Impact of nuclear data revised from JENDL-4.0 to JENDL-5 on PWR spent fuel nuclide composition

Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 60(11), p.1386 - 1396, 2023/11

https://doi.org/10.1080/00223131.2023.2201603

The burnup calculations for estimating the nuclide composition of the spent fuel are highly dependent on nuclear data. Many nuclides in the latest version of the Japanese Evaluated Nuclear Data Library JENDL-5 were modified from JENDL-4.0 and the modification affects the burnup calculations. This study confirmed the validity of JENDL-5 in the burnup calculations. The PIE data of Takahama-3 was used for the validation. The effect of modifications of the parameters, e.g., cross sections and fission yields, from JENDL-4.0 to JENDL-5 on the nuclide compositions was quantitatively investigated. The calculation results showed that JENDL-5 has a similar performance to JENDL-4.0. The calculation results also revealed that the modifications of the cross sections of actinide nuclides, fission yields, and thermal scattering low data of hydrogen in HO affected the nuclide compositions of PWR spent fuels.

Hierarchical Bayes model to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under LOCA conditions

Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka

Proceedings of Asian Symposium on Risk Assessment and Management 2022 (ASRAM 2022) (Internet), 11 Pages, 2022/12

Depletion calculation of subcritical system with consideration of spontaneous fission reaction

Riyana, E. S.; Okumura, Keisuke; Sakamoto, Masahiro; Matsumura, Taichi; Terashima, Kenichi

Journal of Nuclear Science and Technology, 59(4), p.424 - 430, 2022/04

https://doi.org/10.1080/00223131.2021.1974970

Study on mechanism and threshold conditions for fuel fragmentation during loss-of-coolant accident conditions

Narukawa, Takafumi; Udagawa, Yutaka

Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10

HTTR burnup characteristic analysis with detailed axial burning region using MVP-BURN

Ikeda, Reiji*; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo; Fujimoto, Nozomu*

JAEA-Technology 2021-015, 32 Pages, 2021/09

JAEA-Technology-2021-015.pdf:2.74MB

Burnup calculation of the HTTR considering temperature distribution and detailed burning regions was carried out using MVP-BURN code. The results show that the difference in k, as well as the difference in average density of some main isotopes, is insignificant between the cases of uniform temperature and detailed temperature distribution. However, the difference in local density is noticeable, being 6% and 8% for U and Pu, respectively, and even 30% for the burnable poison B. Regarding the division of burning regions to more detail, the change of k is also small of 0.6%k/k or less. The small burning region gives a detailed distribution of isotopes such as U, Pu, and B. As a result, the effect of graphite reflector and the burnup behavior could be evaluated more clearly compared with the previous study.

Improvement of intragranular fission gas behavior model for fuel performance code FEMAXI-8

Udagawa, Yutaka; Tasaki, Yudai

JAEA-Data/Code 2021-007, 56 Pages, 2021/07

JAEA-Data-Code-2021-007.pdf:5.05MB

Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.

Technical basis of ECCS acceptance criteria for light-water reactors and applicability to high burnup fuel

Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki

JAEA-Review 2020-076, 129 Pages, 2021/03

JAEA-Review-2020-076.pdf:3.9MB

Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

Research and development on high burnup HTGR fuels in JAEA

Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X.

Mechanical Engineering Journal (Internet), 7(3), p.19-00571_1 - 19-00571_12, 2020/06

https://doi.org/10.1299/mej.19-00571

JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/t. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

https://doi.org/10.1080/00223131.2019.1637795

no abstracts in English