMP-THERM-019 (2) LEU-COMP-THERM-019 (3) LEU-COMP-THERM-021 (1) LEU-COMP-THERM-021 (2) LEU-COMP-THERM-021 (3) LEU-COMP-THERM-021 (4) LEU-COMP-THERM-021 (5) LEU-COMP-THERM-021 (6) LEU-COMP-THERM-026 (1) LEU-COMP-THERM-026 (2) LEU-COMP-THERM-026 (3) LEU-COMP-THERM-026 (4) LEU-COMP-THERM-026 (5) LEU-COMP-THERM-026 (6)
0.99753 1.00038 1.00231 0.99967 0.99556 0.99621 1.01372 1.00784 1.00376 1.01107 1.01483 1.0137 1.01404 1.0137 1.01221 1.00276 1.00289 1.00535 1.00127 1.00336 1.00185
0.00076 0.00077 0.00076 0.00078 0.00077 0.00076 0.00111 0.00116 0.00088 0.00117 0.00118 0.00119 0.00116 0.00105 0.00105 0.00026 0.00026 0.00025 0.00024 0.00025 0.00025
0.99850 1.00052 1.00186 1.00170 0.99672 0.99378 1.01584 1.01023 1.00768 1.01316 1.01318 1.01238 1.01204 1.01445 1.01411 1.00292 1.00235 1.00512 1.00083 1.00370 1.00199
0.00073 0.00080 0.00073 0.00074 0.00073 0.00077 0.00107 0.00111 0.00091 0.00106 0.00118 0.00116 0.00102 0.00104 0.00106 0.00026 0.00024 0.00024 0.00024 0.00025 0.00025
MCNP
Jendl-4.0
Approach 1 MCNP
Approach 3
Approach 1
Approach 3
keff
s
keff
s
keff
s
0.99831 1.00096 1.00090 0.99975 0.99637 0.99575 1.01742 1.01053 1.0075 1.01204 1.01414 1.01035 1.01317 1.01393 1.01169 1.00311 1.00193 1.00511 1.00045 1.00427 1.00259
0.00074 0.00079 0.00073 0.00073 0.00077 0.00075 0.00115 0.00108 0.00093 0.00121 0.00117 0.00124 0.00109 0.00104 0.00105 0.00025 0.00025 0.00024 0.00024 0.00025 0.00025
0.99889 0.99993 1.00227 1.00029 0.99751 0.99490 1.01692 1.00940 1.00716 1.01397 1.01085 1.01327 1.01389 1.01467 1.01317 1.00313 1.00261 1.00488 1.00086 1.00357 1.00202
0.00078 0.00081 0.00077 0.00074 0.00076 0.00069 0.00110 0.00112 0.00087 0.00125 0.00122 0.00123 0.00103 0.00105 0.00114 0.00025 0.00025 0.00026 0.00023 0.00024 0.00024
0.99821 0.99982 1.00086 1.00083 0.99590 0.99518 1.01565 1.01161 1.00630 1.01308 1.01287 1.01432 1.01375 1.01379 1.01218 1.00345 1.00279 1.00559 1.00192 1.00338 1.00235
0.00075 0.00079 0.00072 0.00071 0.00075 0.00075 0.00108 0.00113 0.00089 0.00123 0.00119 0.00122 0.00108 0.00105 0.00111 0.00026 0.00026 0.00026 0.00023 0.00026 0.00024
every temperature. The final frequency spectra for the range of temperatures are based on (Mattes and Keinert, 2005). Three approaches were used to produce material cards. The first approach is based on the use of module Leapr with the different frequency distributions. In second approach we produced material cards by using pre-calculated scattering law in the ENDF-6 format which is distributed with cross section data files. All of these two approaches have constraints in numbers of pre-calculated data, therefore closest values were used. The third approach used the short definition of Leapr module where frequency distribution can be set only by the first temperature.(MacFarlane, 1994) All benchmark problems were calculated for each material card. MCNP calculations for IEU-COMP-THERM-002 problem were performed with 520 generations of 2000 neutrons each, and the results from the first 10 generations were discarded. Consequently,
MCNP
MCNP
MCNP
keff
sMCNP
0.99994 0.99991 1.00209 1.00215 0.99676 0.99486 1.01344 1.00913 1.0070 1.01609 1.01211 1.01159 1.01475 1.01364 1.01270 1.00315 1.00287 1.00530 1.00149 1.00387 1.00268
0.00074 0.00078 0.00074 0.00076 0.00075 0.00077 0.00110 0.00112 0.00086 0.00120 0.00125 0.00124 0.00114 0.00111 0.00097 0.00025 0.00026 0.00025 0.00024 0.00025 0.00024
the results for each case are based on 1,020,000 active neutron histories. For the LEU-COMP-THERM-019 and LEU-COMP-THERM021 problems 550 generations of 1000 neutrons each were used, and 50 generations were discarded. In total 500,000 active neutrons were produced. In LEU-COMP-THERM-026 problem 1020 generations of 10,000 neutrons each were generated. 10 generations were discarded and total number of 1,010,000 active neutrons paths were calculated. Statistical approach to results data processing was chosen because of their robustness. Every partial bias and uncertainty for each benchmark case was evaluated using terms: rffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi 2 2 ffi sbenchmark biasi ¼ kbenchmark kMCNP ; si ¼ þ sMCNP ; i i i i
si ¼ f1.21g In this terms symbol k stands for k effective. From these partial biases average value labeled biascomb (combined bias)
Table 5 Bias and uncertainty for ENDF/B-VII nuclear data and MCNP5-1.6. Case
IEU-COMP-THERM-002 (1) IEU-COMP-THERM-002 (2) IEU-COMP-THERM-002 (3) IEU-COMP-THERM-002 (4) IEU-COMP-THERM-002 (5) IEU-COMP-THERM-002 (6) LEU-COMP-THERM-019 (1) LEU-COMP-THERM-019 (2) LEU-COMP-THERM-019 (3) LEU-COMP-THERM-021 (1) LEU-COMP-THERM-021 (2) LEU-COMP-THERM-021 (3) LEU-COMP-THERM-021 (4) LEU-COMP-THERM-021 (5) LEU-COMP-THERM-021 (6) LEU-COMP-THERM-026 (1) LEU-COMP-THERM-026 (2) LEU-COMP-THERM-026 (3) LEU-COMP-THERM-026 (4) LEU-COMP-THERM-026 (5) LEU-COMP-THERM-026 (6)
Approach 1
Approach 2
Approach 3
bias
s
bias
s
bias
s
0.00387 0.00152 0.00061 0.00223 0.00584 0.00539 0.01372 0.00784 0.00376 0.01107 0.01483 0.0137 0.01404 0.0137 0.01221 0.00276 0.00329 0.00355 0.00347 0.0027 0.0036
0.00397 0.00407 0.00447 0.00447 0.00437 0.00447 0.00640 0.00591 0.00616 0.00729 0.00730 0.00730 0.00513 0.00511 0.00511 0.00341 0.00341 0.00621 0.00620 0.00411 0.00411
0.00238 0.00002 0.00367 0.00023 0.00389 0.00471 0.01560 0.01034 0.00715 0.01300 0.01237 0.01224 0.01178 0.01356 0.01352 0.00310 0.00354 0.00395 0.00393 0.00344 0.0055
0.00397 0.00408 0.00446 0.00447 0.00436 0.00446 0.00640 0.00591 0.00616 0.00730 0.00730 0.00730 0.00511 0.00511 0.00510 0.00341 0.00341 0.00621 0.00620 0.00411 0.00411
0.00290 0.00138 0.00016 0.00020 0.00468 0.00782 0.01584 0.01023 0.00768 0.01316 0.01318 0.01238 0.01204 0.01445 0.01411 0.00292 0.00275 0.00332 0.00303 0.00300 0.0037
0.00397 0.00408 0.00446 0.00446 0.00436 0.00447 0.00639 0.00591 0.00617 0.00728 0.00730 0.00729 0.00510 0.00511 0.00511 0.00341 0.00341 0.00620 0.00620 0.00411 0.00411
J. Lülely et al. / Progress in Nuclear Energy 59 (2012) 96e99 Table 6 Combined bias and combined uncertainty for each nuclear data set and used material cards. Approach
ENDF/B-VII 1 2 3 JEFF-3.1 1 2 3 JENDL-4.0 1 2 3
mcnp5-1.4
mcnp5-1.6
biascomb
scomb
biascomb
scomb
0.004940476 0.00511767 0.005237143
0.025867975 0.025999238 0.026033796
0.00493095 0.00489333 0.00503381
0.025838934 0.025882309 0.025954867
0.0056510 0.0060548 0.0058157
0.0260713 0.0262340 0.0260809
0.0053010 0.0057052 0.0055267
0.0259099 0.0260235 0.0259830
0.0053367 0.0054352 0.0052524
0.0259829 0.0260036 0.0258720
0.0055110 0.0052624 0.0055724
0.0260014 0.0259684 0.0259197
was calculated. Combined uncertainty was also determined by term: qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi scomb ¼ ðs21 þ . þ s221 þ s2bias Þ, where sbias is standard deviation of the average bias. This combined bias with combined uncertainty represents the total bias of a whole group of specified benchmark problems.
99
5. Conclusions The calculations described in the current report constitute from a variety problems investigated at our Institute. The main goal was to determine the combined values of biases and appropriate uncertainties for different libraries and for our computer cluster systems. Results are presented in Table 6. In addition, we have provided some initial indication of the moving from MCNP5-1.40 to MCNP5-1.60 system. We can conclude, that these results are slightly identical. Three methods of production material cards for hydrogen bound in light water were studied and compared each other as described in Tables 3 and 4. Going forward the most appropriate bias related to the specific solved problem should be used. In the future real criticality experiment results from VVER-440 unit commissioning will be used for MCNP5-1.60 code validation and evaluation of bias value. This bias will be compared and correlated with biases coming from the International Handbook of Evaluated Criticality Benchmark Experiments and biases presented in this paper. Acknowledgements This work has been done through the support of KEGA number 3/7241/09.
4. Results
References
All 21 benchmark problems were calculated for each cross section library in combination with different type of material cards on both clusters using MCNP5-1.4 and MCNP5-1.6. Next tables show results from calculations on Cluster 1. The results from Cluster 2 are identical. Table 3 presents results calculated using both code version with ENDF/B-VII, JEFF-3.1 and JENDL-4.0 nuclear data. Material cards used in this calculations were prepared from precalculate data files in ENDF-6 format. Table 4 shows the results where material cards were prepared by module Leapr in two ways, precisely defined frequency distribution for each temperature and short version. For each computed case bias and uncertainty were determined (Table 5) and finally combined bias and combined uncertainty were determined for whole group of benchmark problems for each nuclear data set and used material cards (Table 6).
ENDF library - Evaluated Data Libraries including INDF/B-VII.0, ordered from IAEA http://www-nds.iaea.org/cd-catalog.html, (April, 2011). International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, 2007. OECD Nuclear Energy Agency. JEFF 3.1 General Purpose Library, ordered from IAEA http://www-nds.iaea.org/cdcatalog.html, (May, 2005). JENDL 4.0-Japanese Evaluated Nuclear Data Library version 4., ordered from IAEA http://www-nds.iaea.org/cd-catalog.html, (2010). E. MacFarlane, D. W. Muir: RSICC Peripheral Shielding Routine Collection NJOY99.0, Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data, Los Alamos National Laboratory, Los Alamos, New Mexico. MacFarlane, R.E., March 1994. New Thermal Neutron Scattering Files for ENDF/B-VI Release 2. LA-12639-MS (ENDF 356). Mattes, M., Keinert, J., April 2005. Institute for Nuclear Technology and Energy Systems (IKE Thermal Neutron Scattering Data for the Moderator Materials H2O, D2O and ZrHx in ENDF-6 Format and as ACE Library for MCNP(X) Codes. INDC(NDS)-0470. X-5 Monte Carlo Team, April, 2003. MCNP e A General N-Particle Transport Code, Version 5. In: Overview and Theory, LA-UR-03-1987, vol. I. Los Alamos National Laboratory.
Contents lists available at SciVerse ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Determination of criticality safety MCNP5 calculation bias by using different libraries of cross section data Jakub Lülely, Branislav Vrban*, Gabriel Farkas, Ján Has cík, Martin Petriska Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovicova 3, 812 19 Bratislava, Slovakia
a r t i c l e i n f o
a b s t r a c t
Article history: Received 12 October 2011 Accepted 23 April 2012
Application of different cross section libraries and different versions of Monte Carlo code MCNP has an influence on the calculation results and therefore determination of criticality safety calculation bias forms part of improving accuracy of simulations using computational systems and codes. In this paper, criticality calculations results are presented for 21 problems coming from the International Handbook of Evaluated Criticality Benchmark Experiments (International Handbook, 2007). All of these problems are related to VVER-440 reactors because of their extensive use in Slovakia. Three libraries of cross section data were investigated:
Keywords: Bias VVER-440 MCNP Cross section Benchmark
JEFF-3.1 General purpose library, ENDF/B-VII library, JENDL 4.0.
Calculations were provided with MCNP5-1.40 and MCNP5-1.60 transport codes. Two cluster systems situated at our Institute were used. Main purpose of this analyses was the determination of the bias which should be used in further simulations. Ó 2012 Elsevier Ltd. All rights reserved.
1. Introduction
2. Material and methods
Determination of criticality safety calculation bias caused by use of different libraries of cross section data and different versions of stochastic Monte Carlo N-Particle Transport Code (X-5 Monte Carlo Team, 2003) plays an essential role in process of tuning used calculation systems. These systems should be needed in analyses of diverse nuclear problems. In this paper we are presenting obtained criticality calculations results for 21 problems coming from the International Handbook of Evaluated Criticality Benchmark Experiments (International Handbook, 2007). Three sets of calculations were performed using JEFF-3.1 General purpose library (JEFF, 2005), ENDF libraries (ENDF, 2011) and Japanese Evaluated Nuclear Data Library version 4.0 (JENDL, 2010). Versions MCNP5-1.40 and MCNP5-1.60 were investigated. All these calculations have been provided on two computer clusters used at our Institute. Main purpose of this analysis was the determination and specification of the best bias which should be used in further calculations at Institute of Nuclear and Physical Engineering.
Problems coming from the International Handbook were chosen in regard to analyses performed at our Institute. Only VVER-440 reactors are situated and operated in Slovakia, so all of these problems are connected to this type of reactor and should be useful for our purposes. We tried to specify benchmarks problems at different temperatures focused on use of low enrichment uranium oxide as a fuel, light water or light water with boric acid as a moderator, gadolinium or cadmium as absorber, stainless steel or zirconium as a clad of fuel rods and use of hexagonal pitched lattices. Four main problems are presented. Each of them consists of another cases. Enrichment of 235U is at the level of 17.4% in IEU-COMP-THERM-002, 5.4% in LEU-COMP-THERM-019 and in LEU-COMP-THERM-021, and finally 4.924% in problem LEU-COMPTHERM-026. Further subdivision of problems can be found in Table 1. All of different libraries of cross section data in ACE format needed for MCNP5 were calculated by cross section processing program NJOY 99.364. (MacFarlane and Muir) At our Institute, two computer clusters for MCNP calculations are in operation. Cluster versions of MCNP5-1.40 and MCNP5-1.60
* Corresponding author. Tel.: þ421 260291504. E-mail address: [email protected] (B. Vrban). 0149-1970/$ e see front matter Ó 2012 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2012.04.007
J. Lülely et al. / Progress in Nuclear Energy 59 (2012) 96e99 Table 1 Specified problems coming from the International Handbook of Evaluated Criticality Benchmark Experiments (International Handbook, 2007). Spectrum
Thermal
Moderator
Pure light water IEU-COMP-THERM-002 LEU-COMP-THERM-019 LEU-COMP-THERM-026 Gadolinium IEU-COMP-THERM -002 (3) IEU-COMP-THERM -002 (4) Stainless steel IEU-COMP-THERM-002 LEU-COMP-THERM-019 LEU-COMP-THERM-026
Boric acid water LEU-COMP-THERM-021
400 K
500 K
IEU-COMP-THERM -002 (4) IEU-COMP-THERM -002 (6)
IEU-COMP-THERM -002 (2) LEU-COMP-THERM -026 (2) LEU-COMP-THERM -026 (4) LEU-COMP-THERM -026 (6)
Absorber
Clad
Temperatures 300 K IEU-COMP-THERM -002 (1) IEU-COMP-THERM -002 (3) IEU-COMP-THERM -002 (5) LEU-COMP-THERM -019 LEU-COMP-THERM -021 LEU-COMP-THERM -026 (1) LEU-COMP-THERM -026 (3) LEU-COMP-THERM -026 (5)
97
were compiled and prepared on these systems. In next Table 2 main parameters of computer clusters are given. 3. Calculation The NJOY code consists of a set of modules, each performing a well-defined processing task. Each of this modules is essentially a separate computer program. In our calculation modules Moder, Broadr, Unresr, Leapr, Therm, Heatr, Purr, Gaspr, Viewr and Acer were used. Moder module converts ENDF tapes back and forth between formatted and blocked binary modes. Unresr computes effective self-shielded pointwise cross sections in the unresolved energy range. Thermr produces cross sections and energy-to-energy matrices for free or bound scatters in the thermal energy range. Heatr generates pointwise heat production cross sections and radiation-damage-production cross sections. Purr is used to prepare unresolved region probability tables for the MCNP continuous-energy Monte Carlo code. Gaspr is used to add gas production to PENDF fie. Viewr makes plots in postscript format. Acer prepares libraries in ACE format for the MCNP. (MacFarlane & Muir) All of cross section data were calculated for desired temperatures specified in the benchmark problems. The Leapr module is used to prepare the scattering law S(a,b) and related quantities, which describe thermal scattering from bound moderators, in the ENDF-6 format used by Thermr module. (MacFarlane, 1994) This module was used in combination with modules mentioned above to produce material cards for hydrogen bound in light water (H2O) for desired temperatures. Leapr requires a uniform grid for the continuous frequency distribution r(u) for
Cadmium IEU-COMP-THERM -002 (5) IEU-COMP-THERM -002 (6) Zirconium LEU-COMP-THERM-021
Table 2 Main clusters parameters. Name
Hardware
Cluster solution
Used MCNP5 compilers
CLUSTER1
12x Intel Core Duo E6850 2 3.00GHz (node) 1x Intel Core Quad Q700 2 2.40GHz (frontend)
gfortran gcc version 4.1.2 20080704
CLUSTER2
5x Intel Core Quad Q700 2 2.40GHz (frontend þ node)
Rocksþ 5.4 Maverick Open MPI 1.4.3 Red Hat 4.1.2-48 Rocks 5.1(V.I) Open MPI 1.2.7 Red Hat 4.1.2-42
gfortran gcc version 4.1.2 20071124
Table 3 Criticality benchmark results for MCNP5-1.4 and MCNP5-1.6 using ENDF/B-VII, JEFF-3.1 and JENDL-4.0 nuclear data. Case
Benchmark
mcnp5-1.4 ENDF/B-VII
IEU-COMP-THERM-002 (1) IEU-COMP-THERM-002 (2) IEU-COMP-THERM-002 (3) IEU-COMP-THERM-002 (4) IEU-COMP-THERM-002 (5) IEU-COMP-THERM-002 (6) LEU-COMP-THERM-019 (1) LEU-COMP-THERM-019 (2) LEU-COMP-THERM-019 (3) LEU-COMP-THERM-021 (1) LEU-COMP-THERM-021 (2) LEU-COMP-THERM-021 (3) LEU-COMP-THERM-021 (4) LEU-COMP-THERM-021 (5) LEU-COMP-THERM-021 (6) LEU-COMP-THERM-026 (1) LEU-COMP-THERM-026 (2) LEU-COMP-THERM-026 (3) LEU-COMP-THERM-026 (4) LEU-COMP-THERM-026 (5) LEU-COMP-THERM-026 (6)
mcnp5-1.6 JEFF-3.1
JENDL-4.0
ENDF/B-VII
JEFF-3.1
JENDL-4.0
keff
sbenchmark keff
sMCNP
keff
sMCNP
keff
sMCNP
keff
sMCNP
keff
sMCNP
keff
sMCNP
1.00140 1.00190 1.00170 1.00190 1.00140 1.00160 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 1.00000 0.99960 1.00180 0.99780 1.00070 0.99830
0.00390 0.00400 0.00440 0.00440 0.00430 0.00440 0.00630 0.00580 0.00610 0.00720 0.00720 0.00720 0.00500 0.00500 0.00500 0.00340 0.00340 0.00620 0.00620 0.00410 0.00410
0.00076 0.00079 0.00070 0.00075 0.00072 0.00077 0.00120 0.00109 0.00088 0.00118 0.00117 0.00124 0.00105 0.00107 0.00107 0.00025 0.00025 0.00024 0.00024 0.00024 0.00025
0.99630 1.00394 1.00192 1.00067 0.99517 0.99475 1.01712 1.01257 1.00856 1.01525 1.01457 1.01410 1.01445 1.01268 1.01318 1.00284 1.00429 1.00446 1.00221 1.00270 1.00352
0.00073 0.00077 0.00074 0.00073 0.00082 0.00077 0.00113 0.00107 0.00091 0.00116 0.00119 0.00120 0.00105 0.00099 0.00106 0.00024 0.00026 0.00024 0.00024 0.00025 0.00025
0.99800 1.00067 1.00289 1.00118 0.99637 0.99464 1.01427 1.00812 1.00602 1.01329 1.01296 1.01283 1.01494 1.01588 1.01496 1.00196 1.00179 1.00528 1.00117 1.00312 1.00190
0.00073 0.00082 0.00074 0.00080 0.00077 0.00077 0.00111 0.00121 0.00089 0.00114 0.00121 0.00122 0.00107 0.00102 0.00105 0.00025 0.00025 0.00025 0.00024 0.00024 0.00025
0.99902 1.00188 1.00537 1.00167 0.99751 0.99689 1.01560 1.01034 1.00715 1.01300 1.01237 1.01224 1.01178 1.01356 1.01352 1.00310 1.00314 1.00575 1.00173 1.00414 1.00380
0.00074 0.00078 0.00073 0.00076 0.00073 0.00073 0.00111 0.00111 0.00085 0.00122 0.00123 0.00121 0.00105 0.00106 0.00099 0.00025 0.00025 0.00025 0.00024 0.00024 0.00025
0.99754 1.00207 1.00198 1.00153 0.99580 0.99445 1.01532 1.01219 1.00770 1.01262 1.01212 1.01410 1.01455 1.01442 1.0117 1.00216 1.00423 1.00455 1.00209 1.00302 1.00381
0.00077 0.00075 0.00075 0.00075 0.00072 0.00074 0.00111 0.00112 0.00092 0.00117 0.00116 0.00120 0.00093 0.00106 0.00108 0.00026 0.00025 0.00024 0.00023 0.00025 0.00025
0.99761 0.99907 1.00233 1.00009 0.99673 0.99514 1.01441 1.01058 1.00746 1.01039 1.01289 1.01367 1.01338 1.01559 1.01496 1.00196 1.00207 1.00479 1.00053 1.00316 1.00180
0.00077 0.00082 0.00073 0.00074 0.00069 0.00075 0.00110 0.00113 0.00089 0.00126 0.00127 0.00121 0.00107 0.00099 0.00105 0.00026 0.00026 0.00025 0.00024 0.00024 0.00025
1.00016 1.00092 1.00360 1.00385 0.99953 0.99600 1.01554 1.00977 1.00799 1.01121 1.01509 1.01431 1.01253 1.01371 1.01406 1.00289 1.00376 1.00631 1.00201 1.00446 1.00360
98
J. Lülely et al. / Progress in Nuclear Energy 59 (2012) 96e99
Table 4 Criticality benchmark results for MCNP5-1.6 using material cards prepared by different approaches. mcnp5-1.6
ENDF/B-VII
Jeff-3.1
Approach 1
Approach 3
Case
keff
s
keff
s
IEU-COMP-THERM-002 (1) IEU-COMP-THERM-002 (2) IEU-COMP-THERM-002 (3) IEU-COMP-THERM-002 (4) IEU-COMP-THERM-002 (5) IEU-COMP-THERM-002 (6) LEU-COMP-THERM-019 (1) LEU-COMP-THERM-019 (2) LEU-COMP-THERM-019 (3) LEU-COMP-THERM-021 (1) LEU-COMP-THERM-021 (2) LEU-COMP-THERM-021 (3) LEU-COMP-THERM-021 (4) LEU-COMP-THERM-021 (5) LEU-COMP-THERM-021 (6) LEU-COMP-THERM-026 (1) LEU-COMP-THERM-026 (2) LEU-COMP-THERM-026 (3) LEU-COMP-THERM-026 (4) LEU-COMP-THERM-026 (5) LEU-COMP-THERM-026 (6)
0.99753 1.00038 1.00231 0.99967 0.99556 0.99621 1.01372 1.00784 1.00376 1.01107 1.01483 1.0137 1.01404 1.0137 1.01221 1.00276 1.00289 1.00535 1.00127 1.00336 1.00185
0.00076 0.00077 0.00076 0.00078 0.00077 0.00076 0.00111 0.00116 0.00088 0.00117 0.00118 0.00119 0.00116 0.00105 0.00105 0.00026 0.00026 0.00025 0.00024 0.00025 0.00025
0.99850 1.00052 1.00186 1.00170 0.99672 0.99378 1.01584 1.01023 1.00768 1.01316 1.01318 1.01238 1.01204 1.01445 1.01411 1.00292 1.00235 1.00512 1.00083 1.00370 1.00199
0.00073 0.00080 0.00073 0.00074 0.00073 0.00077 0.00107 0.00111 0.00091 0.00106 0.00118 0.00116 0.00102 0.00104 0.00106 0.00026 0.00024 0.00024 0.00024 0.00025 0.00025
MCNP
Jendl-4.0
Approach 1 MCNP
Approach 3
Approach 1
Approach 3
keff
s
keff
s
keff
s
0.99831 1.00096 1.00090 0.99975 0.99637 0.99575 1.01742 1.01053 1.0075 1.01204 1.01414 1.01035 1.01317 1.01393 1.01169 1.00311 1.00193 1.00511 1.00045 1.00427 1.00259
0.00074 0.00079 0.00073 0.00073 0.00077 0.00075 0.00115 0.00108 0.00093 0.00121 0.00117 0.00124 0.00109 0.00104 0.00105 0.00025 0.00025 0.00024 0.00024 0.00025 0.00025
0.99889 0.99993 1.00227 1.00029 0.99751 0.99490 1.01692 1.00940 1.00716 1.01397 1.01085 1.01327 1.01389 1.01467 1.01317 1.00313 1.00261 1.00488 1.00086 1.00357 1.00202
0.00078 0.00081 0.00077 0.00074 0.00076 0.00069 0.00110 0.00112 0.00087 0.00125 0.00122 0.00123 0.00103 0.00105 0.00114 0.00025 0.00025 0.00026 0.00023 0.00024 0.00024
0.99821 0.99982 1.00086 1.00083 0.99590 0.99518 1.01565 1.01161 1.00630 1.01308 1.01287 1.01432 1.01375 1.01379 1.01218 1.00345 1.00279 1.00559 1.00192 1.00338 1.00235
0.00075 0.00079 0.00072 0.00071 0.00075 0.00075 0.00108 0.00113 0.00089 0.00123 0.00119 0.00122 0.00108 0.00105 0.00111 0.00026 0.00026 0.00026 0.00023 0.00026 0.00024
every temperature. The final frequency spectra for the range of temperatures are based on (Mattes and Keinert, 2005). Three approaches were used to produce material cards. The first approach is based on the use of module Leapr with the different frequency distributions. In second approach we produced material cards by using pre-calculated scattering law in the ENDF-6 format which is distributed with cross section data files. All of these two approaches have constraints in numbers of pre-calculated data, therefore closest values were used. The third approach used the short definition of Leapr module where frequency distribution can be set only by the first temperature.(MacFarlane, 1994) All benchmark problems were calculated for each material card. MCNP calculations for IEU-COMP-THERM-002 problem were performed with 520 generations of 2000 neutrons each, and the results from the first 10 generations were discarded. Consequently,
MCNP
MCNP
MCNP
keff
sMCNP
0.99994 0.99991 1.00209 1.00215 0.99676 0.99486 1.01344 1.00913 1.0070 1.01609 1.01211 1.01159 1.01475 1.01364 1.01270 1.00315 1.00287 1.00530 1.00149 1.00387 1.00268
0.00074 0.00078 0.00074 0.00076 0.00075 0.00077 0.00110 0.00112 0.00086 0.00120 0.00125 0.00124 0.00114 0.00111 0.00097 0.00025 0.00026 0.00025 0.00024 0.00025 0.00024
the results for each case are based on 1,020,000 active neutron histories. For the LEU-COMP-THERM-019 and LEU-COMP-THERM021 problems 550 generations of 1000 neutrons each were used, and 50 generations were discarded. In total 500,000 active neutrons were produced. In LEU-COMP-THERM-026 problem 1020 generations of 10,000 neutrons each were generated. 10 generations were discarded and total number of 1,010,000 active neutrons paths were calculated. Statistical approach to results data processing was chosen because of their robustness. Every partial bias and uncertainty for each benchmark case was evaluated using terms: rffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi 2 2 ffi sbenchmark biasi ¼ kbenchmark kMCNP ; si ¼ þ sMCNP ; i i i i
si ¼ f1.21g In this terms symbol k stands for k effective. From these partial biases average value labeled biascomb (combined bias)
Table 5 Bias and uncertainty for ENDF/B-VII nuclear data and MCNP5-1.6. Case
IEU-COMP-THERM-002 (1) IEU-COMP-THERM-002 (2) IEU-COMP-THERM-002 (3) IEU-COMP-THERM-002 (4) IEU-COMP-THERM-002 (5) IEU-COMP-THERM-002 (6) LEU-COMP-THERM-019 (1) LEU-COMP-THERM-019 (2) LEU-COMP-THERM-019 (3) LEU-COMP-THERM-021 (1) LEU-COMP-THERM-021 (2) LEU-COMP-THERM-021 (3) LEU-COMP-THERM-021 (4) LEU-COMP-THERM-021 (5) LEU-COMP-THERM-021 (6) LEU-COMP-THERM-026 (1) LEU-COMP-THERM-026 (2) LEU-COMP-THERM-026 (3) LEU-COMP-THERM-026 (4) LEU-COMP-THERM-026 (5) LEU-COMP-THERM-026 (6)
Approach 1
Approach 2
Approach 3
bias
s
bias
s
bias
s
0.00387 0.00152 0.00061 0.00223 0.00584 0.00539 0.01372 0.00784 0.00376 0.01107 0.01483 0.0137 0.01404 0.0137 0.01221 0.00276 0.00329 0.00355 0.00347 0.0027 0.0036
0.00397 0.00407 0.00447 0.00447 0.00437 0.00447 0.00640 0.00591 0.00616 0.00729 0.00730 0.00730 0.00513 0.00511 0.00511 0.00341 0.00341 0.00621 0.00620 0.00411 0.00411
0.00238 0.00002 0.00367 0.00023 0.00389 0.00471 0.01560 0.01034 0.00715 0.01300 0.01237 0.01224 0.01178 0.01356 0.01352 0.00310 0.00354 0.00395 0.00393 0.00344 0.0055
0.00397 0.00408 0.00446 0.00447 0.00436 0.00446 0.00640 0.00591 0.00616 0.00730 0.00730 0.00730 0.00511 0.00511 0.00510 0.00341 0.00341 0.00621 0.00620 0.00411 0.00411
0.00290 0.00138 0.00016 0.00020 0.00468 0.00782 0.01584 0.01023 0.00768 0.01316 0.01318 0.01238 0.01204 0.01445 0.01411 0.00292 0.00275 0.00332 0.00303 0.00300 0.0037
0.00397 0.00408 0.00446 0.00446 0.00436 0.00447 0.00639 0.00591 0.00617 0.00728 0.00730 0.00729 0.00510 0.00511 0.00511 0.00341 0.00341 0.00620 0.00620 0.00411 0.00411
J. Lülely et al. / Progress in Nuclear Energy 59 (2012) 96e99 Table 6 Combined bias and combined uncertainty for each nuclear data set and used material cards. Approach
ENDF/B-VII 1 2 3 JEFF-3.1 1 2 3 JENDL-4.0 1 2 3
mcnp5-1.4
mcnp5-1.6
biascomb
scomb
biascomb
scomb
0.004940476 0.00511767 0.005237143
0.025867975 0.025999238 0.026033796
0.00493095 0.00489333 0.00503381
0.025838934 0.025882309 0.025954867
0.0056510 0.0060548 0.0058157
0.0260713 0.0262340 0.0260809
0.0053010 0.0057052 0.0055267
0.0259099 0.0260235 0.0259830
0.0053367 0.0054352 0.0052524
0.0259829 0.0260036 0.0258720
0.0055110 0.0052624 0.0055724
0.0260014 0.0259684 0.0259197
was calculated. Combined uncertainty was also determined by term: qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi scomb ¼ ðs21 þ . þ s221 þ s2bias Þ, where sbias is standard deviation of the average bias. This combined bias with combined uncertainty represents the total bias of a whole group of specified benchmark problems.
99
5. Conclusions The calculations described in the current report constitute from a variety problems investigated at our Institute. The main goal was to determine the combined values of biases and appropriate uncertainties for different libraries and for our computer cluster systems. Results are presented in Table 6. In addition, we have provided some initial indication of the moving from MCNP5-1.40 to MCNP5-1.60 system. We can conclude, that these results are slightly identical. Three methods of production material cards for hydrogen bound in light water were studied and compared each other as described in Tables 3 and 4. Going forward the most appropriate bias related to the specific solved problem should be used. In the future real criticality experiment results from VVER-440 unit commissioning will be used for MCNP5-1.60 code validation and evaluation of bias value. This bias will be compared and correlated with biases coming from the International Handbook of Evaluated Criticality Benchmark Experiments and biases presented in this paper. Acknowledgements This work has been done through the support of KEGA number 3/7241/09.
4. Results
References
All 21 benchmark problems were calculated for each cross section library in combination with different type of material cards on both clusters using MCNP5-1.4 and MCNP5-1.6. Next tables show results from calculations on Cluster 1. The results from Cluster 2 are identical. Table 3 presents results calculated using both code version with ENDF/B-VII, JEFF-3.1 and JENDL-4.0 nuclear data. Material cards used in this calculations were prepared from precalculate data files in ENDF-6 format. Table 4 shows the results where material cards were prepared by module Leapr in two ways, precisely defined frequency distribution for each temperature and short version. For each computed case bias and uncertainty were determined (Table 5) and finally combined bias and combined uncertainty were determined for whole group of benchmark problems for each nuclear data set and used material cards (Table 6).
ENDF library - Evaluated Data Libraries including INDF/B-VII.0, ordered from IAEA http://www-nds.iaea.org/cd-catalog.html, (April, 2011). International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, 2007. OECD Nuclear Energy Agency. JEFF 3.1 General Purpose Library, ordered from IAEA http://www-nds.iaea.org/cdcatalog.html, (May, 2005). JENDL 4.0-Japanese Evaluated Nuclear Data Library version 4., ordered from IAEA http://www-nds.iaea.org/cd-catalog.html, (2010). E. MacFarlane, D. W. Muir: RSICC Peripheral Shielding Routine Collection NJOY99.0, Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data, Los Alamos National Laboratory, Los Alamos, New Mexico. MacFarlane, R.E., March 1994. New Thermal Neutron Scattering Files for ENDF/B-VI Release 2. LA-12639-MS (ENDF 356). Mattes, M., Keinert, J., April 2005. Institute for Nuclear Technology and Energy Systems (IKE Thermal Neutron Scattering Data for the Moderator Materials H2O, D2O and ZrHx in ENDF-6 Format and as ACE Library for MCNP(X) Codes. INDC(NDS)-0470. X-5 Monte Carlo Team, April, 2003. MCNP e A General N-Particle Transport Code, Version 5. In: Overview and Theory, LA-UR-03-1987, vol. I. Los Alamos National Laboratory.