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Journal Articles

A Pseudo-material method for graphite with arbitrary porosities in Monte Carlo criticality calculations

Okita, Shoichiro; Nagaya, Yasunobu; Fukaya, Yuji

Journal of Nuclear Science and Technology, 58(9), p.992 - 998, 2021/09

 Times Cited Count:1 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Nuclear data processing code FRENDY; A Verification with HTTR criticality benchmark experiments

Fujimoto, Nozomu*; Tada, Kenichi; Ho, H. Q.; Hamamoto, Shimpei; Nagasumi, Satoru; Ishitsuka, Etsuo

Annals of Nuclear Energy, 158, p.108270_1 - 108270_8, 2021/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Lead void reactivity worth in two critical assembly cores with differing uranium enrichments

Fukushima, Masahiro; Goda, J.*; Bounds, J.*; Cutler, T.*; Grove, T.*; Hutchinson, J.*; James, M.*; McKenzie, G.*; Sanchez, R.*; Oizumi, Akito; et al.

Nuclear Science and Engineering, 189, p.93 - 99, 2018/01

 Times Cited Count:6 Percentile:57.41(Nuclear Science & Technology)

To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worths was conducted in a high-enriched uranium (HEU)/Pb system and a low enriched uranium (LEU)/Pb system using the Comet Critical Assembly at NCERC. The critical experiments were designed to provide complementary data sets having different sensitivities to scattering cross sections of lead. The larger amount of the $$^{238}$$U present in the LEU/Pb core increases the neutron importance above 1 MeV compared with the HEU/Pb core. Since removal of lead from the core shifts the neutron spectrum to the higher energy region, positive lead void reactivity worths were observed in the LEU/Pb core while negative values were observed in the HEU/Pb core. Experimental analyses for the lead void reactivity worths were performed with the Monte Carlo calculation code MCNP6.1 together with nuclear data libraries, JENDL 4.0 and ENDF/B VII.1. The calculation values were found to overestimate the experimental ones for the HEU/Pb core while being consistent for the LEU/Pb core.

Journal Articles

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 54(7), p.795 - 805, 2017/07

 Times Cited Count:5 Percentile:53.32(Nuclear Science & Technology)

A series of integral experiments was conducted in FCA assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, $$^{237}$$Np, $$^{238}$$Pu, $$^{239}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am, and $$^{244}$$Cm. Latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, were tested using benchmark models regarding the fission rate ratios relative to $$^{239}$$Pu. For all the libraries, the benchmark tests by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of $$^{244}$$Cm to $$^{239}$$Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of $$^{238}$$Pu to $$^{239}$$Pu measured in the intermediate neutron spectrum. The cause of discrepancy is furthermore clarified by sensitivity analyses.

Journal Articles

Analyses of iron and concrete shielding experiments at JAEA/TIARA with JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0

Konno, Chikara; Ochiai, Kentaro; Sato, Satoshi; Ota, Masayuki

Fusion Engineering and Design, 98-99, p.2178 - 2181, 2015/10

 Times Cited Count:7 Percentile:57.29(Nuclear Science & Technology)

We have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which were supplied from JAEA, BNL and IAEA, respectively, were used for this analysis. The followings are found out from the results. (1) The calculations with JENDL/HE-2007 agree with all the measured ones well; (2) Those with ENDF/B-VII.1 tend to overestimate the measured ones with the thickness of the assemblies largely; (3) Those with FENDL-3.0 agree with the measured ones well for the iron experiment, while they overestimate the measured ones well for the concrete experiment largely. Some data in ENDF/B-VII.1 and FENDL-3.0 should be revised.

Journal Articles

Benchmark experiments of thermal neutron and capture $$gamma$$-ray distributions in concrete using $$^{252}$$Cf

Asano, Yoshihiro; Sugita, Takeshi*; Hirose, Hideyuki; Suzaki,Takenori

Nuclear Science and Engineering, 151(2), p.251 - 259, 2005/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Benchmark analysis of KRITZ-2 critical experiments

Okumura, Keisuke; Kawasaki, Kenji*; Mori, Takamasa

JAERI-Research 2005-018, 64 Pages, 2005/08


In the KRITZ-2 critical experiments, criticality and pin power distributions were measured at room temperature and high temperature (about 245 degree C) for three different cores loading slightly enriched UO$$_{2}$$ or MOX fuels. For nuclear data testing, benchmark analysis was carried out with a continuous-energy Monte Carlo code MVP and its four nuclear data libraries based on JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI.8. As a result, fairly good agreements with the experimental data were obtained with any libraries for the pin power distributions. However, the JENDL-3.3 and ENDF/B-VI.8 give under-prediction of criticality and too negative isothermal temperature coefficients for slightly enriched UO$$_{2}$$ cores, while the older nuclear data JENDL-3.2 and JEF-2.2 give rather good agreements with the experimental data. From the detailed study with an infinite unit cell model, it was found that the differences among the libraries are mainly due to the different fission cross section of U-235 in the energy rage below 1.0 eV.

JAEA Reports

Proposal of new $$^{235}$$U nuclear data to improve k$$_{eff}$$ biases on $$^{235}$$U enrichment and temperature for low enriched uranium fueled lattices moderated by light water

Wu, H.; Okumura, Keisuke; Shibata, Keiichi

JAERI-Research 2005-013, 31 Pages, 2005/06


The under prediction of k$$_{eff}$$ depending on $$^{235}$$U enrichment in low enriched uranium fueled systems was studied in this report. Benchmark testing was carried out with several evaluated nuclear data files, including the new uranium evaluations from preliminary ENDF/B-VII and CENDL-3.1. Another problem reviewed here was k$$_{eff}$$ underestimation vs. temperature increase, which was observed in the slightly enriched system with recent JENDL and ENDF/B uranium evaluations. Through the substitute analysis of nuclear data of $$^{235}$$U and $$^{238}$$U, we propose a new evaluation of $$^{235}$$U data to solve both of the problems. The new evaluation was tested for various uranium fueled systems including low or highly enriched metal and solution benchmarks in the ICSBEP handbook. As a result, it was found that the combination of the new evaluation of $$^{235}$$U and the $$^{238}$$U data from the preliminary ENDF/B-VII gives quite good results for most of benchmark problems.

Journal Articles

Integral validation of minor actinide nuclear data by using samples irradiated at dounreay prototype fast reactor

Tsujimoto, Kazufumi; Oigawa, Hiroyuki; Shinohara, Nobuo

Proceedings of International Conference on Nuclear Data for Science and Technology (ND 2004), 4 Pages, 2004/09

Japan Atomic Energy Research Institute (JAERI) has been developing technologies for partitioning and transmutation of long-lived nuclides in high-level radioactive waste. In the dedicated transmutation systems, reliable neuclear data of minor actinide (MA) are indispensable to obtain a reliable design of a transmutation system. Present status of MA nuclear data is not so satisfactory. To obtain reliable nuclear data of MA, radiochemically analyzed data of the actinide samples irradiated at the Dounreay Prototype Fast Reactor (PFR) were used in this study. The samples were actinide oxides of 21 different isotopes from thorium to curium. The burnup calculations were performed and the calculated results were compared with the experimental data to validate the neutron cross section data of MA in an evaluated nuclear data file JENDL-3.3, ENDF/B-VI, and JEFF-3.0. The results for uraniumu and plutoniumu samples show good agreements with experimental data. On the other hand, in the results for americium and curiumu, relatively large disagreement with experimental data are showed.

JAEA Reports

Production of MVP neutron cross section libraries based on the latest evaluated nuclear data files

Mori, Takamasa; Nagaya, Yasunobu; Okumura, Keisuke; Kaneko, Kunio*

JAERI-Data/Code 2004-011, 119 Pages, 2004/07


The 2nd version of code system, LICEM-2, has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system can process nuclear data in the latest ENDF-6 format and produce cross section libraries for MVP's capability of transport calculation at arbitrary temperature. By using the present system, MVP neutron cross section libraries have been prepared from the latest evaluations of JENDL, ENDF/B and JEFF data bases. This report describes the specification of MVP neutron cross section library, the details of each code in the code system, how to use them and MVP neutron cross section libraries produced with the code system.

JAEA Reports

Analysis of the TRIGA MARK-II benchmark IEU-COMP-THERM-003 with Monte Carlo code MVP

Mahmood, M. S.; Nagaya, Yasunobu; Mori, Takamasa

JAERI-Tech 2004-027, 30 Pages, 2004/03


The benchmark experiments of the TRIGA Mark-II reactor in the ICSBEP handbook have been analyzed with the Monte Carlo code MVP using the cross section libraries based on JENDL-3.3, JENDL-3.2 and ENDF/B-VI.8. MCNP calculations have been also performed with the ENDF/B-VI.6 library for comparison between the MVP and MCNP results. For both cores labeled 132 and 133, which have different core configurations, the ratio of the calculated to the experimental results (C/E) for keff obtained by the MVP code is 0.999 for JENDL-3.3, 1.003 for JENDL-3.2, and 0.998 for ENDF/B-VI.8. For the MCNP code, the C/E values are 0.998 for both the Core 132 and 133. All the calculated results agree with the reference values within the experimental uncertainties. The results obtained by MVP with ENDF/B-VI.8 and MCNP with ENDF/B-VI.6 differ only by 0.02% for Core 132, and by 0.01% for Core 133.

Journal Articles

Analysis of VENUS-2 MOX core measurements with a Monte Carlo code MVP

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10

We have analyzed the VENUS-2 MOX core benchmark exercises by using a continuous-energy Monte Carlo code MVP with the nuclear data set JENDL-3.2 and ENDF/B-VI release 5. The VENUS-2 core is cruciform and consists of three fuel regions; the squared central region, the inner and the outer part of the peripheral region are fueled with 3.3% UO$$_2$$, 4.0% UO$$_2$$ and MOX. We have constructed 3-D quarter-symmetric calculation model as precisely as possible. All calculations were performed for 200 million histories including 1 million histories of 50 cycles for the initial guess. The C/E values of keff are 1.00500, 0.99793 for JENDL-3.2 and ENDF/B-VI, respectively. They are in good agreement with the experimental one. However, the JENDL-3.2 result overestimates slightly by about 0.5%. For the pin power distribution, the systematic overestimation can be observed in the MOX fuel region. The calculated results tend to underestimate the measured one slightly in the UO$$_2$$ fuel regions. However, the dependence on the libraries is not seen.

Journal Articles

Measurements of neutron induced fission cross-section for $$^{242m}$$Am from 0.003eV to 10keV using lead slowing-down spectrometer, thermal neutron facility and time-of-flight method

Kai, Tetsuya; Kobayashi, Katsuhei*; Yamamoto, Shuji*; Cho, H.*; Fujita, Yoshiaki*; Kimura, Itsuro*; Okawachi, Yasushi*; Wakabayashi, Toshio*

Annals of Nuclear Energy, 28(8), p.723 - 739, 2001/05

 Times Cited Count:7 Percentile:49.64(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Critical and subcritical masses of curium-245, -246 and -247 calculated with a combination of MCNP4A code and JENDL-3.2 library

Okuno, Hiroshi; Kawasaki, Hiromitsu*

JAERI-Research 2000-040, 44 Pages, 2000/09


no abstracts in English

Oral presentation

Problems on ENDF/B-VIII.0 and JEFF-3.3 copper nuclear data above a few MeV

Kwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Sato, Satoshi*

no journal, , 

Our analyses of JAEA/FNS copper benchmark experiment with ENDF/B-VIII.0 and JEFF-3.3 pointed out that the calculation with ENDF/B-VIII.0 underestimated and that with JEFF-3.3 overestimated the measured reaction rate of the $$^{93}$$Nb(n,2n)$$^{rm 92m}$$Nb sensitive to neutrons above 10 MeV. As a result of our detailed study, we specified that this issue was due to the (n,np) and (n,n') reaction data above a few MeV, etc.

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