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JAEA Reports

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.

Journal Articles

Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions; Results of BZ-3 and BZ-4 tests

Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Development methodology on determination of instant release fractions for generic safety assessment for direct disposal of spent nuclear fuel

Kitamura, Akira; Akahori, Kuniaki; Nagata, Masanobu*

Genshiryoku Bakkuendo Kenkyu (CD-ROM), 27(2), p.83 - 93, 2020/12

Direct disposal of spent nuclear fuel (SNF) in deep underground repositories (hereafter "direct disposal") is a concept that disposal canisters stored fuel assemblies dispose without reprocessing. Behavior of radionuclide release from SNF must be different from that from vitrified glass. The present study established a methodology on determination of instant release fraction (IRF) of radionuclides from SNF, which is the one of the parameters on radionuclide release based on the latest safety assessment reports in other countries, especially for IRF values proportional to a fission gas release ratio (FGR). Recommended and maximum values of FGR have been estimated using the fuel performance code FEMAXI-7 after collecting FGR values on Japanese SNFs. Furthermore, recommended and maximum values of IRF for Japanese SNFs used in a pressurized water reactor (PWR) have been estimated using the presently obtained FGR values and experimentally obtained IRF values on foreign SNFs. The recommended and maximum IRF values obtained in the present study have been compared with those of the latest safety assessment reports in other countries.

Journal Articles

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

Journal Articles

Feasibility study of tritium recoil barrier for neutron reflectors

Ishitsuka, Etsuo; Sakamoto, Naoki*

Physical Sciences and Technology, 6(2), p.60 - 63, 2019/12

Tritium release into the primary coolant of the research and test reactors during operation had been studied, and it is found that the recoil release from chain reaction of $$^{9}$$Be is dominant. To reduce tritium concentration of the primary coolant, feasibility study of the tritium recoil barrier for the beryllium neutron reflectors was carried out, and the tritium recoils of various materials were calculated by PHITS. From these calculation results, it is clear that the thickness of tritium recoil barrier depends on the material and 20$$sim$$40 $$mu$$m is required for three orders reduction.

Journal Articles

Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Journal Articles

Calculation of tritium release from driver fuels into primary coolant of research reactors

Ho, H. Q.; Ishitsuka, Etsuo

Physical Sciences and Technology, 5(2), p.53 - 56, 2019/00

Increasing of tritium concentration in the primary coolant of the research and test reactors during operation had been reported. To check the source for tritium release into the primary coolant during operation of the JMTR and the JRR-3M, the tritium release from the driver fuels was calculated by MCNP6 and PHITS. It is clear that the calculated values of tritium release from fuels are as about 10$$^{7}$$ and 10$$^{6}$$ Bq for the JMTR and JRR-3M, respectively, and that calculated values are about 4 order of magnitude smaller than that of the measured values. These results show that the tritium release from fuels is negligible for both the reactors.

JAEA Reports

Calculations of Tritium Recoil Release from Li and U Impurities in Neutron Reflectors (Joint research)

Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2018-010, 33 Pages, 2018/11

JAEA-Technology-2018-010.pdf:2.58MB

As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.

JAEA Reports

Comparison between HTFP code and minory changed FORNAX-A code

Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi

JAEA-Technology 2018-002, 70 Pages, 2018/06

JAEA-Technology-2018-002.pdf:1.46MB

HTFP code is code for calculation of additional release amount of fission product (FP) from fuel rod in high temperature gas-cooled reactor (HTGR) after stop of fission. Minory changed Fornax-A code also can calculate that. Therefore, release behavior of Cs calculated with HTFP code was compared with that calculated with minory modified FORNAX-A code in this report. Release constants of Cs evaluated with minory modified FORNAX-A code are rather different from default values for HTFP code.

Journal Articles

Evaluation of tritium release curve in primary coolant of research reactors

Ishitsuka, Etsuo; Kenzhina, I. E.*

Physical Sciences and Technology, 4(1), p.27 - 33, 2018/06

Increase of tritium concentration in the primary coolant for the research and testing reactors during reactor operation had been reported. To clarify the tritium sources, a curve of the tritium release rate into the primary coolant for the JMTR and the JRR-3M are evaluated. It is also observed that the amount of released tritium is lower in the case of new beryllium components installation, and increases with the reactor operating cycle. These results show the beryllium components in core strongly affect to the tritium release into the primary coolant. As a result, the tritium release rate is related with produced $$^{6}$$Li by (n,$$alpha$$) reaction from $$^{9}$$Be, and evaluation results of tritium release curve are shown as the dominant source of tritium release into the primary coolant for the JMTR and the JRR-3M are beryllium components. Scattering of the tritium release rate with irradiation time were observed, and this phenomena in the JMTR occurred in earlier time than that of the JRR-3M.

Journal Articles

Current state of atmospheric and oceanic environmental researches on the Fukushima Daiichi nuclear accident; What is known about/from the accident

Aoyama, Michio*; Yamazawa, Hiromi*; Nagai, Haruyasu

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(1), p.46 - 50, 2018/01

no abstracts in English

Journal Articles

Development of experimental and analytical technologies for fission product chemistry under LWR severe accident condition

Miyahara, Naoya; Miwa, Shuhei; Nakajima, Kunihisa; Osaka, Masahiko

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 9 Pages, 2017/09

This paper presents the development of a reproductive experimental setup for FP release and transport and an analysis tool considering chemical reaction kinetics for the construction of the FP chemistry database. The performance test of the reproductive experimental setup TeRRa using CsI compounds show that TeRRa can reproduce well a FP chemistry-related behavior such as aerosol formation, growth and deposition behavior. An analytical tool has been developed based on the commercial ANSYS-FLUENT code. Some additional models was added to evaluate detailed FP chemistry during release and transport in this study. A test analysis simulating the CsI heating test in steam atmosphere was carried out to demonstrate the performance of the improved code. The result shows the appropriateness of the additional models.

Journal Articles

Challenges for enhancing Fukushima environmental resilience, 1; Status and lessons

Miyahara, Kaname; Ohara, Toshimasa*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 59(5), p.282 - 286, 2017/05

This review highlights JAEA and NIES's challenges for enhancing Fukushima environmental resilience based on carrying out multifaceted research working with many public and private sector organizations and academia.

Journal Articles

An Overview of progress in environmental research on radioactive materials derived from the Fukushima Nuclear accident

Ohara, Toshimasa*; Miyahara, Kaname

Global Environmental Research (Internet), 20(1&2), p.3 - 13, 2017/03

Toward the environmental regeneration in Fukushima Prefecture and other areas after the Fukushima Daiichi Nuclear Power Station accidents, JAEA and NIES working with many public and private sector organizations and academia have carried out multifaceted research that will help to restore the environment of affected areas. These challenging efforts need to be further strengthened.

JAEA Reports

Calculation by PHITS code for recoil tritium release rate from beryllium under neutron irradiation (Joint research)

Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2016-022, 35 Pages, 2016/10

JAEA-Technology-2016-022.pdf:3.73MB

As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, the calculation methods by PHITS code is studied to evaluate the recoil tritium release rate from beryllium core components. Calculations using neutron and triton sources were compared, and it is clear that the tritium release rates in both cases show similar values. However, the calculation speed for the triton source cases is two orders faster than that for the neutron source case. It is also clear that the calculation up to history number per unit volume of 2$$times$$10$$^{4}$$ (cm$$^{-3}$$) is necessary to determine the recoil tritium release rate of two effective digits precision. Furthermore, the relationship between the beryllium shape and recoil tritium release rate using the triton sources was studied. Recoil tritium release rate showed linear relation to the surface area per volume of beryllium, and the recoil tritium release rate showed about half of the conventional equation value.

JAEA Reports

Solvent extraction and release behavior of ruthenium and europium in fire accident conditions in reprocessing plants (Contract research)

Amano, Yuki; Watanabe, Koji; Masaki, Tomoo; Tashiro, Shinsuke; Abe, Hitoshi

JAEA-Technology 2016-012, 21 Pages, 2016/06

JAEA-Technology-2016-012.pdf:1.81MB

To contribute to safety evaluation of fire accident in fuel reprocessing plants, solvent extraction behavior of ruthenium, which could form volatile species, was investigated. Distribution ratios of ruthenium at fire accident conditions were obtained by extraction experiments with several solvent composition at different temperature as parameters. In order to investigate release behavior of ruthenium and europium at fire accident, release ratios of ruthenium and europium were also obtained by solvent combustion experiments.

Journal Articles

Release of radioactive materials from high active liquid waste in small-scale hot test for boiling accident in reprocessing plant

Yamane, Yuichi; Amano, Yuki; Tashiro, Shinsuke; Abe, Hitoshi; Uchiyama, Gunzo; Yoshida, Kazuo; Ishikawa, Jun

Journal of Nuclear Science and Technology, 53(6), p.783 - 789, 2016/06

 Times Cited Count:3 Percentile:36.52(Nuclear Science & Technology)

The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory scale reprocessing, release ratio was measured for the FP nuclides such as Ru, $$^{99}$$Tc, Cs, Sr, Nd, Y, Mo, Rh and actinides such as $$^{242}$$Cm, $$^{241}$$Am. As a result, the release ratio was 0.20 for Ru and 1$$times$$$$10^{-4}$$ for the FP and Ac nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to the initial concentration in the test solution. The release ratio decreased with the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the initial concentration. The release ratio of Ru and NOx concentration increased with temperature of the test solutions. They were released almost at the same temperature between 200 and 300$$^{circ}$$C. Size distribution of the mist and other particle was measured.

Journal Articles

Formation and release of molecular iodine in aqueous phase chemistry during severe accident with seawater injection

Kido, Kentaro; Hata, Kuniki; Maruyama, Yu; Nishiyama, Yutaka; Hoshi, Harutaka*

NEA/CSNI/R(2016)5 (Internet), p.204 - 212, 2016/05

Journal Articles

Recent research activities using NSRR on safety related issues

Udagawa, Yutaka; Sugiyama, Tomoyuki*; Amaya, Masaki

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04

206 (Records 1-20 displayed on this page)