JAEA-Conf 2018-001, p.87 - 91, 2018/12
Status and plan of JENDL will be presented. After the release of JENDL-4.0 in 2010, six special purpose files have been developed. Four of them were already released and two are under preparation for the release. New decay and yield data for fission products were released as JENDL/FPD-2011 and JENDL/FPY-2011 in 2011, respectively. JENDL-4.0/HE released in 2015 includes proton and neutron induced reaction data up to 200 MeV. Comprehensive decay data were released as JENDL/DDF-2015 which contains data for 3,237 nuclides. New photonuclear reaction data JENDL/PD-2016 and an activation file JENDL/AD-2017 are under preparation for release. Regarding general purpose file, two activities are in progress. One is JENL-4.0u which is created for maintenance of JENDL-4.0 and the other is development of next version of JENDL. For the next JENDL, evaluation for light nuclei and structure material are in progress. It is planed that next version of JENDL will be JENDL-5 which contains nuclear data for all nuclei having natural abundance. Addition of covariance data will be one of the main targets.
Hashimoto, Shintaro; Sato, Tatsuhiko; Iwamoto, Yosuke; Ogawa, Tatsuhiko; Furuta, Takuya; Abe, Shinichiro; Niita, Koji*
Kaku Deta Nyusu (Internet), (120), p.26 - 34, 2018/06
Particle and heavy-ion transport code system PHITS has been used for calculations of radiation shielding in accelerator facilities. PHITS describes physical phenomena induced by radiation as combination of transport and collision processes. The collision process including nuclear reactions is simulated by the three-step calculation: a generation of a reaction, pre-equilibrium, and compound processes. In the simulation, many physics models are used. This report explains roles of the models in PHITS and shows their developments we recently performed.
Suyama, Kenya; Yokoyama, Kenji
Kaku Deta Nyusu (Internet), (119), p.38 - 47, 2018/02
We have developed numerous neutronics calculation codes in Japan. However, development of the one-point burnup calculation code which replaces the still widely used ORIGEN2 code has not been successful. The one point burnup code is indispensable to evaluate the characteristics of the used nuclear fuel increasing in Japan, and it uses all evaluated nuclear data including the fission yield and decay data as well as cross section data. It means that it could be the Killer Application in the field of the nuclear data and neutronics code. This report describes the necessity of the one point burnup calculation code development in Japan and required function and performance which have been considered by authors.
Iwamoto, Osamu; Shibata, Keiichi; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Ichihara, Akira; Nakayama, Shinsuke
EPJ Web of Conferences (Internet), 146, p.02005_1 - 02005_6, 2017/09
Kaku Deta Nyusu (Internet), (117), p.5 - 14, 2017/06
The benchmark calculation is one of the main activities of the Nuclear Science Committee under the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/NSC). The international benchmark relatively frequently means the benchmark activity carried out by the NEA. In this manuscript, the author discusses the significance of the international benchmark by describing (i) the current status of the benchmark in the field of the reactor physics conducted by the OECD/NEA/NSC, (ii) revision of the neutronics calculation code system to reflect the results of the benchmark, (iii) the benchmark calculation as the asset for the future research and development, (iv) examples of the benchmark calculation based on the experimental data, and (v) how to propose the benchmark in the OECD/NEA/NSC.
Katakura, Junichi; Yoshida, Tadashi*; Oyamatsu, Kazuhiro*; Tachibana, Takahiro*
JAERI 1343, 79 Pages, 2001/07
no abstracts in English
*; Takemura, Morio*
JNC-TJ9440 2000-005, 157 Pages, 2000/03
With use of the two-dimensional discrete ordinates code DORT and the standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the representative configurations in the Radial Shield Attenuation Experiment of the JASPER were performed. The results were compared with those obtained with use of traditional method DOT3.5/JSDJ2 for the previous JASPER experimetal analyses. In general, the change of the cross section library gives higher results and the change of the transport code gives lower results. Finally the new analysis method gives better agreement with the experimental results and also less deviations of calculational errors between various detectors. Experimental analyses for the thick concrete configulation in the Gap Streaming Experiment of the JASPER was also performed with the new analysis method, after solving the poor agreement found in last year with the original JASPER experimental analyses. The same tendency due to the library change was confirmed with the above mentioned analyses of the Radial Shield Attenuation Experiment. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments were continued through the above-mentioned experimental analyses and related informations were added for repletion of the database preserved in a computer disk holding previously accumulated data. Input data descriptions were made for auxiliary routines needed for the experimental analyses and their sample data were compiled and stored in the database.
JNC-TJ9400 2000-007, 46 Pages, 2000/03
For fission cross section and prompt fission neutron spectrum, which largely influence core characteristics of a fast reactor, we have performed experimental and analytical studies for developing an advanced technique to measure absolute fission cross section and neutron fission spectrum for actinide nuclides such as Np237. As the results, we could develop an advanced technique, which combines a normalization technique for the well-known differential cross section and a correction method by a Monte-Carlo code for sample effects. This advanced technique accurately provides both absolute fission cross section and prompt fission neutron spectrum individually. By employing this technique, in this study, we have measured for three actinides (Np, Th and U), then, have obtained the fission cross sections and fission spectrum parameter data for those nuclides. Furthermore, we have also performed an analytical study to examine sensitivity of fission spectrum parameter to core multiplication factor by using the standard calculation code for a first reactor.
JNC-TJ9400 2000-005, 182 Pages, 2000/03
The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radial heterogereous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By coparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and O, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. Bycoaparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it ...
JNC-TJ9400 2000-009, 63 Pages, 2000/02
The present status of nuclear data for technetium (Tc)-99, which is a well-known fission product (FP), has been reviewed and investigated. And making use of the Kyoto university Lead Slowing-down Spectrometer (KULS), the cross section of the Tc (n, ) Tc reaction has been measured in the energy range from thermal to keV neutron energy with an Ar-gas proportinal counter. The neutron flux/spectrum has been monitored with a BF proportional counter, and the relative measurement has been normalized to the well-known standard capture cross section value for the Tc (n, ) Tc reaction at 0.0253 eV. Self-shielding corrections, especially near the resonance peaks, were made by the calculations with the MCNP code. Although the experimental data measured by Chou et al with a lead slowing-down spectrometer are higher in general, the energy dependency is similar to the present measurement. The evaluated data in ENDF/B-VI and JENDL-3.2 are higher near the resonances at 5.6 and 20 eV and above several 100 eV. A lead slowing-down spectrometer was installed coupled to a 46 MeV electron linac at the Research Reactor Institute, Kyoto university (KURRI). Characteristics of the Kyoto University Lead Slowing-down Spectrometer (KULS) were measured and (1)the relation between neutron slowing-down time t(s) and energy E(keV) (E=190/t in Bi hole and E=156/t in Pb hole) and (2)the energy resolution (40% in Bi and Pb holes) were experimentally investigated. (3)The neutron energy spectrum in the KULS was also measured by the neutron TOF method. The results obtained by the MCNP code were in general agreement with these experimental ones.
JNC-TJ9400 2000-008, 61 Pages, 2000/02
For studies on nuclear transmutation of long-lived fission products (LLFPs) in a fast reactor, detailed characteristics of reactor core such as transmutation performance have to be investigated, so accurate neutron cross section data of LLFPs become necessary. Therefore, the keV-neutron capture cross sections of Tc-99, which is one of important LLFPs, were measured in the present study to obtain the accurate data. The measurement was relative to the standard capture cross sections of Au-197. A neutron time-of-flight method was adopted with a ns-pulsed neutron source by a Pelletron accelerator and a large anti-Compton NaI(TI) gamma-ray detector. As a result, the capture cross sections of Tc-99 were obtained with the error of about 5 % in the incident neutlon energy region of 10 to 600 keV. The present data were compared with other experimental data and the evaluated values of JENDL-3.2, and it was found that the evaluations of JENDL-3.2 were 15-20 % smaller than the present measurements.
JNC-TJ9400 2000-004, 109 Pages, 2000/02
We estimated covariances of the JENDL-3.2 data on the nuclides and reactions needed to analyze fast-reactor cores for the past three years, and produced covariance files. The present work was undertaken to re-examine the covariance files and to make some improvements. The covariances improved are the ones for the inelastic scattering cross section of O, the total cross section of Na, the fission cross section of U, the capture cross section of U, and the resolved resonance parameters for U. Moreover, the covariances of U data were newly estimated by the present work. The covariances obtained were compiled in the ENDF-6 format.
JNC-TN9400 99-049, 74 Pages, 1999/04
This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2d) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of keff was 1.1%k/k higher than the measured value, Na void worth C/E values were 1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes, though the efect should be investigated in any future experiments.) several sample worth values were small compared with calculational uncertaint
PNC-TJ9500 98-002, 126 Pages, 1998/03
PNC-TJ9500 98-001, 102 Pages, 1998/03
PNC-TJ9604 97-001, 108 Pages, 1997/03
A lead slowing-down spectrometer was installed coupled to the 46 MeV electron linac at Research Reactor Institute, Kyoto university (KURRI). Characteristics of the Kyoto University Lead Slowing-down Spectrometer (KULS) were measured for (1)the relation between neutron slowing-down time t(s) and energy E (keV) (E=190/t in Bi hole and E=156/ in Pb hole), (2)energy resolutlon (40 % in Bi and Pb holes), and (3)neutron energy spectrum by the neutron TOF method. The results obtained by the MCNP code were in general agreement with these experimental ones. The KULS has been applied to the fission cross section measurements of Am-241, Am-243 and Am-242m relative to that of U-235 from 0.1 eV to 10 keV, making use of the back-to-back type double fission chambers. For Am-241, Dabbs and ENDF/B-VI data are in good agreement with the present measurement. The JEND L-3.2 data are smaller by a factor of 2 between 10 and 200 eV. The ENDF/B-VI data for Am-243 are lower between 15 and 60 eV, and the JENDL-3.2 are lower in general above 100 eV. It has been found that the preliminary result for the Am-242m(n,f) reaction is close to the ENDF/B-VI and the JENDL-3.2 data. Thermal neutron cross sections for Am-241 and Am-243 have also been measured in a standard Maxwellian distribution spectrum field. Finally, aiming at the measurement of capture cross section for MA nuclides, the experimental investigation for Np-237 sample (2 mg) has been performed with the KULS. Due to the comparable background counts to the foreground ones, the capture events from the sample have scarecely been detected with an Ar-gas proportional counter.
PNC-TJ9055 97-001, 112 Pages, 1997/03
With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the Axial Shield Experiment (homogeneous and central blockage type shield configurations with BC or stainless steel shield material) were performed. The results were compared with those obtained by the same analysis method and input data using JSDJ2 library that had been applied consistently to the JASPER experiment analyses. In general, the results with JSSTDL analyses are higher than those by JSDJ2 as were found in analyses in last year for the Radial Shield Attenuation Experiment and the Special Materials Experiment. Consideration was made on the discrepancies between JSSTDL and JSDJ2 analysis results of the Axial Shield Experiment and also those of the sodium configulation in the Radial Shield Attenuation Experiment. The former was done by exchange of macro cross section of each region, and the latter forcused on sodium cross section was done with use of cross section sensitivity analysis method. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected in previous study in last year, were continued and new data were added into the computer disk holding previous ones.
Shibata, Keiichi; Nakagawa, Tsuneo; Asami, Tetsuo*; Fukahori, Tokio; ; Chiba, Satoshi; Mizumoto, Motoharu; ; ; Nakajima, Yutaka; et al.
JAERI 1319, 516 Pages, 1990/06
no abstracts in English
Igarashi, Shinichi; Asami, Tetsuo; Kikuchi, Yasuyuki; Nakagawa, Tsuneo; Narita, Tsutomu; Shibata, Keiichi
Nippon Genshiryoku Gakkai-Shi, 26(3), p.191 - 198, 1984/00
no abstracts in English
PNC-TJ250 82-12, 56 Pages, 1982/03