Internal event level-1 PRA for sodium-cooled fast reactor considering safety measures of defense-in-depth level 1 to 4

Nishino, Hiroyuki; Kurisaka, Kenichi; Naruto, Kenichi*; Gondai, Yoji; Yamamoto, Masaya

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

The effectiveness evaluation of safety measures against severe accident is necessary for restart of experimental sodium-cooled fast reactor "Joyo" in Japan. These safety measures correspond to those in defense-in-depth (DiD) level 4. In the previous study, a level-1 probabilistic risk assessment (PRA) at power was performed to calculate frequencies of the accident sequences of failure of safety measures in DiD level 1 to 3, to identify dominant accident sequence groups, and to identify dominant accident sequence for selecting important accident sequences in each accident sequence group which are needed for implementing the effectiveness evaluation of safety measures in DiD level 4. Based on this, the present study implemented level-1 PRA at power to show quantitatively reduction of those occurrence frequency by the safety measure in the DiD level 4. As the result, the frequency of each accident sequence group decreased significantly, and total frequency of the accident sequence groups decreased to about 110 /reactor-year which is about 1/1000 times the one estimated in the previous study. The protected loss of heat sink was the largest contributor in all the accident groups and a dominant accident sequence in each accident group was also identified in this study.

Internal event level-1 PRA for sodium-cooled fast reactor considering safety measures of defense-in-depth level 1 to 3

Nishino, Hiroyuki; Kurisaka, Kenichi; Naruto, Kenichi*; Gondai, Yoji; Yamamoto, Masaya; Yamano, Hidemasa

Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 12 Pages, 2020/11

The objective of this study is to evaluate the occurrence frequency of accident sequences which may lead to core damage if provisions in defense in depth (DiD) level 1 to 3 are the only safety measures. For this objective, the existing safety measures in this SFR are categorized into those for the DiD level 1-3 and those for the DiD level 4. The safety measures for the DiD level 1-3 are as follows; (1) main reactor shutdown system, (2) double boundary structure in the primary main and auxiliary cooling system and the reactor vessel, which maintain the reactor coolant level sufficient for coolant circulation in the primary main cooling system, (3) decay heat removal in a forced circulation mode. Accident sequences are categorized into typical SFR-specific groups and station blackout (SBO) in this study. The SFR-specific groups are unprotected loss of flow, unprotected transient over power, unprotected loss of heat sink, loss of reactor level, and protected loss of heat sink (PLOHS). The occurrence frequency of these accident sequence groups was quantified to identify major contributors. As the result, PLOHS excluding SBO was indicated as the dominant contribution of 80% or more in the all accident sequence groups and the annual occurrence frequency of the PLOHS was 1.0E-4 order of magnitude. For the PLOHS, loss of offsite power (LOOP) was indicated as major contribution of 30% in initiating events. In the accident sequences of the PLOHS initiated from LOOP, a dominant sequence was combination of common cause failure of primary pumps in the main cooling system and failure-to-start of the auxiliary cooling system after LOOP. The second dominant contribution (15% or more) in the all accident sequence groups is PLOHS in SBO (i.e., decay heat removal failure due to SBO). Each of the other accident sequence groups was 1%.

An Experimental study on self-leveling behavior of debris beds with comparatively higher gas velocities

Cheng, S.; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Gondai, Yoji*; Nakamura, Yuya*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12

Experimental studies and empirical models for the transient self-leveling behavior in debris bed

Cheng, S.*; Tanaka, Yohei*; Gondai, Yoji*; Kai, Takayuki*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*; Fukuda, Kenji*; Yamano, Hidemasa; Suzuki, Toru; et al.

Journal of Nuclear Science and Technology, 48(10), p.1327 - 1336, 2011/10

https://doi.org/10.3327/jnst.48.1327

To clarify this behavior, a series of experiments have been performed in which nitrogen gas has been percolated uniformly through a particle bed. In these experiments, solid particles and water contained in a rectangular tank simulate respectively fuel debris and coolant. Based on the data obtained, an empirical model was developed to describe the transient variation in the bed inclination angle during the self-leveling process. Good agreement has been obtained between calculated and experimental values. Verification of the model has been confirmed through detailed analysis of the effects of experimental parameters such as particle size, particle density, and gas flow rate. Its applicability to extended conditions was further discussed by performing modeling simulations and comparing results against experimental data obtained from a larger-scale experimental system that employed a conventional boiling method.

Experimental investigation on self-leveling behavior in debris bed

Cheng, S.*; Tanaka, Yohei*; Gondai, Yoji*; Kai, Takayuki*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*; Fukuda, Kenji*; Yamano, Hidemasa; Suzuki, Toru; et al.

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11

Study on characteristics of particle-bed self-leveling driven by gas injection

Nakamura, Yuya*; Gondai, Yoji*; Cheng, S.*; Takeda, Shohei*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*; Yamano, Hidemasa; Tagami, Hirotaka; Suzuki, Toru; et al.

no journal, , 

In order to clarify the characteristics of debris bed behavior in the post accident heat removal phase in the core disruptive accidents of FBR, an experimental study to simulate the coolant boiling in debris bed by gas injection from the bottom of the bed and basic knowledge on the self-leveling characteristic for large vapor velocity was obtained.