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Nakayama, Shinsuke; Iwamoto, Osamu; Kimura, Atsushi
EPJ Web of Conferences, 294, p.07001_1 - 07001_6, 2024/04
Graphite is a candidate of moderator in innovative nuclear reactors such as molten salt reactors. Scattering of thermal neutrons by the moderator material has a significant impact on the reactor core design. To contribute to the development of innovative nuclear reactors, an evaluation method of thermal neutron scattering law for reactor grade graphite was studied. The inelastic scattering component due to lattice vibration was evaluated based on the phonon density of states computed with first-principles lattice dynamics simulations. The simulations were performed for ideal crystalline graphite. The coherent elastic scattering component due to crystal structure was evaluated based on neutron transmission and scattering experiments recently performed in the J-PARC/MLF facility. In comparison with the neutron transmission experiments, it was found that the quantification of small-angle neutron scattering due to structures larger than crystal, such as pores in graphite, is important. Based on the above methods, thermal neutron scattering law data for reactor-grade graphite at room temperature were evaluated.
Tada, Kenichi
no journal, ,
In JAEA, the nuclear data processing system FRENDY (FRom Evaluated Nuclear Data librarY to any application) has been developed. In this presentation, the overview and verification of FRENDY is described.
Tada, Kenichi; Ikehara, Tadashi; Ono, Michitaka*; Tojo, Masayuki*
no journal, ,
JAEA develops a nuclear data processing code FRENDY. We can comparison and verification of conventional nuclear data processing code using FRENDY. In this presentation, we focus on the thermal scattering law data. We found some problems of NJOY to process the thermal scattering law data as follows, (1) The generation of input file is complex and we found some inputting error in the official ACE library, (2) The maximum energy of ACE file is not identical to the inputted maximum energy, (3) If user uses iwt=2 option in ACER module, MCNP6.1 cannot treat this generated ACE file appropriately and the calculation will not completed This presentation explains the overview of these problems.
Tada, Kenichi
no journal, ,
We have compared the processing results of FRENDY to those of NJOY for verification of FRENDY. In this presentation, we focus on the processing of thermal scattering law data. This presentation explains the problems of NJOY. This presentation also explains the implementation of input checking functions for collect nuclear data processing and the recent development of FRENDY, e.g., the development of neutron-induced multi-group cross-section file generation function.
Nakayama, Shinsuke; Iwamoto, Osamu
no journal, ,
In molten salt reactors and small modular reactors (SMRs), the use of graphite and CaH
as moderators is being considered, respectively. Thermal neutron scattering law of moderator material has a large influence on the reactor core design. In addition, charged-particle emission reactions such as (n,p) and (n,a) on K-39 in molten salt and on Cu-63 in heat pipes of SMRs can produce nuclides that are problematic for waste management. Therefore, accurate data on thermal neutron scattering laws for graphite and CaH, and charged-particle emission reaction cross sections for K-39 and Cu-63 are important for the core design of these innovative reactors. Based on the above, we have been studying the evaluation method of these data. The progress to date will be presented.
O of JENDL-5 at several moderator temperatures on neutronics calculationTada, Kenichi; Watanabe, Tomoaki; Endo, Tomohiro*; Yamamoto, Akio*
no journal, ,
The PWR pin-cell calculations using the cross section data of each evaluated nuclear data library were compared at several moderator temperatures for the verification of the thermal scattering law data for H in HO of JENDL-5. Compared with the commonly used evaluated nuclear data libraries JENDL-4.0 and ENDF-B/VII.1, the relative differences of k-infinity varied with the moderator temperature. This difference may affect the moderator temperature coefficient. However, it is difficult to judge which library is good since there are not so much experimental data for the cross section and double differential cross section measurement of high-temperature H in HO data. Additional experimental data are required to improve the prediction accuracy of the thermal scattering law data for H in HO data.
Be thermal scattering law data considering crystallite size, 2; Effect on nuclear properties of A-FNS test modulesKwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Sato, Satoshi*
no journal, ,
The analyses of the beryllium benchmark experiment at JAEA/FNS and the test modules of a fusion neutron source A-FNS were performed with the thermal scattering law data of Be considering the crystallite size of beryllium produced by the European Spallation Source group. It was concluded that the overestimation of the calculated reaction rates sensitive to low energy neutrons in the beryllium benchmark experiment at JAEA/FNS decreased with increasing the crystallite size of beryllium. The effect of the Be thermal scattering law data appeared for neutron flux below 0.1 eV in the test modules of A-FNS but the effect on tritium production rate, etc. was small because the neutron flux below 0.1 eV was small.
Nakayama, Shinsuke; Iwamoto, Osamu
no journal, ,
The use of graphite and hydrogen compounds as moderators has been considered for molten salt reactors and small modular reactors. Scattering of thermal neutrons by moderators has a significant impact on reactor core design. In addition, charged-particle emission reactions on nuclides contained in molten salts and structural materials can produce nuclides that pose a problem for waste management. Therefore, accurate thermal neutron scattering law and charged-particle emission reaction cross section data for the above materials are important for the development of such innovative reactors. Based on the above, these nuclear data were evaluated and compiled as an evaluated nuclear data file in the MEXT Innovative Nuclear Research and Development Program entitled "Development of Nuclear Data Evaluation Framework for Innovative Reactor". The evaluation method of these nuclear data is outlined.
Kimura, Atsushi; Endo, Shunsuke; Nakamura, Shoji; Rovira Leveroni, G.
no journal, ,
no abstracts in English
Harada, Yoshinari*; Yamaguchi, Hibiki*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi
no journal, ,
The data assimilation using the prompt neutron decay constant was performed to reduce the uncertainty in nuclear calculations due to the thermal neutron scattering law (TSL) for light water. The uncertainty in the effective multiplication factor due to TSL for light water can be reduced using the bias factor method based on the deterministic sampling method when there is a strong correlation between nuclear properties through TSL for light water.
