Development of technology for separating Am-241 in aged plutonium

Emori, Tatsuya; Kitatsuji, Yoshihiro; Ban, Yasutoshi

JAEA-Technology 2024-025, 20 Pages, 2025/03

JAEA-Technology-2024-025.pdf:1.65MB

Radioisotope Thermoelectric Generators (RTGs) using the decay heat of Pu-238 has been applied for outer planet missions far from Jupiter, where solar power is limited. However, no facilities are available to produce Pu-238 for space probes in Japan. Moreover, the use of nuclear materials for the space exploration is difficult in term of the regulation. Thus, we focused on Am-241 whose half-life is around 432 years as an alternative heat source for RTGs. This report describes the procedure of separating Am-241 decayed from Pu-241 in aged plutonium oxide. Two experiments were performed: one using solid-liquid extraction and the other combining liquid-liquid extraction and solid-liquid extraction. Packed columns were used in the experiments, with their number reduced by less than one-fifth in the latter experiment compared to the former. Furthermore, the time required for separation in the latter experiment was less than half that of the former. We performed the separation experiments six times, collecting a total of approximately 0.43 g of Am-241 as an oxalate salt.

JENDL-5 benchmarking for advanced test reactor for preparing burnup analysis using isotopic data from HTGR type fuel irradiation tests

Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio

Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11

https://doi.org/10.1115/ICONE31-131748

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2022)

Department of HTTR

JAEA-Review 2024-034, 70 Pages, 2024/10

JAEA-Review-2024-034.pdf:3.22MB

This report summarizes the activities carried out in the fiscal year 2022 about the operation and maintenance of the High Temperature Engineering Test Reactor (HTTR), the R&Ds using the HTTR and so on. The HTTR is the first Japanese test reactor of High Temperature Gas-cooled Reactor (HTGR) type with 30MW in thermal power and whose maximum outlet coolant temperature achieved 950C. HTGRs are regarded as the promising candidates of the Next Generation Nuclear Plants conformed to the future decarbonized society because of the inherent safety characteristics as well as high temperature heat supply capability for not only a power generation but for wide-ranging industrial uses such as a hydrogen production and so on. The purpose of the HTTR is establishment of basic HTGR technologies, demonstration of HTGR safety characteristics and so on. The HTTR has had a lot of experience of HTGRs' operation and maintenance throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2022, we conducted maintenance of the HTTR such as countermeasures of differential pressure rise event for the primary helium gas circulator's filters occurred at an operation in the year 2021.

Current status of high temperature gas-cooled reactor development in Japan

Nagatsuka, Kentaro; Noguchi, Hiroki; Nagasumi, Satoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki

Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08

https://doi.org/10.1016/j.nucengdes.2024.113338

HTGR has a potential to contribute to decarbonization of hard-to-abate industries by supplying a large amount of hydrogen and high temperature heat or steam without carbon dioxide emission. JAEA has been conducting R&Ds for HTGR technologies with High Temperature Engineering Test Reactor (HTTR). This paper shows that HTTR's tests including the loss of core cooing test as a joint the OECD/NEA international research project and a HTTR heat application test plan which demonstrate hydrogen production by coupling the HTTR with a hydrogen production test facility. Additionally, aiming for operation start from the latter half of 2030s, the basic design of the HTGR demonstration reactor has been shown. The Japan's HTGR technology capabilities established by the HTTR project will be fully utilized for the construction of HTGR demonstration reactor.

Comparison on safety features among HTGR's Reactor Cavity Cooling Systems (RCCSs)

Takamatsu, Kuniyoshi; Funatani, Shumpei*

Nuclear Engineering and Technology, 56(3), p.832 - 845, 2024/03

https://doi.org/10.1016/j.net.2023.09.036

The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Therefore, the authors concluded that the proposed RCCS based on atmospheric radiation has the advantage that the temperature of the RPV can be stably maintained against disturbances in the outside air (ambient air). Moreover, methodology to utilize all the heat emitted from the RPV surface for increasing the degree of waste-heat utilization was discussed.

Development of an RPV cooling system for HTGRs

Takamatsu, Kuniyoshi

Kakushinteki Reikyaku Gijutsu; Mekanizumu Kara Soshi, Shisutemu Kaihatsu Made, p.179 - 183, 2024/01

The HTGR has excellent safety, and even in the event of an accident where the reactor coolant is lost, the decay heat and residual heat in the core can be dissipated from the outer surface of the RPV, so the fuel temperature never exceeds the limit value, and the core stabilizes. On the other hand, regarding the cooling system that transports the heat emitted from the RPV to the final heat sink, an active cooling system using forced circulation of water by a pump, etc., and a passive cooling system using natural circulation of the atmosphere have been proposed. However, there is a problem that the cooling performance is affected by the operation of dynamic equipment and weather conditions. This paper presents an overview of a new cooling system concept using radiative cooling, which has been proposed to solve the above problem, and introduces the results of analysis and experiments aimed at confirming the feasibility of this concept.

Development of coupling technology for high temperature gas-cooled reactors and hydrogen production facility; HTTR heat application test project plan

Ishii, Katsunori; Morita, Keisuke; Noguchi, Hiroki; Aoki, Takeshi; Mizuta, Naoki; Hasegawa, Takeshi; Nagatsuka, Kentaro; Nomoto, Yasunobu; Shimizu, Atsushi; Iigaki, Kazuhiko; et al.

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2023/09

https://doi.org/10.1299/jsmepes.2023.27.A242

Reactor physics experiment on a graphite-moderated core to construct integral experiment database for HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Nuclear Science and Engineering, 197(8), p.2251 - 2257, 2023/08

https://doi.org/10.1080/00295639.2022.2087836

Evaluation of failure fraction for high burnup High Temperature Gas-cooled Reactor fuel

Sawa, Kazuhiro*; Haseda, Masaya*; Aihara, Jun

Nihon Kikai Gakkai Rombunshu (Internet), 89(921), p.22-00314_1 - 22-00314_6, 2023/05

https://doi.org/10.1299/transjsme.22-00314

In high temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In the high burnup coated fuel particle, stress due to fission gas pressure and irradiation-induced pyrolytic carbon (PyC) shrinkage is introduced into the coating layers and consequently the stress could cause failure of coating layers under high burnup irradiation condition. A failure model has developed to predict failure fraction of TRISO-coated particle under high burnup irradiation. In the model, failure probability is strongly dependent on the irradiation characteristics of PyC. This paper describes the outline of the failure model and evaluation result of high burnup fuel irradiation experiment by the model.

Research on improvement of HTGR core power-density, 4; Feasibility study for a reactor core

Okita, Shoichiro; Mizuta, Naoki; Takamatsu, Kuniyoshi; Goto, Minoru; Yoshida, Katsumi*; Nishimura, Yosuke*; Okamoto, Koji*

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05