Conceptual design study of a high performance commercial HTGR for early introduction

Fukaya, Yuji; Mizuta, Naoki; Goto, Minoru; Ohashi, Hirofumi; Yan, X. L.

Nuclear Engineering and Design, 361, p.110577_1 - 110577_6, 2020/05

https://doi.org/10.1016/j.nucengdes.2020.110577

Conceptual design study of a commercial High Temperature Gas-cooled Reactor (HTGR) for early introduction has been performed based on the cumulated experience in design, construction, and operation of the High Temperature engineering Test Reactor (HTTR) and design of the commercial Gas Turbine High Temperature Reactor 300 (GTHTR300). The power output is 165 MWt and the inlet and outlet coolant temperatures are 325C and 750C, respectively, to provide steam for industrial utilization. However, given a requirement for the reactor pressure vessel to be smaller even that of the 30 MWt HTTR, several challenging technical problems have to be dealt with to arrive in a high performance core design that provides extended fuel burnup, prolonged refueling period, improved fuel refueling scheme, improved fuel element and so on from the HTTR.

Burn-up characteristics and criticality effect of impurities in the graphite structure of a commercial-scale prismatic HTGR

Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo

Nuclear Engineering and Design, 326, p.108 - 113, 2018/01

https://doi.org/10.1016/j.nucengdes.2017.11.003

Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.

Overview of Japanese development of accident tolerant FeCrAl-ODS fuel claddings for BWRs

Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 7 Pages, 2017/09

http://www.wrfpm2017.org/upload/draft_file/D_033-PD1.pdf

This paper will show the overview of current status of development of accident tolerant FeCrAl-ODS fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. This program is being carried out to create the technical basis for the practical use of the accident tolerant fuels and the other components in LWRs through multifaceted activities. In the development of FeCrAl-ODS fuel claddings both the experimental and the analytical studies have been performed. The acquisition and accumulation of key material properties of FeCrAl-ODS fuel claddings were conducted by using bar, sheet and tube shaped FeCrAl-ODS materials fabricated in this program to support the evaluations in the analytical studies. A neutron irradiation test was also started in the ORNL High Flux Isotope Reactor (HFIR) to examine the effect of neutron irradiation on the mechanical properties.

Multiple-wavelength neutron holography with pulsed neutrons

Hayashi, Koichi*; Oyama, Kenji*; Happo, Naohisa*; Matsushita, Tomohiro*; Hosokawa, Shinya*; Harada, Masahide; Inamura, Yasuhiro; Nitani, Hiroaki*; Shishido, Toetsu*; Yubuta, Kunio*

Science Advances (Internet), 3(8), p.e1700294_1 - e1700294_7, 2017/08

http://doi.org/10.1126/sciadv.1700294

Effect of thermo-mechanical treatments on nano-structure of 9Cr-ODS steel

Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Onuma, Masato*

Nuclear Materials and Energy (Internet), 9, p.346 - 352, 2016/12

https://doi.org/10.1016/j.nme.2016.10.007

Heat transport analysis in a district heating system applying waste heat from GTHTR300, a commercial design of high-temperature gas-cooled reactor

Kasahara, Seiji; Murata, Tetsuya*; Kamiji, Yu; Terada, Atsuhiko; Yan, X.; Inagaki, Yoshiyuki; Mori, Michitsugu*

Mechanical Engineering Journal (Internet), 3(3), p.15-00616_1 - 15-00616_16, 2016/06

http://doi.org/10.1299/mej.15-00616

A district heating system for household heating and road snow melting utilizing waste heat from GTHTR300, a heat-electricity cogeneration design of high temperature gas-cooled reactor, was analyzed. The application area was Sapporo and Ishikari, cities with heavy snowfall in northern Japan. The heat transport analyses were performed by modeling components to estimate heat supply profile; the secondary loops between the GTHTR300s and the heat-application area; heat exchangers connecting the secondary loops to the tertiary loops of the district-heating pipes; and the tertiary loops between the heat exchangers and houses and roads. Though double pipes for the secondary loops were advantageous for having less heat loss and a smaller excavation area, these advantages did not compensate for the higher construct cost of the pipes. To satisfy heat demand in the month of maximum requirement, 520-529 MW of heat were supplied by 3 GTHTR300s and delivered by 6 secondary loops, 3,450 heat exchangers about 90 m long, and 3,450 tertiary loops. Heat loss to the ground from the tertiary loops comprised 80%-90% of the heat loss. More than 90% of the construction cost went into thermal insulators. The thickness and properties of the thermal insulator must be reevaluated for economical heat delivery.

A Study on transmutation of LLFPs using various types of HTGRs

Kora, Kazuki*; Nakaya, Hiroyuki*; Matsuura, Hideaki*; Goto, Minoru; Nakagawa, Shigeaki; Shimakawa, Satoshi*

Nuclear Engineering and Design, 300, p.330 - 338, 2016/04

https://doi.org/10.1016/j.nucengdes.2016.01.030

In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system "GTHTR300", a small modular HTGR plant "HTR50S" and two types of plutonium burner HTGRs "Clean Burn with MA" and "Clean Burn without MA" were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

Developments of high-performance moderator vessel for JRR-3 cold neutron source

Arai, Masaji; Tamura, Itaru; Hazawa, Tomoya

JAEA-Technology 2015-010, 52 Pages, 2015/05

JAEA-Technology-2015-010.pdf:7.11MB

In the Department of Research Reactor and Tandem Accelerator, developments of high-performance CNS moderator vessel that can produce cold neutron intensity about two times higher compared to the existing vessel have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the design and manufacture of new vessel. In the present study, design strength evaluation, mockup test, simulation for thermo-fluid dynamics of the liquid hydrogen and strength evaluation of the different-material-bonding were studied. By these evaluation results, we verified that the developed new vessel can be applied to CNS moderator vessel of JRR-3.

Mechanical properties of weldments using irradiated stainless steel welded by the laser method for ITER blanket replacement

Yamada, Hirokazu*; Kawamura, Hiroshi; Nagao, Yoshiharu; Takada, Fumiki; Kono, Wataru*

Journal of Nuclear Materials, 355(1-3), p.119 - 123, 2006/09

https://doi.org/10.1016/j.jnucmat.2006.04.010

In this study, the bending properties of welding joint of irradiated material and un-irradiated material (irradiated/un-irradiated joints) were investigated using SS316LN-IG, which is the candidate material for the cooling pipe of ITER. The results of this study showed that the bending position of joints using un-irradiated material was un-irradiated part and that the bending position of irradiated/irradiated joints was fusion area or HAZ (heat affected zone). Although the bending position of joints was different bor the combination pattern between irradiated and un-irradiated materials, the bending strength of joint was almost same. Additionally, it is confirmed that bending strength did not depend on the combination pattern between the irradiated and un-irradiated materials, nor on the relationship between the heat input direction and the bending load direction.

Basic concept on structural design criteria for zirconia ceramics applying to nuclear components

Shibata, Taiju; Sumita, Junya; Baba, Shinichi; Yamaji, Masatoshi*; Ishihara, Masahiro; Iyoku, Tatsuo; Tsuji, Nobumasa*

Key Engineering Materials, 297-300, p.728 - 733, 2005/11

no abstracts in English