Proceedings of International Conference on the Physics of Reactors; Transition To A Scalable Nuclear Future (PHYSOR 2020) (USB Flash Drive), 8 Pages, 2020/03
The probability table is widely used for continuous energy Monte Carlo calculation codes to treat the self-shielding effect in the unresolved resonance region. The ladder method is used to calculate the probability table. This method generates a lot of pseudo resonance structures using random numbers based on the averaged resonance parameters. The probability table affects the ladder number. i.e., number of pseudo resonance structures. The ladder number has large impact on the generation time of the cross section library. In this study, the appropriate ladder number is investigated. The probability table of all nuclides prepared in JENDL-4.0 is used to investigate the appropriate ladder number. The comparison results indicate that the differences of the probability table are enough small when the ladder number is 100.
JAEA-Conf 2019-001, p.29 - 34, 2019/11
JAEA has developed a new nuclear data processing code FRENDY (FRom Evaluated Nuclear Data librarY to any application) to generate a cross-section data library from evaluated nuclear data library JENDL. In this presentation, author explains how to generate cross-section data library and overview and features of FRENDY.
Nakayama, Shinsuke; Iwamoto, Osamu; Watanabe, Yukinobu*
Physical Review C, 100(4), p.044603_1 - 044603_8, 2019/10
The weakly-bound nature of the deuteron brings the complexity of deuteron-induced reactions compared to nucleon-induced ones, and is expected to affect various physical quantities observed in deuteron-induced reactions. Aiming to deep understanding and accurate prediction for the emission of light composite particle (LCP) in deuteron-induced reactions, we revise the computational system dedicated to deuteron-induced reactions, called DEURACS. The model by Iwamoto and Harada describing pre-equilibrium cluster emission which was successfully applied to LCP emission innucleon-induced reactions is integrated into the framework of DEURACS, in which the breakup processes of incident deuteron are explicitly taken into account. The phenomenological model by Kalbach is also adopted to estimate the contribution from the direct pickup process. Using the revised DEURACS, we analyze the , , and reactions in the target mass range . Regardless of the targets, the calculation results successfully reproduced the experimental data for each reaction, simultaneously. These results demonstrates that the LCP emission from the pre-equilibrium and compound nucleus processes in deuteron-induced reactions,which occupies a large part of the total LCP emission,can be described by the same theoretical models as used in nucleon-induced reactions when the breakup processes of incident deuteron are properly considered.
Nuclear Data Newsletter (Internet), (67), P. 2, 2019/07
This is an advertisement of our nuclear data processing system FRENDY for Nuclear Data Newsletter published by IAEA nuclear data section.
Nakayama, Shinsuke; Iwamoto, Osamu; Iwamoto, Nobuyuki; Hashimoto, Shintaro
Kaku Deta Nyusu (Internet), (123), p.53 - 59, 2019/06
The 2019 International Conference on Nuclear Data for Science and Technology (ND2019) was held at the China National Convention Center on May 19-24, 2019. The series of the ND conferences are the largest conferences in nuclear data research field that are held every three years. In this paper, as a part of the conference reports of ND2019, the authors gave summaries of the presentations on nuclear data evaluation and theory conducted at the conference.
Kaku Deta Nyusu (Internet), (122), p.9 - 21, 2019/02
This paper reports the overview of the technical meeting of nuclear data processing in IAEA to Japanese researchers. In this technical meeting, the current status of nuclear data processing codes and verification of them are described.
Robutsuri No Kenkyu (Internet), (71), 13 Pages, 2019/02
The nuclear data processing is very important to connect between the evaluated nuclear data library and the particle transport calculation code. However, many nuclear engineers do not know well about the nuclear data processing. This paper describes the overview of nuclear data processing and our nuclear data processing code FRENDY. This paper also lists references about the nuclear data processing.
Tada, Kenichi; Kunieda, Satoshi; Nagaya, Yasunobu
JAEA-Data/Code 2018-014, 106 Pages, 2019/01
A new nuclear data processing code FRENDY has been developed in order to process the evaluated nuclear data library JENDL. Development of FRENDY helps to disseminate JENDL and various nuclear calculation codes. FRENDY is developed not only to process the evaluated nuclear data file but also to implement the FRENDY functions to other calculation codes. Users can easily use many functions e.g., read, write, and process the evaluated nuclear data file, in their own codes when they implement the classes of FRENDY to their codes. FRENDY is coded with considering maintainability, modularity, portability and flexibility. The processing method of FRENDY is similar to that of NJOY. The current version of FRENDY treats the ENDF-6 format and generates the ACE file which is used for the continuous energy Monte Carlo codes such as PHITS and MCNP. This report describes the nuclear data processing methods and input instructions for FRENDY.
Yokoyama, Kenji; Kitada, Takanori*
Journal of Nuclear Science and Technology, 56(1), p.87 - 104, 2019/01
no abstracts in English
Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi
Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.1493 - 1502, 2019/00
A perturbation capability of ACE formatted cross section files was developed using the modules of FRENDY. Uncertainty quantification using MCNP was carried out for the Godiva critical experiment by the RS method. We verified the results of the RS method by comparing with those obtained by the conventional sensitivity analyses. Moreover, uncertainty reduction using the bias factor method with the RS technique was applied to kinetic parameter, i.e., neutron generation time.
Nishio, Katsuhisa; Utsuno, Yutaka; Chiba, Satoshi*; Koura, Hiroyuki; Iwamoto, Osamu; Nakamura, Shoji
JAEA-Conf 2018-001, 226 Pages, 2018/12
The 2017 Symposium on Nuclear Data was held at iVil in Tokai on November 16-17, 2017. The symposium was hosted by the Nuclear Data Division of the Atomic Energy Society of Japan (AESJ) and Advanced Science Research Center of Japan Atomic Energy Agency, and co-hosted by Japanese Nuclear Data Committee of AESJ and North Kanto Branch of AESJ. In the symposium, a tutorial was given by Prof. Rykaczewski (ORNL) "New nuclear data from total absorption spectroscopy and beta-delayed neutron measurements", as well as six oral sessions, "Nuclear Physics and Nuclear Data" (two sessions), "Nuclear Theory and Nuclear Data", "Reactors" and "Nuclear Data and Their Applications" (two sessions). In addition, recent advances in experiment, theory, evaluation, benchmark, and application were presented in the poster session. The symposium had 79 participants, who contributed to very active and fruitful discussions. This report consists of 37 papers, including those of 14 oral and 23 poster presentations.
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.
Stankovskiy, A.*; Iwamoto, Hiroki; elik, Y.*; Van den Eynde, G.*
Annals of Nuclear Energy, 120, p.207 - 218, 2018/10
Propagation of high-energy (above 20-MeV) nuclear data uncertainties on the safety related neutronic responses in accelerator driven systems has been assessed. The total core power and production of radionuclides contributing to radiation source terms were focused on. The article features a method based on the Monte Carlo sampling of random nuclear data files from the covariance matrices generated from the sets of reaction cross sections obtained with model calculations of high-energy particle interactions with matter or picked up from already existing nuclear data libraries. It has been demonstrated that nuclear data uncertainties do not need to be propagated through particle transport calculations to obtain uncertainties on the responses. This advantage allowed to investigate the convergence of the sample average to the best estimate. The number of random nuclear data file sets needed to obtain reliable uncertainty on the total core power is around 300 that results in the uncertainty of 14%. The uncertainties on the concentrations of nuclides most important for the safety assessment that are accumulated in lead-bismuth eutectic during irradiation, range from 5 to 60%. Concentrations of some nuclides exemplified by Tritium converge much slower than neutron multiplicities so that several thousands of samples are needed to ensure reliable uncertainty estimates.
Nakayama, Shinsuke; Furutachi, Naoya; Iwamoto, Osamu; Watanabe, Yukinobu*
Physical Review C, 98(4), p.044606_1 - 044606_8, 2018/10
Use of deuteron-induced spallation reactions at intermediate energies has recently been proposed for transmutation of several long-lived fission products (LLFPs). In the design study of a transmutation system using a deuteron primary beam, accurate cross section data of deuteron-induced reactions on the LLFPs are indispensable. In the present study, production cross sections of residual nuclei in the deuteron-induced reactions on Zr and Pd at MeV/nucleon are analyzed using DEURACS, in which the breakup processes are explicitly taken into account. The calculated values reproduced the experimental data quantitatively well. From a component-by-component analysis, it was found that the components of nucleon absorption make the significant contributions to residual nuclei production. This result strongly indicates that consideration of the breakup processes is essentially important to predict production of residual nuclei in deuteron-induced reactions.
Journal of Nuclear Science and Technology, 55(6), p.614 - 622, 2018/06
Toward the development of the next version of Japanese Evaluated Nuclear Data Library (JENDL) general-purpose file, we calculate neutron cross-sections on Cu from 50 keV to 20MeV, which is the incident energy range above the resolved resonance region in JENDL-4.0. A dispersive optical model potential is adopted with a coupled-channel method for interaction between neutron and Cu. Direct, pre-equilibrium, and compound processes are taken into account in the calculation. All cross-sections, differential and double-differential cross-sections are consistently calculated with a single set of model parameters. The calculation results reproduce the measured data very well. In addition, disagreement between the calculated and experimental values seen in an integral test for the Cu reaction is improved by using the cross-section data obtained from the present work.
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.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Yokoyama, Kenji; Tada, Kenichi
Kaku Deta Nyusu (Internet), (120), p.35 - 46, 2018/06
We report 30th WPEC meeting, expert group meeting, and subgroup meeting in Paris, May 14-18, 2018.
Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.2929 - 2939, 2018/04
JAEA develops a new nuclear data processing system FRENDY. We investigated all processing methods and we focused on the probability table generation using the ladder method which is adopted in the PURR module in NJOY. To improve the probability table generation, the more sophisticated method was introduced in the calculation methods of the Chi-Squared random numbers and the complex error function. We also investigated the appropriate ladder number. To investigate the impact of the difference of the complex error function calculation method, the K values of the benchmark experiments with the probability tables by the both methods were compared. The calculation results indicated that the appropriate ladder number is 100 and the difference of the calculation methods of the Chi-Squared random numbers and the complex error function has no significant impact on the neutronics calculation.
Tada, Kenichi; Kosako, Kazuaki*; Yokoyama, Kenji; Konno, Chikara
Nippon Genshiryoku Gakkai-Shi, 60(3), p.168 - 172, 2018/03
The neutronics calculation codes cannot treat the evaluated nuclear data file directly. The nuclear data processing is required to use the nuclear data file in the neutronics calculation codes. The nuclear data processing is not just a converter but also many processes to evaluate the physical values for the neutronics calculation codes. In this paper, we describe the overview of the nuclear data processing and validation of the nuclear data.
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.