Release behaviors of elements from an Ag-In-Cd control rod alloy at temperatures up to 1673 K

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi*

Nuclear Technology, 208(3), p.484 - 493, 2022/03

https://doi.org/10.1080/00295450.2021.1905472

An Ag-In-Cd control rod alloy was heated in argon or oxygen at 1073-1673 K for 60-3600 s and the release behavior of the elements was examined. Complete liquefaction of the alloy occurred between 1123 and 1173 K, and elemental release was quite limited below the liquefaction temperature. In argon, almost all of the Cd content was released within 3600 s at 1173 K and within 60 s at 1573 K, while the released fractions of Ag and In were 3% and 8%, respectively. In oxygen, the release of Cd, which was quite small at temperatures up to 1573 K, drastically increased to 30-50% at 1673 K for short periods. Releases of Ag and In were also small in oxygen under the examined conditions. Comparison with the experimental data suggests that conventional empirical release models may underestimate the Cd release at lower temperatures just after control rod failure in severe accidents.

Impact of MOX fuel use in light-water reactors; Long-term radiological consequences of disposal of high-level waste in a geological repository

Minari, Eriko*; Kabasawa, Satsuki; Mihara, Morihiro; Makino, Hitoshi; Asano, Hidekazu*; Nakase, Masahiko*; Takeshita, Kenji*

Journal of Nuclear Science and Technology, 11 Pages, 2022/00

https://doi.org/10.1080/00223131.2022.2146016

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.

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

Modelling of cesium chemisorption under nuclear power plant severe accident conditions

Miradji, F.; Suzuki, Chikashi; Nishioka, Shunichiro; Suzuki, Eriko; Nakajima, Kunihisa; Osaka, Masahiko; Barrachin, M.*; Do, T. M. D.*; Murakami, Kenta*; Suzuki, Masahide*

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 21 Pages, 2019/03

Thermal-hydraulics technological strategy roadmap 2017; An Approach for continuous safety improvement of LWRs

Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*

Nihon Genshiryoku Gakkai-Shi ATOMO, 60(4), p.221 - 225, 2018/04

no abstracts in English

Compilation of the data book on light water reactor benchmark to develop the next version of JENDL; Utilization of criticality data in ICSBEP and IRPhEP open databases

JENDL Committee, Reactor Integral Test Working Group

JAEA-Data/Code 2017-006, 152 Pages, 2017/05

JAEA-Data-Code-2017-006.pdf:13.46MB
JAEA-Data-Code-2017-006-appendix1(CD-ROM).zip:115.88MB
JAEA-Data-Code-2017-006-appendix2(CD-ROM).zip:110.88MB

A benchmark database which is devoted to the evaluation of the future JENDL against the criticality of light water reactors was prepared, where the ICSBEP and IRPhEP handbooks by OECD/NEA were utilized effectively. Specific features of this report can be described as follows: (1) The recommendation for benchmarking is based on careful reviewing for the document and related information. Validity of the original benchmark evaluation is carefully checked, and numerical results obtained with JENDL-4.0 are considered. (2) Heterogeneity effect of PuO particles dispersed in fuel medium is consistently quantified for the MOX fuel-loaded experimental data. This precise evaluation is realized by the newly developed finite fuel pin bundle model with the Monte Carlo neutron transport code. (3) Sensitivity analysis is conducted in order to specify nuclear data whose difference between recent nuclear data libraries significantly affects the critical parameter calculation.

Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki*; Torii, Kazutaka*

JAEA-Research 2015-019, 90 Pages, 2016/01

JAEA-Research-2015-019.pdf:1.95MB

At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For the purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core.

Evaluation of Nuclear Knowledge Management; An Outcome in JAERI

Yanagisawa, Kazuaki

International Journal of Nuclear Knowledge Management, 2(2), p.91 - 104, 2006/00

https://doi.org/10.1504/IJNKM.2006.010286

Taking into consideration of national funds invested to individual research divisions in JAERI during the course of safety project studies, one found that a total input was approximately corresponded to . Qualitatively, JAERI contributed to (1) the governmental policies, (2) the scientific technologies and scholarship, (3) the industries (technology transfer, patents, commissioned research etc.), (4) the local society in the lab location area, and (5) the international cooperation. Quantitatively, JAERI made the creation of added value (cost benefit), cost reduction and reduction of cost loss produced either in direct or indirect manner as research outputs. One found that a total output during the course of nuclear safety project study was approximately taking into consideration of nuclear market creation at electricity and facilities related to nuclear. This ex post evaluation tells that cost benefit made by nuclear safety project of JAERI is 1.5 () and such outcome contributes to the increase of general domestic products (GDP).