Initial verification and validation of a new CASMO5 JENDL-5 nuclear data library for typical LWR applications

Watanabe, Tomoaki; Suyama, Kenya; Tada, Kenichi; Ferrer, R. M.*; Hykes, J.*; Wemple, C. A.*

Nuclear Science and Engineering, 198(11), p.2230 - 2239, 2024/11

https://doi.org/10.1080/00295639.2023.2295075

A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, many essential nuclides for conventional LWR analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5-based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1-based library as possible with JENDL-5. This study verified and validated the new library. Verifications were performed based on the OECD/NEA burnup credit criticality safety benchmark phase III-C, and the calculated k and fuel compositions of the BWR fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, the TCA critical experiment and Takahama-3 post-irradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1-based library in predictions of reactivity and fuel compositions for LWR systems.

Proposal on how to proceed with Verification and Validation of radiation shielding analyses

Okumura, Keisuke; Sakamoto, Yukio*; Tsukiyama, Toshihisa*

Shahei Kaiseki No V&V Gaidorain Sakutei Ni Mukete, p.4 - 8, 2023/03

https://www.aesj.net/uploads/com_vvshieldingcalc20230428_sup.pdf

no abstracts in English

Multi-group cross section library generation by FRENDY for fast reactor neutronics calculations

Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi; Endo, Tomohiro*

Transactions of the American Nuclear Society, 124(1), p.556 - 558, 2021/06

The FRENDY nuclear data processing code has been used to generate multi-group cross section libraries for the CBZ reactor physics code system. The newly generated libraries have been applied to neutronics calculations of a fast reactor core MET-1000, and several neutronics parameters are calculated. Calculations with other libraries generated by NJOY2016 have been also conducted, and differences in obtained neutronics parameters between the FRENDY-based library and the NJOY-based library have been quantified. Generally reasonable agreement between them has been obtained, so it has been demonstrated that the multi-group libraries for fast reactor neutronics calculations can be generated successfully by FRENDY. Detailed investigation on the impact of the difference in the processing codes on k-effective has been also carried out with a help of the perturbation theory, and the causes of the differences have been identified.

Establishment of guideline for credibility assessment of nuclear simulations in the Atomic Energy Society of Japan

Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08

Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.

State-of-the-art approach and issue to establish simulation credibility

Nakada, Kotaro*; Kudo, Yoshiro*; Koshizuka, Seiichi*; Tanaka, Masaaki

Nihon Genshiryoku Gakkai-Shi ATOMO, 60(3), p.173 - 177, 2018/03

The Atomic Energy Society of Japan (AESJ) published "Guideline for Credibility Assessment of Nuclear Simulations 2015" in June, 2016 which specifies the concepts on methodology for the prediction with uncertainty quantification and the quality management based on the concept of verification and validation (V&V) of modeling and simulation. In this report, the outlines of activities in AESJ for publication of the guideline and the expectation for effective implementation of the guideline are described including that of the lectures with major respondents of the questionnaires.

Comparative study on burnup characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

Verification of alternative dew point hygrometer for CV-LRT in MONJU; Short- and long-term verification for capacitance-type dew point hygrometer (Translated document)

Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*

JAEA-Research 2017-001, 40 Pages, 2017/03

JAEA-Research-2017-001.pdf:5.19MB

In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (2.04C) required by the JEAC4203-2008.

Comparative study on neutronics characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 6 Pages, 2016/06

https://doi.org/10.1115/ICONE24-60455

Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, core characteristic parameters at the beginning of cycle were evaluated by the best estimate deterministic and stochastic methodologies of ANL and JAEA. The results obtained by both institutions are agreed well with less than 200 pcm of discrepancy on the neutron multiplication factor, and less than 3% of discrepancy on the sodium void reactivity, Doppler reactivity, and control rod worth. The results by the stochastic and deterministic were compared in each party to investigate impacts of the deterministic approximation and to understand potential variations in the results due to different calculation methodologies employed. Impacts of the nuclear data libraries were also investigated using a sensitivity analysis methodology.

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.8141 - 8155, 2015/08

In this paper, the authors focus on four kinds of thermal-hydraulic issues associated with the SDC, i.e. fuel assembly thermal-hydraulics, natural circulation decay heat removal, thermal striping phenomena, and core disruptive accidents, and provide a description of their evaluation method developments including verification and validation and necessary experimental studies for the Japan Sodium-cooled Fast Reactor (JSFR). These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all phenomena envisioned in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing down of knowledge/technologies.

Recent activities on clearance by IAEA and Japanese organizations

Okoshi, Minoru

Dekomisshoningu Giho, (31), p.32 - 44, 2005/03

http://www.randec.or.jp/publish/documents/gihou/Decommissioning%20gihou_31.pdf

The International Atomic Energy Agency (IAEA) published a Safety Guide (RS-G-1.7) on the application of the concepts of exclusion, exemption and clearance in 2004. In this Safety Guide, the IAEA proposes the activity concentrations can be applied to clearance. The Nuclear Safety Commission reevaluated their clearance levels to reflect the new technological information given in the IAEA Safety Guide and other reports. The Regulatory Authorities have been investigating and discussing regulatory rules on clearance including the verification system of clearance levels. And the Atomic Society of Japan has deliberated the technological standards for the verification of clearance levels performed by nuclear operators. In this paper the activities and outcomes by those organizations are outlined, and the issues to be solved on clearance are listed and the basic ideas for solution are discussed.