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.

Collection of strength characteristic data used for analysis evaluation in reactor pressure vessel and in-core structures in accident

Shimomura, Kenta; Yamashita, Takuya; Nagae, Yuji

JAEA-Data/Code 2022-012, 270 Pages, 2023/03

JAEA-Data-Code-2022-012.pdf:38.25MB

In a light water reactor, which is a commercial nuclear power plant, a severe accident such as loss of cooling function in the reactor pressure vessel (RPV) and exposure of fuel rods due to a drop in the water level in the reactor can occur when a trouble like loss of all AC power occurs. In the event of such a severe accident, the RPV may be damaged due to in-vessel conditions (temperature, molten materials, etc.) and leakage of radioactive materials from the reactor may occur. Verification and estimation of the process of RPV damage, molten fuel debris spillage and expansion, etc. during accident progression will provide important information for decommissioning work. Possible causes of RPV failure include failure due to loads and restraints applied to the RPV substructure (mechanical failure), failure due to the current eutectic state of low-melting metals and high-melting oxides with the RPV bottom members (failure due to inter-material reactions), and failure near the melting point of the structural members at the RPV bottom. Among the failure factors, mechanical failure is verified by numerical analysis (thermal hydraulics and structural analysis). When conducting such a numerical analysis, the heat transfer properties (thermal conductivity, specific heat, density) and material properties (thermal conductivity, Young's modulus, Poisson's ratio, tensile, creep) of the materials (zirconium, boron carbide, stainless steel, nickel-based alloy, low alloy steel, etc.) constituting the RPV and in-core structures to near the melting point are required to evaluate the creep failure of the RPV. In this document, we compiled data on the properties of base materials up to the melting point of each material constituting the RPV and in-core structures, based on published literature. In addition, because welds exist in the RPV and in-core structures, the data on welds are also included in this report, although they are limited.

Constraint effect on fracture mechanics evaluation for an under-clad crack in a reactor pressure vessel steel

Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08

https://doi.org/10.1115/PVP2020-21441

In JEAC 4206 which prescribes the methodology for assessing the structural integrity of reactor pressure vessels (RPVs), an under-clad crack (UCC) at the inner surface of RPV is postulated, and it is required that the fracture toughness of RPV steels is higher than stress intensity factor for at the crack tip during the pressurized thermal shock event. In the present study, to investigate the effect of cladding on the fracture toughness, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for an RPV steel containing an UCC or a surface crack, and the constraint effect for UCC was also discussed. As the result, we found that the fracture toughness for UCC was considerably higher than that for surface crack. On the other hand, the FEAs showed that the cladding decreased the constraint effect for UCC.

A Rapid evaluation method of the heat removed by a VCS before rise-to-power tests

Takamatsu, Kuniyoshi

Journal of Thermal Science, 24(3), p.295 - 301, 2015/06

https://doi.org/10.1007/s11630-015-0787-x

Before rise-to-power tests, the actual measured value of heat released from the Reactor Pressure Vessel (RPV) or removed by the Vessel Cooling System (VCS) cannot be obtained. It is difficult for operators to evaluate the reactor outlet coolant temperature supplied from the High Temperature Engineering Test Reactor (HTTR) before rise-to-power tests. Therefore, when the actual measured value of heat released from the RPV or removed by the VCS are changed during rise-to-power tests, operators need to evaluate quickly, within a few minutes, the heat removed by the VCS and the reactor outlet coolant temperature of 30 (MW), at the 100% of the reactor power, before the temperature achieves to 967 (C) which is the maximum temperature limit generating the reactor scram. In this paper, a rapid evaluation method for use by operators is presented.

Post-irradiation annealing and re-irradiation technique for LWR reactor pressure vessel material

Matsui, Yoshinori; Ide, Hiroshi; Itabashi, Yukio; Kikuchi, Taiji; Ishikawa, Kazuyoshi; Abe, Shinichi; Inoue, Shuichi; Shimizu, Michio; Iwamatsu, Shigemi; Watanabe, Naoki*; et al.

KAERI/GP-195/2002, p.33 - 40, 2002/00

Studies on the irradiation damage of the material of the RPV are inevitable for the LWR. Recently, the researches of annealing effect on the irradiation damage of RPV material were extensively carried out using specimens irradiated in the JMTR of the JAERI. As the next step, an annealing test of irradiated specimens and re-irradiation of annealed specimens were planned. The aim of the test is to evaluate the effect of annealing by comparing the damage of irradiated specimen, its recovery by annealing and the damage after re-irradiation. For the re-irradiation test of this study, JAERI developed a new capsule in which the specimens can be exchanged before and after annealing, and, re-irradiated afterward. The development of the capsule consisted of the design and fabrication of airtight connector for thermocouples and mechanical seal device which was fit to remote handling. Remote operation procedures for handling the radioactive capsule and for exchanging specimens were carefully performed. The results of the re-irradiation proved that the development was technically successful.

Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water

Maruyama, Yu; ; Moriyama, Kiyofumi; H.S.Park*; Kudo, Tamotsu; Y.Yang*; Sugimoto, Jun

NEA/CSNI/R(98)18, p.243 - 250, 1999/02

no abstracts in English

Evaluation of Life Predictive for Light Water Reactor Pressure Vessel and its Relevant Theme for the Material Testing

; Nakajima, Hajime; Kondo, Tatsuo

JAERI-M 84-208, 25 Pages, 1984/11

JAERI-M-84-208.pdf:0.95MB

no abstracts in English

Development and systematic application of surface inspection methods used for in-service inspection of reactor pressure vessel

;

Journal of Nuclear Science and Technology, 16(10), p.750 - 763, 1979/00

https://doi.org/10.3327/jnst.16.750

no abstracts in English

Estimation of the in-depth debris status of Fukushima Unit-2 and Unit-3 with multi-physics modeling, 4; Consideration on possible RPV-boundary failure modes of Unit 2

Sato, Ikken; Yamaji, Akifumi*; Furuya, Masahiro*; Oishi, Yuji*; Li, X.*; Madokoro, Hiroshi; Fukai, Hirofumi*

no journal, , 

no abstracts in English