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
JAEA-Data/Code 2020-014, 30 Pages, 2020/10
The decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. In the process of the decommissioning, preventing radiation exposure of workers is imperative originating in nuclear criticality of fuel debris. This study provides the handy tool enabling the analysis on nuclear criticality of fuel debris. The developed analysis tool named as HAND enables estimation of the criticality of fuel debris in short time. HAND deduces the range of parameters such as the size and composition, in which the criticality of fuel debris is specified. By selecting the range of the parameters using HAND in advance, total calculation time of the detail analysis will be reduced. Since the input data of HAND is designed to be simple and the output of HAND is to be user friendly, this tool is expected to be also an intuitive tool to study the criticality of fuel debris. This report explains the outline of the HAND and input instructions for HAND.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Tada, Kenichi
Kaku Deta Nyusu (Internet), (127), p.1 - 10, 2020/10
The 32nd annual meeting and the subgroup meeting of the Working Party on International Nuclear Data Evaluation Co-operation (WPEC) under the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) was held via a web meeting system from 11 to 15 in May in 2020. The activities about nuclear data measurement and evaluation of each region or country were reported at the annual meeting, and the SG activities were discussed at the subgroup meetings. The summary of these meetings are reported.
Yamamoto, Akio*; Endo, Tomohiro*; Tada, Kenichi
Transactions of the American Nuclear Society, 122(1), p.714 - 717, 2020/06
A generation capability of multi-group cross sections from point-wise cross sections in ACE files is being developed as a function of the nuclear data processing code FRENDY. This presentation describes features of this function and comparison of the processing results between this function and GROUPR module in NJOY.
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.
Tada, Kenichi; Kunieda, Satoshi
KURNS-EKR-5, p.229 - 232, 2019/12
The R-matrix limited formula is formatted by the current nuclear data format and it is adopted some nuclei in the latest evaluated nuclear data library. Since the processing of the R-matrix limited formula is significantly different to the other resonance formulae, it is difficult to treat this formula without large modification of the nuclear data processing code. In this study, we implemented one of the Rmatrix code AMUR to treat this formula in FRENDY. The processing results of FRENDY are compared to those of NJOY2016 to verify FRENDY. The comparison results indicate that FRENDY appropriately treat the R-matrix limited formula with similar computational time.
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.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Yokoyama, Kenji; Tada, Kenichi
Kaku Deta Nyusu (Internet), (124), p.23 - 34, 2019/10
The 31st annual meeting and the subgroup meeting of the Working Party on International Nuclear Data Evaluation Co-operation (WPEC) under the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) was held at the head quarter of OECD/NEA located at Boulogne-Billancourt near Paris from 24 to 28 in June in 2019. The activities about nuclear data measurement and evaluation of each region or country were reported at the annual meeting, and the SG activities were discussed at the subgroup meetings. The summary of these meetings are reported.
Tada, Kenichi; Sakino, Takao*
Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09
Criticality safety of the fuel debris is one of the most important issues, and the adoption of burnup credit is desired. To adopt the burnup credit, validation of the burnup calculation codes is required. In this study, assay data of the used nuclear fuel (2F2DN23, 2F1ZN2, and 2F1ZN3) are evaluated to validate the SWAT4.0 code. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data. To investigate the applicability of SWAT4.0 to the criticality safety evaluation of fuel debris, we evaluated the effect of isotopic composition difference on . The differences in the number densities of U-235, Pu-239, Pu-241, and Sm-149 have a large impact on . However, the reactivity uncertainty related to the burnup analysis was less than 3%. SWAT4.0 appropriately analyses the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.
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.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05
The decommissioning of Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. The criticality safety of fuel debris is imperative to prevent exposure of workers. The investigating criticality monitoring system cannot detect the criticality of fuel debris quickly. The estimation of criticality of fuel debris is required for the fuel debris retrieval. Though the expert knowledge of reactor physics is necessary to estimate the criticality of fuel debris, many people who make a plan of fuel debris retrieval may not know well about criticality analysis. We developed a handy criticality analysis tool HAND to quickly estimate the criticality of fuel debris without expert knowledge of reactor physics. Since the input data of HAND is so simple and users can intuitively understand the calculation results, this tool is expected to be the effective tool to estimate the criticality of fuel debris.
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.
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.
Yamashita, Susumu; Tada, Kenichi; Yoshida, Hiroyuki; Suyama, Kenya
Nihon Genshiryoku Gakkai Wabun Rombunshi, 17(3/4), p.99 - 105, 2018/12
In order to reveal melt relocation behaviors of core internals phenomenologically and to reduce the uncertainties of the melt relocation analysis in existing SA analysis codes, in JAEA, the numerical simulation code for melt relocation and accumulation behaviors based on computational fluid dynamics named JUPITER has been developed. In this paper, to consider the estimation method for fuel debris composition and its re-criticality, we performed the melt accumulating and spreading simulation to the pedestal region by JUPITER and also performed re-criticality analysis by Monte Carlo Codes for Neutron Transport Calculations based on Continuous Energy and Multi-group Methods (MVP) using detailed fuel debris composition data obtained by JUPITER. From the coupled analysis on fuel debris distribution by JUPITER and MVP, we had prospects for a detailed possibility of re-criticality of fuel debris with detailed fuel debris distribution.
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.
Kikuchi, Takeo; Tada, Kenichi; Sakino, Takao; Suyama, Kenya
JAEA-Research 2017-021, 56 Pages, 2018/03
The criticality management of the fuel debris is one of the most important research issues in Japan. The current criticality management adopts the fresh fuel assumption. The adoption of the fresh fuel assumption for the criticality control of the fuel debris is difficult because the k of the fuel debris could exceed 1.0 in most of cases which the fuel debris contains water and does not contain neutron absorbers such as gadolinium. Therefore, the adoption of the burnup credit is considered. The prediction accuracy of the isotopic composition of used nuclear fuel must be required to adopt the burnup credit for the treatment of the fuel debris. JAEA developed a burnup calculation code SWAT4.0 to obtain reference calculation results of the isotopic composition of the used nuclear fuel. This code is used to evaluate the composition of fuel debris. In order to investigate the prediction accuracy of SWAT4.0, we analyzed the PIE of BWR obtained from 2F2DN23.