Validation of the hybrid turbulence model in detailed thermal-hydraulic analysis code SPIRAL for fuel assembly using sodium experiments data of 37-pin bundles

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Nuclear Technology, 210(5), p.814 - 835, 2024/05

https://doi.org/10.1080/00295450.2023.2249707

In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.

Elucidation of solid particle interfacial phenomena in liquid sodium; Magnetic interactions on liquid metal and solid atoms at the solid interface

Tei, C.; Otaka, Masahiko; Kuwahara, Daisuke*

Chemical Physics Letters, 829, p.140755_1 - 140755_6, 2023/10

https://doi.org/10.1016/j.cplett.2023.140755

We were able to detect the nuclear magnetic resonance (NMR) signal of a liquid sodium clinging to the interface of solid metal particles for the first time. In this study, we confirmed the difference in the relaxation times due to the difference in the interactions between liquid sodium and metal particles suspended in the liquid sodium. It was found that the surface of the micro titanium particles and liquid metallic sodium interact physically, not chemically.

Development of Lagrangian particle method for temperature distribution formed by sodium-water reaction in a tube bundle system

Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi; Yanagisawa, Hideki*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08

https://doi.org/10.13182/NURETH20-40116

The leakage of pressurized water from a steam generator (SG) and the progress after that are a key issue in the safety assessment or design of a SG in sodium-cooled fast reactor. The analysis code LEAP-III can evaluate a rate of water leakage during the long-term event progress, i.e., from the self-wastage initiated by an occurrence of a microscopic crack in a tube wall to the water leak detection and water/water-vapor blowdown. Since LEAP-III consists of semi-empirical formulae and one-dimensional equations of conservation, it has an advantage in short computation time. Thus, LEAP-III can facilitate the exploration of various new SG designs in the development of innovative reactors. However, there are several problems, such as an excessive conservative result in some case and the need for numerous experiments or preliminary analyses to determine tuning parameters of models in LEAP-III. Hence, we have developed a Lagrangian particle method code, which is characterized by a simpler computational principle and faster calculation. In this study, we have improved the existing particle pair search method for interparticle interaction in this code and developed an alternative model without the pair search. Through the trial analysis simulating in a tube bundle system, it was confirmed that new models reduced the computation time. In addition, it was shown that representative temperatures of the heat-transfer tubes evaluated by this particle method code, which is used to predict the tube failure in LEAP-III, were good agreement with that by SERAPHIM, which is a detailed mechanistic analysis method code.

Effectiveness evaluation methodology of the measures for improving resilience of nuclear structures against excessive earthquake

Kurisaka, Kenichi; Nishino, Hiroyuki; Yamano, Hidemasa

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05

The objective of this study is to develop an effectiveness evaluation methodology of the measures for improving resilience of nuclear structures against excessive earthquake by applying the failure mitigation technology. This study regarded those measures for improving resilience of important structures, systems, and components for safety to enlarge their seismic safety margin. To evaluate effectiveness of those measures, seismic core damage frequency (CDF) is selected as an index. Reduction of CDF as an effectiveness index is quantified by applying seismic PRA technology. Accident sequences leading to loss of decay heat removal are significant contributor to seismic CDF of sodium-cooled fast reactors (SFRs), and those sequences result in core damage via ultra-high temperature condition. This study improved the methodology to evaluate not only the measures against shaking due to excessive earthquake but also the measures at the ultra-high temperature condition. To examine applicability of the improved methodology, a trial calculation was implemented with some assumptions for a loop-type SFR. Within the assumption, the measures for improving resilience were significantly effective for decreasing CDF in excessive earthquake up to several times of a design basis ground motion. Through the applicability examination, the methodology for the effectiveness evaluation was developed successfully.

Development of dynamic PRA methodology for external hazards in sodium-cooled fast reactor via applying Markov chain Monte Carlo method to severe accident analysis code; Assessment of accident management of assigning independent emergency diesel generators to each air cooler

Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

Experimental study on prevention of high cycle thermal fatigue at the core outlet of advanced sodium-cooled fast reactor; Characteristics of temperature fluctuations and countermeasures to mitigate temperature fluctuations at a bottom of upper internal structure

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Nagasawa, Kazuyoshi*; Kurihara, Akikazu; Tanaka, Masaaki

JAEA-Research 2022-009, 125 Pages, 2023/01

JAEA-Research-2022-009.pdf:29.22MB

The design studies of an advanced loop-type sodium-cooled fast reactor (Advanced- SFR) have been carried out by the Japan Atomic Energy Agency (JAEA). At the core outlet, temperature fluctuations occur due to mixing of hot sodium from the fuel assembly with cold sodium from the control rod channels and radial blanket assembly. These temperature fluctuations may cause high cycle thermal fatigue around a bottom of Upper Internal Structure (UIS) located above the core. Therefore, we conducted a water experiment using a 1/3 scale 60 degree sector model that simulated the upper plenum of the advanced loop-type sodium-cooled reactor. And we proposed some countermeasures against large temperature fluctuations that occur at the bottom of the UIS. In this report, we have summarized that the effect of the countermeasure structure to mitigate the temperature fluctuation generated at the bottom of UIS is confirmed, and the Reynolds number dependency of the countermeasure structure and the characteristics of the temperature fluctuation on the control rod surface.

The Development of Petri Net-based continuous Markov Chain Monte Carlo methodology applying to dynamic probability risk assessment for multi-state resilience systems with repairable multi-component interdependency under longtermly thereat

Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi

Journal of Nuclear Science and Technology, 23 Pages, 2023/00

https://doi.org/10.1080/00223131.2023.2290067

A 3D particle-based simulation of heat and mass transfer behavior in the EAGLE ID1 in-pile test

Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Annals of Nuclear Energy, 179, p.109389_1 - 109389_10, 2022/12

https://doi.org/10.1016/j.anucene.2022.109389

Investigation on natural circulation behavior for decay heat removal in reactor vessel of sodium-cooled fast reactor under severe accident condition, 1; Effect of decay-heat conditions on natural circulation behavior under dipped-type DHX operation conditions

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 6 Pages, 2022/10

In sodium-cooled fast reactors (SFRs), decay heat removal after a core disruptive accident (CDA) is an important issue for the safety enhancement. Therefore, water experiments using a 1/10 scale experimental apparatus (PHEASANT) that simulates the reactor vessel of an SFR are conducted to investigate the natural circulation phenomena in the reactor vessel. In this study, experiments under the operation of the dipped-type DHX were conducted to investigate the effect of the heat generation ratio between the fuel debris on the core catcher in lower plenum and the reactor core remnant on the natural circulation behavior in the reactor vessel. The temperature distribution and the velocity distribution were measured under two heat generation conditions. Thus, the effect of the heat generation ratio between the fuel debris in the lower plenum and the reactor core remnant on the natural circulation behavior was quantitatively grasped under the dipped-type DHX operating conditions.

Development of safety design criteria and safety design guidelines for Generation IV sodium-cooled fast reactors

Futagami, Satoshi; Kubo, Shigenobu; Sofu, T.*; Ammirabile, L.*; Gauthe, P.*

Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10