Decommissioning report for Wastewater Treatment Facility (Part 2); Chapter on contamination inspection section

Yamamoto, Keisuke; Nakagawa, Takuya; Shimojo, Hiroto; Kijima, Jun; Miura, Daiya; Onose, Yoshihiko*; Namba, Koji*; Uchida, Hiroaki*; Sakamoto, Kazuhiko*; Ono, Chika*; et al.

JAEA-Technology 2024-019, 211 Pages, 2025/02

JAEA-Technology-2024-019.pdf:35.35MB

The uranium enrichment facilities at the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency (JAEA) were constructed sequentially to develop uranium enrichment technology with centrifugal separation method. The developed technologies were transferred to Japan Nuclear Fuel Limited until 2001. And the original purpose has been achieved. Wastewater Treatment Facility, one of the uranium enrichment facilities, was constructed in 1976 to treat radioactive liquid waste generated at the facilities, and it finished the role in 2008. In accordance with the Medium/Long-Term Management Plan of JAEA Facilities, interior equipment installed in this facility had been dismantled and removed since November 2021 to August 2023. This report summarizes the findings obtained through the work related to the contamination inspection methods cancellation the controlled area of Wastewater Treatment Facility from September 2023 to March 2024.

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.

Development of safety requirements for HTGRs design

Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tokuhara, Kazumi; Nishihara, Tetsuo; Kunitomi, Kazuhiko

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.330 - 340, 2016/11

The safety requirements for the design of HTGRs has been developed by the research committee established in the Atomic Energy Society of Japan so as to incorporate the HTGR safety features demonstrated by HTTR, lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the coupling of the hydrogen production plants with nuclear plant. The safety design approach was determined to establish a high level of safety design standards by utilizing inherent safety features of HTGRs. This paper describes the process to develop the HTGR specific safety requirements and overview of the proposed HTGR specific safety requirements.

Characteristic confirmation test by using HTTR and investigation of absorbing thermal load fluctuation

Honda, Yuki; Tochio, Daisuke; Sato, Hiroyuki; Nakagawa, Shigeaki; Ono, Masato; Fujiwara, Yusuke; Hamamoto, Shimpei; Iigaki, Kazuhiko; Takada, Shoji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 5 Pages, 2016/06

https://doi.org/10.1115/ICONE24-60355

The characteristic confirmation test has been demonstrating by using the High Temperature engineering Test Reactor (HTTR). The thermal load fluctuation test, which is one of marginal performance test is planned to be carried out after restarting of the HTTR. The preliminary analysis for the thermal load fluctuation test has been investigated. In the analysis, the reactor outlet temperature can continue to be stable against the reactor inlet temperature changing by thermal fluctuation. It means that HTGR have the capability of absorbing thermal fluctuation. This paper focuses on the investigation of mechanism of absorbing thermal fluctuation. With additional analysis, it is cleared that the large negative graphite moderator reactivity enhances the capability of absorbing thermal fluctuation. In addition, in the middle of the core, graphite moderator reactivity insertion trend are inverted. This trend is unique to HTGR because of large temperature difference between core inlet and outlet.

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.

Validation and application of thermal hydraulic system code for analysis of helically coiled heat exchanger in high-temperature environment

Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

Journal of Nuclear Science and Technology, 51(11-12), p.1324 - 1335, 2014/11

https://doi.org/10.1080/00223131.2014.928605

A qualification of the thermal hydraulic system code RELAP5 code is conducted for the analysis of helically-coiled heat exchangers used in high temperature environment. The experimental data obtained from the HTTR are utilized to compare with calculated data by RELAP5-based model with built-in closure models. A set of closure model is also suggested considering the heat transfer enhancement by thermal radiation based on the past separate effect test data and validated against the measured data. In addition, the modified RELAP5 code is tested for the analysis of the HTTR-IS system. The comparison of calculated and measure data with steady state operation showed that the prediction temperature with the suggested model generally agreed well. As a conclusion of the present study, the use of thermal hydraulic system code with the suggested closure model is acceptable for the analysis of the IHX in HTGR nuclear hydrogen production systems in the safety evaluation.

Spontaneous stabilization of HTGRs without reactor scram and core cooling; Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

Takamatsu, Kuniyoshi; Yan, X.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

Nuclear Engineering and Design, 271, p.379 - 387, 2014/05

https://doi.org/10.1016/j.nucengdes.2013.12.005

In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the strength of the neutron source, the larger the minimum reactor power.

A Proposal for safety design philosophy of HTGR for coupling hydrogen production plant

Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Imai, Yoshiyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

JAEA-Technology 2013-015, 68 Pages, 2013/06

JAEA-Technology-2013-015.pdf:2.22MB

In present study, requirements in order to design, construct and operate hydrogen production plants coupled to HTGRs under conventional chemical plant standards are identified. In addition, design considerations for safety design of nuclear facility are suggested. Furthermore, feasibility of proposed safety design and design considerations are clarified.

Status and perspective of the NbAl development

Takeuchi, Takao*; Kikuchi, Akihiro*; Banno, Nobuya*; Kitaguchi, Hitoshi*; Iijima, Yasuo*; Tagawa, Kohei*; Nakagawa, Kazuhiko*; Tsuchiya, Kiyosumi*; Mitsuda, Shiori*; Koizumi, Norikiyo; et al.

Cryogenics, 48(7-8), p.371 - 380, 2008/07

https://doi.org/10.1016/j.cryogenics.2008.04.007

NbAl has advantages of better tolerance to strain/stress and a higher critical magnetic field (30 T at 4.2 K) for stoichiometric composition over NbSn. The rapid-heating, quenching and transformation annealing (RHQT) process enables to form a stoichiometric NbAl with fine grain structures via metastable bcc supersaturated-solid-solution. As a result a large critical current density of NbAl is achieved over the whole range of magnetic fields without trading off the excellent strain tolerance. A long-length of RHQ processing has been established, and a rectangular but Cu stabilized NbAl strand is about be commercially available for NMR uses. Ag or Cu internal stabilization and Cu ion-plating/electroplating techniques have been also developed to enable the stabilized round wire for accelerator and fusion magnets. Successfully energized test coils that were manufactured with a wind-and-react technique have demonstrated that a long piece of Cu stabilized RHQT NbAl wire is really available for practical applications.

Integrated on-line plant monitoring system for HTTR with neural networks

Nabeshima, Kunihiko; Subekti, M.*; Matsuishi, Tomomi*; Ono, Tomio*; Kudo, Kazuhiko*; Nakagawa, Shigeaki

Journal of Power and Energy Systems (Internet), 2(1), p.92 - 103, 2008/00

https://doi.org/10.1299/jpes.2.92

The neural networks have been utilized in on-line monitoring-system of High Temperature Engineering Tested Reactor (HTTR) with thermal power of 30 MW. In this system, several neural networks can independently model the plant dynamics with different architecture, input and output signals and learning algorithm. One of main task is real-time plant monitoring by Multi-Layer Perceptron (MLP) in auto-associative mode, which can model and estimate the whole plant dynamics by training normal operational data only. Other tasks are on-line reactivity prediction, reactivity and helium leak monitoring, respectively. From the on-line monitoring results at the safety demonstration tests, each neural network shows good prediction and reliable detection performances.