Evaluation of thermal neutron scattering law of nuclear-grade isotropic graphite

Nakayama, Shinsuke; Iwamoto, Osamu; Kimura, Atsushi

EPJ Web of Conferences, 294, p.07001_1 - 07001_6, 2024/04

https://doi.org/10.1051/epjconf/202429407001

Graphite is a candidate of moderator in innovative nuclear reactors such as molten salt reactors. Scattering of thermal neutrons by the moderator material has a significant impact on the reactor core design. To contribute to the development of innovative nuclear reactors, an evaluation method of thermal neutron scattering law for reactor grade graphite was studied. The inelastic scattering component due to lattice vibration was evaluated based on the phonon density of states computed with first-principles lattice dynamics simulations. The simulations were performed for ideal crystalline graphite. The coherent elastic scattering component due to crystal structure was evaluated based on neutron transmission and scattering experiments recently performed in the J-PARC/MLF facility. In comparison with the neutron transmission experiments, it was found that the quantification of small-angle neutron scattering due to structures larger than crystal, such as pores in graphite, is important. Based on the above methods, thermal neutron scattering law data for reactor-grade graphite at room temperature were evaluated.

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.

Numerical simulation technologies for safety evaluation in plant lifecycle optimization method, ARKADIA for advanced reactors

Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.

Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11

https://doi.org/10.1016/j.nucengdes.2023.112492

The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.

Structural analysis of a reactor vessel in a sodium-cooled fast reactor under extremely high temperature conditions

Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori

Mechanical Engineering Journal (Internet), 10(4), p.23-00043_1 - 23-00043_12, 2023/08

https://doi.org/10.1299/mej.23-00043

This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.

Development of stress intensity factor solution for surface crack at nozzle corner in reactor pressure vessel

Yamaguchi, Yoshihito; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.

Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 9 Pages, 2023/07

The stress intensity factor (SIF) for crack at nozzle corner is a key parameter in structural integrity assessment of nozzle in reactor pressure vessel (RPV). Although various SIF solutions for surface cracks at nozzle corners have been proposed, most of them are only focusing on the deepest point of the crack, and the information about geometric dimension of the nozzle corner is not clear. According to the previous fatigue test results regarding the surface crack at the nozzle corner, the amounts of crack growth at the surface points were larger than that at the deepest point of the crack. Such results imply that SIFs at the surface points may be higher than that at the deepest point. To increase the reliability of the structural integrity assessment, it is necessary to provide SIF solutions for both surface and deepest points. In this study, SIF solutions for two surface points and the deepest point of surface crack at nozzle corners are developed through finite element analyses and the solutions are provided corresponding to the geometric dimensions of nozzle corner and crack size.

Improvement of cooling performance of reactor pressure vessel using passive cooling

Banno, Masaki*; Funatani, Shumpei*; Takamatsu, Kuniyoshi

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05

A fundamental study on the safety of a passive cooling system for the RPV with radiative cooling is conducted. The object of this study is to demonstrate that passive RPV cooling system with radiative cooling is extremely safe and reliable even in the event of natural disasters. Therefore, an experimental apparatus, which is about 1/20 scale of the actual cooling system, was fabricated with several stainless steel containers. The surface of the heating element in the experimental apparatus simulates the surface of the RPV, and the heating element generates natural convection and radiation. A comparison of the Grashof number between the actual cooling system and the experimental apparatus confirmed that both were turbulent, and the experimental results as a scale model are valuable. Moreover, the experimental results confirmed that the heat generated from the surface of the RPV during the rated operation can be removed.

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.5 for reactor pressure vessels

Takamizawa, Hisashi; Lu, K.; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.

JAEA-Data/Code 2022-006, 221 Pages, 2023/02

JAEA-Data-Code-2022-006.pdf:4.79MB

As a part of the structural integrity assessment research for aging light water reactor (LWR) components, a probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in Japan Atomic Energy Agency. The PASCAL code can evaluate failure probabilities and failure frequencies of core region in reactor pressure vessel (RPV) under transients by considering the uncertainties of influential parameters. The continuous development of the code aims to improve the reliability by introducing the analysis methodologies and functions base on the state-of-the-art knowledge in fracture mechanics and domestic data. In the first version of PASCAL, which was released in FY2000, the basic framework was developed for analyzing failure probabilities considering pressurized thermal shock events for RPVs in pressurized water reactors (PWRs). In PASCAL Ver. 2 released in FY 2006, analysis functions including the evaluation methods for embedded cracks and crack detection probability models for inspection were introduced. In PASCAL Ver. 3 released in FY 2010, functions considering weld-overlay cladding on the inner surface of RPV were introduced. In PASCAL Ver. 4 released in FY 2017, we improved several functions such as the stress intensity factor solutions, probabilistic fracture toughness evaluation models, and confidence level evaluation function by considering epistemic and aleatory uncertainties related to influential parameters. In addition, the probabilistic calculation method was also improved to speed up the failure probability calculations. To strengthen the practical applications of PFM methodology in Japan, PASCAL code has been improved since FY 2018 to enable PFM analyses of RPVs subjected to a broad range of transients corresponding to both PWRs and boiling water reactors, including pressurized thermal shock, low-temperature over pressure, and normal operational transients. In particular, the stress intensi

Issues and proposals for activation calculation of reactor decommissioning using recent evaluated nuclear data library

Okumura, Keisuke

Kaku Deta Nyusu (Internet), (133), p.63 - 67, 2022/10

http://www.aesj.or.jp/~ndd/ndnews/No133.html

no abstracts in English

Common evaluation procedure radioactivity concentration by theoretical calculation for radioactive waste generated from the decommissioning of research reactors

Okada, Shota; Murakami, Masashi; Kochiyama, Mami; Izumo, Sari; Sakai, Akihiro

JAEA-Testing 2022-002, 66 Pages, 2022/08

JAEA-Testing-2022-002.pdf:2.46MB

Japan Atomic Energy Agency is an implementing organization of burial disposal for low-level radioactive waste generated from research, industrial and medical facilities in Japan. Radioactivity concentrations of the waste are essential information for design of the disposal facility and for licensing process. A lot of the waste subjected to the burial disposal is arising from dismantling of nuclear facilities. Radioactive Wastes Disposal enter has therefore discussed a procedure to evaluate the radioactivity concentrations by theoretical calculation for waste arising from the dismantling of the research reactors facilities and summarized the common procedure. The procedure includes evaluation of radioactive inventory by activation calculation, validation of the calculation results, and determination of the disposal classification as well as organization of the data on total radioactivity and maximum radioactivity concentration for each classification. For the evaluation of radioactive inventory, neutron flux and energy spectra are calculated at each region in the reactor facility using two- or three-dimensional neutron transport code. The activation calculation is then conducted for 140 nuclides using the results of neutron transport calculation and an activation calculation code. The recommended codes in this report for neutron transport calculation are two-dimensional discrete ordinate code DORT, three-dimensional discrete ordinate code TORT, or Monte Carlo codes MCNP and PHITS, and for activation calculation is ORIGEN-S. Other recommendation of cross-section libraries and calculation conditions are also indicated in this report. In the course of the establishment of the procedure, Radioactive Wastes Disposal Center has discussed the commonly available procedure at meetings. It has periodically held to exchange information with external operators which have research reactor facilities. The procedure will properly be reviewed and be revised by reflecting future situ

Development of ARKADIA for the innovation of advanced nuclear reactor design process (Overview of optimization process development in design optimization support tool, ARKADIA-Design)

Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07

https://doi.org/10.1299/jsmepes.2022.26.A221

To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.