Analysis and mapping of detailed inner information of crystalline grain by wavelength-resolved neutron transmission imaging with individual Bragg-dip profile-fitting analysis

Sakurai, Yosuke*; Sato, Hirotaka*; Adachi, Nozomu*; Morooka, Satoshi; Todaka, Yoshikazu*; Kamiyama, Takashi*

Applied Sciences (Internet), 11(11), p.5219_1 - 5219_17, 2021/06

https://doi.org/10.3390/app11115219

Microstructural stability of ODS steel after very long-term creep test

Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki

Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04

https://doi.org/10.1016/j.jnucmat.2021.152833

In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.

Evaluation of breach characteristics of fast reactor fuel pins during steady state irradiation

Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi

Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12

https://doi.org/10.1016/j.nucengdes.2020.110894

The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.

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

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

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

AA2015-0964.pdf:0.64MB

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.