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Journal Articles

Thoughts at the departure from OECD/NEA

Suyama, Kenya

Kaku Deta Nyusu (Internet), (129), p.44 - 53, 2021/06

I have returned to Japan after finishing 3 year contract from with OECD/NEA. This manuscript summarizes my thoughts in France, that I had been considering during the continuous teleworking because of COVID-19 pandemic broke out from the last year.

Journal Articles

Overview of OECD/NEA and contribution of Japan, 5; Nuclear Science Committee and the Data Bank

Tsuda, Shuichi; Suyama, Kenya*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(3), p.278 - 282, 2021/03

Series of articles have been posted to Journal of the Atomic Energy Society of Japan (ATOMO$$Sigma$$) by Japanese secretaries working in Nuclear Energy Agency (NEA) in Organisation for Economic Co-operation and Development (OECD) to share the recent information. The Division of Nuclear Science has been playing an important role to provide a place for exchanging information about nuclear physics and nuclear data, recently about fuels and materials as well, and the author had been working in the Division as a secretariat for three years since September 2017. The Data bank is in charge of management and distribution of databases developed in NEA, and has been strengthening the user support system and cooperation with other NEA committees. This paper will report the recent activities related to Nuclear Science and the Data bank, and hopefully also provide information for further deepening relations between NEA and the Japanese participants by introducing their contributions.

JAEA Reports

Development of inventory calculation modules using ORIGEN-S for decommissioning

Matsuda, Norihiro; Konno, Chikara; Ikehara, Tadashi; Okumura, Keisuke; Suyama, Kenya*

JAEA-Data/Code 2020-003, 33 Pages, 2020/03


Data handling modules for the radioactivity calculation code, ORIGEN-S, are developed for the reliable evaluations of radioactivity inventory. By using these modules, an activation cross-section data library for the ORIGEN-S code is updated easily and effectively based on a facility-specific neutron spectrum and multi-group neutron activation cross-section library for decommissioning of nuclear facilities, MAXS2015. In order to guarantee the reliability of the radioactivity calculations, functions of data verification in a visual way and numerical comparison between before and after the data processing are also prepared.

JAEA Reports

Applicability of statistical geometry model to light water moderating systems

Mori, Takamasa; Kojima, Kensuke*; Suyama, Kenya

JAEA-Research 2018-010, 57 Pages, 2019/02


In order to estimate applicability of the statistical geometry model (STGM) of MVP/GMVP, a parametric study in infinite geometry and criticality safety analyses for direct disposal of spent fuel in simple finite geometry have been carried out by using the MVP Monte Carlo code. It has been found that calculations with STGM for larger fuel spheres give larger thermal utilization factors and larger infinite multiplication factors compared with explicit random models in the range of fuel sphere packing fraction between 6.5 % and 63.3 %. Substantial differences are not observed between the results with two nearest neighbor distributions (NNDs); that given by the MCRDF code and the analytical expression based on a statistically uniform distribution. It is inferred that the overestimation by STGM is caused by the facts that STGM cannot take account of the surroundings of each neutron, whether a fuel sphere rich region or a water moderator rich one, because STGM always uses an NND averaged over such surroundings and that STGM, therefore, cannot take the effect of consecutive scatterings in the water moderator into account.

Journal Articles

Coupled analysis of fuel debris distribution and recriticality by both multiphase/multicomponent flow and continuous energy neutron transport Monte Carlo simulations

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.

JAEA Reports

Analysis of post irradiation examination of used BWR fuel with SWAT4.0

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$$_{rm eff}$$ 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.

Journal Articles

Analysis of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya

Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02

 Times Cited Count:2 Percentile:33.97(Nuclear Science & Technology)

The criticality safety of the fuel debris in Fukushima Daiichi Nuclear Power Plant is one of the most important issues and the adoption of the burnup credit is desired for the criticality analysis. The assay data of used nuclear fuel irradiated in 2F2 is evaluated to validate SWAT4.0 for BWR fuel burnup problem. The calculation results revealed that number density of many heavy nuclides and FPs showed good agreement with the experimental data except for $$^{235}$$U, $$^{237}$$Np, $$^{238}$$Pu and Sm isotopes. The cause of the difference is assumption of the initial number density and void ratio and overestimation of the capture cross section of $$^{237}$$Np. The C/E-1 values do not depend on the types of fuel rods (UO$$_{2}$$ or UO$$_{2}$$-Gd$$_{2}$$O$$_{3}$$) and it is similar to that for the PWR fuel. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of the BWR fuel and it has sufficient accuracy to be adopted in the burnup credit evaluation of the fuel debris.

Journal Articles

Another important piece; One point burnup calculation code as a Killer Application

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.

Journal Articles

SFCOMPO-2.0; An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

Michel-Sendis, F.*; Gauld, I.*; Martinez, J. S.*; Alejano, C.*; Bossant, M.*; Boulanger, D.*; Cabellos, O.*; Chrapciak, V.*; Conde, J.*; Fast, I.*; et al.

Annals of Nuclear Energy, 110, p.779 - 788, 2017/12

 Times Cited Count:31 Percentile:98.78(Nuclear Science & Technology)

Journal Articles

Cutting-edge studies on nuclear data for continuous and emerging need, 1; Diversifying nuclear applications and need for the nuclear data

Suyama, Kenya; Kunieda, Satoshi; Fukahori, Tokio; Chiba, Go*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 59(10), p.598 - 602, 2017/10

The nuclear data is the data on the reaction probability between the neutron and the nuclide in a narrow sense. However generally speaking, it is the data describing the physical change of the nuclide and the status of the nuclear ration. Since Japan had started the nuclear energy development, the nuclear data has been one of the most important technical development theme. Now, the nuclear data library of Japan, i.e., JENDL, is well recognized internationally because of the highest-accuracy and fully-furnished types of the included data. This serial lecture describes the significance and the status of the nuclear data development, the international trend, and the direction of the future development.

Journal Articles

FRENDY; A New nuclear date processing system being developed at JAEA

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

EPJ Web of Conferences, 146, p.02028_1 - 02028_5, 2017/09

 Times Cited Count:0 Percentile:0.08

JAEA has started to develop new nuclear data processing system FRENDY (FRom Evaluated Nuclear Data libralY to any application). In this presentation, the outline of the development of FRENDY is presented. And functions and performances of FRENDY are demonstrated by generation and validation of the continuous energy cross section data libraries for MVP, PHITS and MCNP codes.

Journal Articles

Development and verification of a new nuclear data processing system FRENDY

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

Journal of Nuclear Science and Technology, 54(7), p.806 - 817, 2017/07


 Times Cited Count:20 Percentile:95.35(Nuclear Science & Technology)

JAEA has developed an evaluated nuclear data library JENDL and several nuclear analysis codes such as MARBLE2, SRAC, MVP and PHITS. Though JENDL and these computer codes have been widely used in many countries, the nuclear data processing system to generate the data library for application programs had not been developed in Japan and foreign nuclear data processing systems, e.g., NJOY and PREPRO are used. To process the new library for JAEA's computer codes immediately and independently, JAEA started to develop the new nuclear data processing system FRENDY in 2013. In this paper, outline, function, and verification of FRENDY are described.

JAEA Reports

A Guide to introducing burnup credit, preliminary version (English translation)

Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu*

JAEA-Review 2017-010, 93 Pages, 2017/06


There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee.

Journal Articles

Atomic Energy Society of Japan 2017 Annual Meeting, joint session of "sigma advisory committee", "subcommittee on nuclear data" and "subcommittee on reactor physics"; Current status and future perspective of the Verification and Validation (V&V) of JENDL and neutronics calculation codes by use of the benchmark problems and integral experiments, 2; International benchmarks of OECD/NEA in the field of the neutronics calculation

Suyama, Kenya

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.

Journal Articles

Development of an automatic nuclear data validation system VACANCE

Tada, Kenichi; Suyama, Kenya

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 4 Pages, 2017/04

JAEA provides the evaluated nuclear data library JENDL. Usually, the integral experiments are used for the validation. Since this validation process takes long time and much effort, the automated system has been desired. To realize the automated system, nuclear data processing, analysis of the integral experiments and editing calculation results are required. With regard to the nuclear data processing, JAEA has started to develop the new nuclear data processing system FRENDY. Using FRENDY, the nuclear data can be automatically processed. Taking advantage of FRENDY, we developed the automatic nuclear data validation system VACANCE. VACANCE has many functions, e.g., searching and modifying the input file, available for the parallel computation and restart calculation, editing the calculation results. Combination of FRENDY and VACANCE enables us to carry out the efficient nuclear data validation cycle. In this presentation, the outline and functions of VACANCE are demonstrated.

Journal Articles

Examination of analytical method of rare earth elements in used nuclear fuel

Ozawa, Mayumi; Fukaya, Hiroyuki; Sato, Makoto; Kamohara, Keiko*; Suyama, Kenya; Tonoike, Kotaro; Oki, Keiichi; Umeda, Miki

Proceedings of 53rd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2016) (Internet), 9 Pages, 2016/11

Journal Articles

Re-structuring of the NEA Data Bank Management Committee; New Management Committee MBDAV

Suyama, Kenya

Kaku Deta Nyusu (Internet), (115), p.61 - 69, 2016/10

In recent years, discussion on the reform of the governing body of OECD/NEA Data Bank has been carried out. This document explains its background and outline.

Journal Articles

Re-structuring of the scientific program of the NEA (No.41, p.37 (1992))

Suyama, Kenya

Kaku Deta Nyusu (Internet), (115), p.70 - 79, 2016/10

In 1992, Dr. Yasuyuki Kikuchi, the general manager of the Nuclear Data Center of JAERI, presented his article "Re-structuring of the Scientific Program of NEA" in the Nuclear Data News volume 41. This is the English translation version of it.

Journal Articles

Burn-up credit criticality safety benchmark phase III-C; Nuclide composition and neutron multiplication factor of a boiling water reactor spent fuel assembly for burn-up credit and criticality control of damaged nuclear fuel

Suyama, Kenya; Uchida, Yuriko*; Kashima, Takao; Ito, Takuya*; Miyaji, Takamasa*

NEA/NSC/R(2015)6 (Internet), 253 Pages, 2016/03

The Expert Group on Burnup Credit Criticality Safety (EGBUC) of Working Party of Nuclear Criticality Safety (WPNCS) under the Nuclear Science Committee (NSC) of OECD/NEA has been assessing the accuracy of the burnup calculation code systems by organizing several international benchmarks. This Phase IIIC benchmark specification for BWR 9 by 9 type fuel assembly infinite two-dimensional model was proposed and approved in the meeting of the OECD/NEA/NSC/WPNCS Expert group on burnup credit criticality safety in September 2012 and distributed in October 2012 to the members of the WPNCS. We have set of thirty-five calculation results from sixteen institutes of nine countries. This report presents the results of the benchmark phase IIIC. By this benchmark results, we can confirm the certain progress of the burnup calculation capability than the time of Phase IIIB benchmark. The difference of the neutron multiplication factor generated by the difference of the burnup calculation results by the latest code systems is less than 3%.

Journal Articles

Current status and problems of computer codes for nuclear safety evaluation

Suyama, Kenya; Hirao, Yoshihiro*; Sakamoto, Hiroki*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 57(12), p.787 - 791, 2015/12

In the measures to pursue the world's highest level of safety of the nuclear installations, it is required to maintain the technical revel of the safety analyses codes as higher as possible. Because many of them were introduced from US at the initial phase of the nuclear energy introduction, development of computer codes and relevant tools in Japan have not been continued successfully. Accordingly, many old US-oriented code has been used. This manuscript presents the current status and the problems of computer codes for nuclear safety evaluation, and a scheme to introduce the computer codes of Japan, which in-cooperate the latest knowledge, in the scene of nuclear safety regulation and practical purpose.

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