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

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.

Criticality configuration design methodology applied to the design of fuel debris experiment in the new STACY

Gunji, Satoshi; Tonoike, Kotaro; Clavel, J.-B.*; Duhamel, I.*

Journal of Nuclear Science and Technology, 58(1), p.51 - 61, 2021/01

https://doi.org/10.1080/00223131.2020.1798825

The new critical assembly STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core designs are in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of the new STACY core to measure the criticality characteristics of pseudo fuel debris that simulated Molten Core Concrete Interaction (MCCI) of the fuel debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes of interest and their k sensitivity to their cross sections. In the case of the fuel debris described in this study, especially for the concrete composition, silicon is the nucleus with the highest k sensitivity to the cross section. For this purpose, some parameters of the core configuration, as for example the lattice pitches or the core dimensions, were adjusted using optimization algorithm to find efficiently the optimal core configurations to obtain high sensitivity of silicon capture cross section. Based on these results, realistic series of experiments for fuel debris in the new STACY could be defined to obtain an interesting feedback for the MCCI. This methodology is useful to design other experimental conditions of the new STACY.

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

Technical design of the pressure-resistant chamber for open inspections of the storage containers of nuclear fuel materials

Marufuji, Takato; Sato, Takumi; Ito, Hideaki; Suzuki, Hisashi; Fujishima, Tadatsune; Nakano, Tomoyuki

JAEA-Technology 2019-006, 22 Pages, 2019/05

JAEA-Technology-2019-006.pdf:2.84MB

Radioactive contamination incident occurred at Plutonium Fuel Research Facility (PFRF) in Oarai Research and Development Institute, Japan Atomic Energy Agency on June 6, 2017. During inspection work of storage container containing nuclear fuel materials, the PVC bag packaging in the storage container ruptured when a worker opened the lid in the hood, and a part of contents was spattered over the room. The cause of the increase of internal pressure of the storage container was gas generation by alpha radiolysis of the epoxy resin mixed with nuclear fuel materials. Opening inspection of about 70 similar containers stored in PFRF has been planned to confirm the condition of the contents and to stabilize the stored materials containing organic compounds. For safe and reliable open inspection of the storage containers with high internal pressure in the glove box, it is necessary to develop a pressure-resistant chamber in which the storage containers are opened and the contents are inspected under gastight condition. This report summarizes the concerns and countermeasures of the chamber design and the design results of the chamber.

Study on Pu-burner high temperature gas-cooled reactor in Japan; Design study of fuel and reactor core

Goto, Minoru; Aihara, Jun; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

JAEA has conducted design studies of a Pu-burner HTGR. The Pu-burner HTGR incinerates Pu by fission, and hence a high burn-up is required for the efficient incineration. In the fuel design, a thin ZrC layer, which acts as an oxygen getter and suppresses the internal pressure, was coated on the fuel kernel to prevent the CFP failure at the high burn-up. A stress analysis of the SiC layer, which acts as a pressure vessel for the CFP, was performed for with consideration of the depression effect due to the ZrC layer. As a result, the CFP failure fraction at high burn-up of 500 GWd/t satisfied the target value. In the reactor core design, an axial fuel shuffling was employed to attain the high burn-up, and the nuclear burn-up calculations with the whole core model and the fuel temperature calculations were performed. As a result, the nuclear characteristics, which are the shutdown margin and the temperature coefficient of reactivity, and the fuel temperature satisfied their target values.

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

Design study of a 750 MWe Japan sodium-cooled fast reactor with metal fuel

Ohgama, Kazuya; Ota, Hirokazu*; Ikusawa, Yoshihisa; Oki, Shigeo; Ogata, Takanari*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

Development of fuel temperature calculation code for HTGRs

Inaba, Yoshitomo; Nishihara, Tetsuo

Annals of Nuclear Energy, 101, p.383 - 389, 2017/03

https://doi.org/10.1016/j.anucene.2016.11.006

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.

Development of fuel temperature calculation file for high temperature gas-cooled reactors

Inaba, Yoshitomo; Isaka, Kazuyoshi; Fukaya, Yuji; Tachibana, Yukio

JAEA-Data/Code 2014-023, 64 Pages, 2015/01

JAEA-Data-Code-2014-023.pdf:7.15MB

The Japan Atomic Energy Agency has performed the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the deployment of the systems to overseas such as developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. In the core thermal and hydraulic designs of HTGRs, it is important to evaluate the maximum fuel temperature so that the thermal integrity of the fuel is ensured. In order to calculate and evaluate the fuel temperature on personal computers (PCs) in a convenient manner, the calculation file based on the Microsoft Excel were developed. In this report, the basic equations used in the calculation file, the calculation method and procedure, and the results of the validation calculation are described.