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.

Corrigendum to "Neutron diffraction study on the deuterium composition of nickel deuteride at high temperatures and high pressures" [Phys. B Condens. Matter. 587 (2020) 412153]

Saitoh, Hiroyukki*; Machida, Akihiko*; Hattori, Takanori; Sano, Asami; Funakoshi, Kenichi*; Sato, Toyoto*; Orimo, Shinichi*; Aoki, Katsutoshi*

Physica B; Condensed Matter, 714, p.417234_1 - 417234_3, 2025/10

https://doi.org/10.1016/j.physb.2025.417234

Corrigendum to "Neutron diffraction study on the deuterium composition of nickel deuteride at high temperatures and high pressures" [Phys. B Condens. Matter. 587 (2020) 412153] was reported.

Density of a molten stainless steel-BC alloy measured in the electrostatic levitation furnace onboard the international space station

Ishikawa, Takehiko*; Oda, Hirohisa*; Koyama, Chihiro*; Shimonishi, Rina*; Ikeuchi, Rumiko*; Paradis, P.-F.*; Okada, Jumpei*; Fukuyama, Hiroyuki*; Yamano, Hidemasa

International Journal of Microgravity Science and Application, 42(2), p.420202_1 - 420202_10, 2025/04

https://doi.org/10.15011/jasma.42.420202

Microstructural observation of simulated fuel kernels for Pu-burner high temperature gas-cooled reactor in Japan

Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*

JAEA-Research 2024-012, 98 Pages, 2025/02

JAEA-Research-2024-012.pdf:32.24MB

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.

High-temperature strength of modified type 316 steel for fast reactor fuel before and after neutron irradiation

Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji

JAEA-Technology 2024-009, 140 Pages, 2024/10

JAEA-Technology-2024-009.pdf:8.03MB

For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.

Countermeasure against Beyond Design Basis Accident of HTTR by using fire engine

Shimazaki, Yosuke; Jidaisho, Tatsuya; Ishii, Toshiaki; Inoi, Hiroyuki; Iigaki, Kazuhiko

JAEA-Technology 2024-005, 23 Pages, 2024/06

JAEA-Technology-2024-005.pdf:5.53MB

HTTR has newly assumed Beyond Design Basis Accident (BDBA) as part of conformity assessment with the new regulatory standards and has established measures to prevent the spread of BDBA. Among these measures, to prevent the spread of BDBA caused by cooling water leaks from spent fuel storage pool, the Oarai Research Institute's fire engine was selected as an equipment to prevent the spread of BDBA, and required performances such as pumping water performance were determined. After all required performances were confirmed by inspections, the fire engine passed the operator's pre-use inspection and contributed to the restart of the HTTR operations.

Effect of dissolved oxygen concentration on dynamic strain aging and stress corrosion cracking of SUS304 stainless steel under high temperature pressurized water

Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*

The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01

https://doi.org/10.1063/5.0192888

Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.

New market opened up by advanced nuclear reactors (Chapter 3, 4, 5, 7)

Kamide, Hideki; Kawasaki, Nobuchika; Hayafune, Hiroki; Kubo, Shigenobu; Chikazawa, Yoshitaka; Maeda, Seiichiro; Sagayama, Yutaka; Nishihara, Tetsuo; Sumita, Junya; Shibata, Taiju; et al.

Jisedai Genshiro Ga Hiraku Atarashii Shijo; NSA/Commentaries, No.28, p.14 - 36, 2023/10

Developments of next generation nuclear reactors, e.g., Fast Reactor, and High Temperature Gas cooled Reactor, are in progress. They can contribute to markets of electricity and industrial heat utilization in the world including Japan. Here, current status of reactor developments in Japan and also situation in the world are summarized, especially for activities of Generation IV International Forum (GIF), developments of Fast Reactor and High Temperature Gas cooled Reactor in Japan, and SMR movements in the world.

High-temperature gaseous reaction of cesium with siliceous thermal insulation; The Potential implication to the provenance of enigmatic Fukushima cesium-bearing material

Rizaal, M.; Nakajima, Kunihisa; Saito, Takumi*; Osaka, Masahiko; Okamoto, Koji*

ACS Omega (Internet), 7(33), p.29326 - 29336, 2022/08

https://doi.org/10.1021/acsomega.2c03525

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

https://doi.org/10.1016/C2018-0-05383-9

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.