Pilot study on thermal, physico-chemical, and mechanical behavior of concrete to understand the failure behavior of Fukushima Daiichi Nuclear Power Station reactor pressure vessel pedestals (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokai National Higher Education and Research System*

JAEA-Review 2025-034, 83 Pages, 2025/12

JAEA-Review-2025-034.pdf:6.9MB

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 FY2023. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), 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 FY2023, this report summarizes the research results of the "Pilot study on thermal, physico-chemical, and mechanical behavior of concrete to understand the failure behavior of Fukushima Daiichi Nuclear Power Station reactor pressure vessel pedestals" conducted in FY2023. The present study aims to examine the mechanism of the collapse of only concrete with rebar remaining in the pedestal in the containment vessel (PCV) of 1F. In verifying concrete-specific factors, (1) to clarify the short-term dissolution mechanism by high temperature, we investigated data acquisition methods in melting experiments, established an analytical framework for determining dissolution, and developed a numerical analysis method for volume change by heating. Additionally, (2) to clarify long-term dissolution mechanism by temperature history, we organized the temperature and water injection history, determined concrete exposure conditions during experiments, and established a method for selecting materials and measuring expansion. Furthermore, we summarized existing knowledge of the expansion phenomenon caused by water supply after high temperature heating. In the verification of special external environmental factors, (1) to evaluate thermal conditions of PCV concrete during an accident, a preliminary heat transfer analysis of fuel debris was conducted. In addition, (2) as elemental behavior tests and comprehensive tests, a preliminary high temperature storage test on concrete materials in a water vapor atmosphere and a preliminary reaction test on the reaction behavior of metal debris and concrete were conducted. Furthermore, uranium-containing suboxides were prepared. This study provided comprehensive insight into the mechanism of concrete failure in 1F Unit 1.

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04

https://doi.org/10.1115/1.4045450

Evaluation and demonstration of cutting the fuel assembly heating examination by AWJ

Maruyama, Shinichiro*; Watatani, Satoshi*

Mitsui Sumitomo Kensetsu Gijutsu Kenkyu Kaihatsu Hokoku, (15), p.107 - 112, 2017/10

https://www.smcon.co.jp/company/development-center/report-2017/

It is essential to estimate characteristics and forms of fuel debris for safe and reliable removing at the decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F). For the estimation, melting behavior of fuel assembly in the accident is being researched. To proceed the research, the fuel debris were need to cut, and the abrasive water jet (AWJ) which had enough results for cutting ceramic material or mixed material of zirconium alloy and stainless. The test results demonstrated that AWJ could cut the fuel assembly and accumulated the cutting data which will be subservient when removing the fuel debris in future.

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

https://doi.org/10.1115/1.4032478

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Computational and experimental examination of simulated core damage and relocation dynamics of a BWR fuel assembly

Hanus, G.*; Sato, Ikken; Iwama, Tatsuya*

Proceedings of International Waste Management Symposia 2016 (WM2016) (Internet), 12 Pages, 2016/03

JAEA plans a large-scale test to evaluate damage and relocation behavior of BWR core materials consisting of fuel rods, channel boxes, control blade and lower support structures. Its purpose is to contribute to understanding of core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing, JAEA desires preliminary investigations to examine melting test pieces. The purpose of such tests is to verify the materials and test piece will be heated by plasma to the target temperature (ca.2900K) and to collect data about the material relocation behavior. Results from preliminary computational simulations are presented illustrating the effectiveness of a 150 kW non-transferred plasma jet. An experimental test program using the computational analyses as a basis and a plasma torch is described.

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Development of a heat-resistant neutron shielding resin for the national centralized tokamak

Morioka, Atsuhiko; Sakurai, Shinji; Okuno, Koichi*; Tamai, Hiroshi

Purazuma, Kaku Yugo Gakkai-Shi, 81(9), p.645 - 646, 2005/09

http://doi.org/10.1585/jspf.81.645

A 300 C heat-resistant neutron shielding material is newly developed, which consists of phenol-based resin with 5 weight-% boron. The neutron shielding performance of the developed resin, examined by the Cf neutron source, is almost the same as that of the polyethylene. The resin is applicable to the port section of vacuum vessel of the DD plasma device to suppress the streaming neutrons and to reduce the nuclear heating of the superconducting coils.

Accelerator R&D for JT-60U and ITER NB systems

Inoue, Takashi; Hanada, Masaya; Iga, Takashi*; Imai, Tsuyoshi; Kashiwagi, Mieko; Kawai, Mikito; Morishita, Takatoshi; Taniguchi, Masaki; Umeda, Naotaka; Watanabe, Kazuhiro; et al.

Fusion Engineering and Design, 66-68, p.597 - 602, 2003/09

https://doi.org/10.1016/S0920-3796(03)00156-X

The neutral beam (NB) injection has been one of the most promising methods for plasma heating and current drive in tokamak fusion devices. JAERI has developed high energy electrostatic accelerators for the NB systems in JT-60U and ITER. Recent progress on this R&D are as follows: 1) In the JT-60U NB system, some of the beams has been deflected due to distorted electric field in the accelerator, resulting in an excess heat load on the NB port. By correcting the electric field, a continuous injection of H beam was succeeded for 10 s with the NB power of 2.6 MW at 355 keV. 2) To increase the beam energy, a metal structure called stress ring was designed. The ring reduces electric field concentration at the triple junction point (interface between metal and dielectric insulator inside vacuum). Initial test of the accelerators with the stress rings has shown higher voltage hold off performance in both accelerators for JT-60U and ITER R&D than that without rings.

Design of neutral beam system for ITER-FEAT

Inoue, Takashi; Di Pietro, E.*; Hanada, Masaya; Hemsworth, R. S.*; Krylov, A.*; Kulygin, V.*; Massmann, P.*; Mondino, P. L.*; Okumura, Yoshikazu; Panasenkov, A.*; et al.

Fusion Engineering and Design, 56-57, p.517 - 521, 2001/10

https://doi.org/10.1016/S0920-3796(01)00339-8

no abstracts in English

Premoderator extension effect for a decoupled hydrogen moderator

Harada, Masahide; Teshigawara, Makoto; Kai, Tetsuya; Sakata, Hideaki*; Watanabe, Noboru; Ikeda, Yujiro

JAERI-Research 2000-014, p.40 - 0, 2000/03

JAERI-Research-2000-014.pdf:2.99MB

no abstracts in English