Maxwellian-averaged neutron-induced reaction cross sections and astrophysical reaction rates for = 1 keV to 1 MeV calculated from microscopic neutron cross section library JENDL-3.3

Nakagawa, Tsuneo; Chiba, Satoshi; Hayakawa, Takehito; Kajino, Toshitaka*

Atomic Data and Nuclear Data Tables, 91(2), p.77 - 186, 2005/11

https://doi.org/10.1016/j.adt.2005.08.002

no abstracts in English

Benchmark experiments of thermal neutron and capture -ray distributions in concrete using Cf

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

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

https://doi.org/10.13182/NSE05-A2544

no abstracts in English

Study on the prediction accuracy of nuclide generation and depletion with JENDL

Okumura, Keisuke; Oki, Shigeo*; Yamamoto, Munenari*; Matsumoto, Hideki*; Ando, Yoshihira*; Tsujimoto, Kazufumi; Sasahara, Akihiro*; Katakura, Junichi; Matsumura, Tetsuo*; Aoyama, Takafumi*; et al.

JAERI-Research 2004-025, 154 Pages, 2005/01

JAERI-Research-2004-025.pdf:19.46MB

This report summarizes the activity (FY2000-2003) of Working Group (WG) on Evaluation of Nuclide Generation and Depletion under Subcommittee on Nuclear Fuel Cycle of Japanese Nuclear Data Committee. In the WG, analyses of Post Irradiation Examinations have been carried out for UO and MOX fuels irradiated in PWRs, BWRs and FBRs, and for actinide samples irradiated in fast reactors, by using ORIGEN or more detailed calculation codes with their libraries based on JENDL-3.2, JENDL-3.3 and other foreign nuclear data files. From these results, current prediction accuracy and problems for evaluation of nuclide generation and depletion are discussed. Furthermore, this report covers other products of our activity; development of the ORIGEN libraries for PWR, BWR and FBR based on JENDL-3.3, study on introduction of neutron spectrum index to ORIGEN calculations, and results of questionnaire survey on desirable accuracy of ORIGEN calculations.

A Set of ORIGEN2 cross section libraries based on JENDL-3.3 library; ORLIBJ33

Katakura, Junichi; Kataoka, Masaharu*; Suyama, Kenya; Jin, Tomoyuki*; Oki, Shigeo*

JAERI-Data/Code 2004-015, 115 Pages, 2004/11

JAERI-Data-Code-2004-015.pdf:16.52MB

A set of cross section libraries for ORIGEN2 code, ORLIBJ33, has been produced based on the latest Japanese Evaluated Nuclear Data Library JENDL-3.3. The produced libraries are for LWR's which include PWR, BWR and their MOX fuels. The libraries for FBR's are also produced. Using the libraries for LWR, comparisons with old libraries based on JENDL-3.2 were performed. The comparisons with measured PIE data were also carried out. For the libraries for FBR, the comparisons with the calculations using the old libraries were performed and the effects using different libraries were discussed.

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.

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

JAERI-Tech-2004-027.pdf:2.26MB

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.

Evaluation of neutronic characteristics of TRACY water-reflected core system

Sono, Hiroki; Yanagisawa, Hiroshi*; Miyoshi, Yoshinori

JAERI-Tech 2003-096, 84 Pages, 2004/01

JAERI-Tech-2003-096.pdf:3.6MB

Prior to the supercritical experiments using a water-reflected core of the TRACY Facility, neutronic characteristics regarding criticality and reactivity of the core system were evaluated. In the analyses, a continuous energy Monte Carlo code, MVP, and a two-dimensional transport code, TWOTRAN, were used together with a nuclear data library, JENDL-3.3. By comparison to the characteristics in the former-used bare core system of TRACY, the water reflector was estimated not to change the kinetic parameter and to reduce the critical solution level by 20 %, the temperature coefficient of reactivity by 610 %, and the void coefficient of reactivity by 18 %, respectively. According to the Nordheim-Fuchs model, the first peak power during a power excursion was evaluated to be 15 % smaller than that in the bare system under the same conditions of fuel and reactivity insertion. The influence of the void feedback effect of reactivity, which is left out of consideration in the model, on the power characteristics will be evaluated from the results of the experiments.

Evaluation of the Bi/Bi branching ratio of the Bi()Bi cross section in the neutron energy range from 200keV to 3.0MeV

Ichihara, Akira; Shibata, Keiichi

Journal of Nuclear Science and Technology, 40(11), p.980 - 982, 2003/11

https://doi.org/10.3327/jnst.40.980

no abstracts in English

Validating JENDL-3.3 for water-reflected low-enriched uranium solution systems using STACY ICSBEP benchmark models

Yamamoto, Toshihiro; Miyoshi, Yoshinori; Kiyosumi, Takehide*

Nuclear Science and Engineering, 145(1), p.132 - 144, 2003/09

https://doi.org/10.13182/NSE03-A2369

no abstracts in English

The Libraries FSXLIB and MATXSLIB based on JENDL-3.3

Kosako, Kazuaki*; Yamano, Naoki*; Fukahori, Tokio; Shibata, Keiichi; Hasegawa, Akira

JAERI-Data/Code 2003-011, 38 Pages, 2003/07

JAERI-Data-Code-2003-011.pdf:1.29MB

The third revision of JENDL-3 (JENDL-3.3) was released in May 2002. The library is useful for many applications. For users' convenience, we have produced two JENDL-3.3 based libraries FSXLIB-J33 and MATXSLIB-J33 for transport calculation codes such as MCNP and ANISN. These two libraries are available on request.