Human resource development related to remote control technology for monitoring inside rpv pedestal during retrieval of fuel debris (Contract research); FY2022 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2023-023, 99 Pages, 2024/03

JAEA-Review-2023-023.pdf:6.0MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2022. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Human Resource Development Related to Remote Control Technology for Monitoring Inside RPV Pedestal during Retrieval of Fuel Debris" conducted in FY2022. The present study aims to construct a monitoring platform for understanding the status inside a reactor during fuel debris removal, and measurement and visualization by sensors moving on the platform. In addition, to develop research personnel through research education by participating in such research projects, classroom lectures, and facility tours is also a goal of this project. In FY2022, each system was improved and expanded toward the final year, and verification experiments were conducted in simulated environments.

Analysis of the radioactivity concentrations in radioactive waste generated from JRR-3, JRR-4 and JRTF facilities, 2

Tobita, Minoru*; Goto, Katsunori*; Omori, Takeshi*; Osone, Osamu*; Haraga, Tomoko; Aono, Ryuji; Konda, Miki; Tsuchida, Daiki; Mitsukai, Akina; Ishimori, Kenichiro

JAEA-Data/Code 2023-011, 32 Pages, 2023/11

JAEA-Data-Code-2023-011.pdf:0.93MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field as trench and pit. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to the study of radioactivity concentration evaluation methods for radioactive wastes generated from nuclear research facilities, we collected and analyzed concrete samples generated from JRR-3, JRR-4 and JAERI Reprocessing Test Facility. In this report, we summarized the radioactivity concentrations of 23 radionuclides (H, C, Cl, Ca, Co, Ni, Sr, Nb, Ag, Cs, Ba, Eu, Eu, Ho, U, U, U, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2021-2022.

Alloy design and characterization of a recrystallized FeCrAl-ODS cladding for accident-tolerant BWR fuels; An Overview of research activity in Japan

Ukai, Shigeharu; Sakamoto, Kan*; Otsuka, Satoshi; Yamashita, Shinichiro; Kimura, Akihiko*

Journal of Nuclear Materials, 583, p.154508_1 - 154508_24, 2023/09

https://doi.org/10.1016/j.jnucmat.2023.154508

Data report of ROSA/LSTF experiment IB-HL-01; 17% hot leg intermediate break LOCA with totally-failed high pressure injection system

Takeda, Takeshi

JAEA-Data/Code 2023-007, 72 Pages, 2023/07

JAEA-Data-Code-2023-007.pdf:3.24MB

An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.

Analysis of the radioactivity concentrations in radioactive waste generated from JRR-3, JRR-4 and JRTF facilities

Tobita, Minoru*; Konda, Miki; Omori, Takeshi*; Nabatame, Tsutomu*; Onizawa, Takashi*; Kurosawa, Katsuaki*; Haraga, Tomoko; Aono, Ryuji; Mitsukai, Akina; Tsuchida, Daiki; et al.

JAEA-Data/Code 2022-007, 40 Pages, 2022/11

JAEA-Data-Code-2022-007.pdf:1.99MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete, ash, ceramic and brick samples generated from JRR-3, JRR4 and JRTF facilities. In this report, we summarized the radioactivity concentrations of 24 radionuclides (H, C, Cl, Ca, Co, Ni, Sr, Nb, Tc, Ag, I, Cs, Ba, Eu, Eu, Ho, U, U, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2020-2021.

Human resource development related to remote control technology for monitoring inside RPV pedestal during retrieval of fuel debris (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2021-030, 79 Pages, 2021/12

JAEA-Review-2021-030.pdf:3.82MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Human resource development related to remote control technology for monitoring inside RPV pedestal during retrieval of fuel debris" conducted in FY2020. This study deals with construction of a monitoring platform for understanding the status inside a reactor during fuel debris removal, and measurement and visualization by sensors moving on the platform. In addition, to develop research personnel through research education by participating in such research projects, classroom lectures, and facility tours is also a goal of this project. In FY2020, the main activities were base design and underlying technology development.

Data report of ROSA/LSTF experiment SB-PV-09; 1.9% pressure vessel top small break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2021-006, 61 Pages, 2021/04

JAEA-Data-Code-2021-006.pdf:2.78MB

An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.

Human resource development related to remote control technology for monitoring inside RPV pedestal during retrieval of fuel debris (Contract research); FY2019 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2020-028, 68 Pages, 2021/01

JAEA-Review-2020-028.pdf:4.01MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2019. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Human Resource Development Related to Remote Control Technology for Monitoring Inside RPV Pedestal during Retrieval of Fuel Debris". This study deals with construction of a monitoring platform for understanding the status inside a reactor during fuel debris removal, and measurement and visualization by sensors moving on the platform. In addition, to develop research personnel through research education by participating in such research projects, classroom lectures, and facility tours is also a goal of this project. In FY2019, the main activities were conceptual design, prototyping, and conceptual planning.

Major outcomes through recent ROSA/LSTF experiments and future plans

Takeda, Takeshi; Wada, Yuki; Shibamoto, Yasuteru

World Journal of Nuclear Science and Technology, 11(1), p.17 - 42, 2021/01

https://doi.org/10.4236/wjnst.2021.111002

Overview of accident-tolerant fuel R&D program in Japan

Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09

After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.