Analyses for Passive Safety of Fusion Reactor during Ex-Vessel Loss of Coolant Accident

1995 ◽  
Vol 32 (4) ◽  
pp. 265-274 ◽  
Author(s):  
Takuro HONDA ◽  
Takashi OKAZAKI ◽  
Koichi MAKI ◽  
Tatuhiko UDA ◽  
Yasushi SEKI ◽  
...  
1996 ◽  
Vol 31 (3) ◽  
pp. 259-263 ◽  
Author(s):  
T. Hino ◽  
Y. Hirohata ◽  
T. Yamashina

Author(s):  
Pan Wu ◽  
Junli Gou ◽  
Jianqiang Shan ◽  
Bo Zhang ◽  
Xiang Li

This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.


2020 ◽  
Author(s):  
Andrea Zappatore ◽  
Antonio Froio ◽  
Gandolfo Alessandro Spagnuolo ◽  
Roberto Zanino

Author(s):  
Jun Liao ◽  
Vefa N. Kucukboyaci

Passive safety design that utilizes gravity, natural circulation, heat sink and stored potential energy for reactor safety functions is being increasingly adopted in advanced reactors, especially in the small modular reactor (SMR) designs. The passive safety design of the Westinghouse SMR is described in details and compared with the AP1000® passive safety design. The natural circulation loops and heat transfer mechanism in a postulated Westinghouse SMR loss of coolant accident (LOCA) are discussed. The key thermal hydraulic phenomena pertinent to the passive safety design of the Westinghouse SMR have been identified in the small break LOCA Phenomena Identification and Rank Table (PIRT). Among the identified phenomena, condensation on the containment wall and natural circulation in core makeup tank (CMT) loop are highly ranked. Those passive safety phenomena are expected to be assessed using the WCOBRA/TRAC-TF2 LOCA thermal hydraulic code, which will provide the design basis LOCA analysis in the SMR design control documentation. In this paper, the progress on the assessing two key phenomena in passive safety of Westinghouse SMR is reported. The preliminary assessments against UCB tube condensation tests and Westinghouse core makeup tank tests reveals the capability of WCOBRA/TRAC-TF2 code to reasonably predict the condensation on the containment wall and natural circulation in the core makeup tank (CMT) loop.


2021 ◽  
Vol 2108 (1) ◽  
pp. 012088
Author(s):  
Mengdi Dai ◽  
Xiaomo Wang

Abstract Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) is a kind of concept for the European demonstration fusion reactor (DEMO). The blanket attachment system plays an important role in the mechanical connection of the BB and vacuum vessel. Typically, the mechanical and thermal loads should meet the requirement to avoid collapse of the system with off-normal conditions, e.g., under ex-vessel Loss of Coolant Accident (LOCA. This paper investigates the loading requirement corresponding to the maximum stress that can sustain to avoid the LOCA condition. Firstly, a model of the BB is constructed using SolidWorks. Then, stress analysis is carried out based on the cross section of the blanket. Through simulation, the critical condition for the LOCA case and the maximum stress value for the model are obtained. According to the relevant size dimension from the reference, the blanket’s cross section is drawn, and one can get the stress field under the ex-vessel LOCA through stress analysis. The stress distribution under the ex-vessel LOCA condition is simulated to find out the maximum stress field that the blanket can sustain through this paper. The significance is to predict the possible conditions leading to an accident and find possible methods to avoid them.


Author(s):  
Wang Yuqi ◽  
Yu Aimin ◽  
Yang Qingming

This paper researched the behavior of 20mm break Loss of Coolant Accident (LOCA) which is located in the Direct Vessel Injection (DVI) line of the integrated small modular reactor (SMR) in case of full power with RELAP5-3KEYMASTER simulation system. The response of passive safety systems is analyzed and compared with the Primary Safety Analysis Report (PSAR) post-accident. Tendency for the variation of main parameters after the accident agree well with the PSAR, which validates the accuracy and rationality of the model, and solves the new problems in the process of modeling and provides an important tool for the research and development of SMR. Cooling and depressurization are calculated post-accident. The variation of main parameters post-accident and the accident advancement and results have been analyzed. Operation intervention is given and the effects with it are discussed. And the emergency strategy for development and verification of Emergency Operating Procedures (EOP) is given.


Author(s):  
Hwang Bae ◽  
Sung Uk Ryu ◽  
Hyo Bong Ryu ◽  
Woo Shik Kim ◽  
Sung-Jae Yi ◽  
...  

A passive injection test was conducted using a core makeup tank (CMT), a safety injection tank (SIT) and an automatic depressurization system (ADS), which consists of a passive safety system (PSS) of the SMART reactor. This paper investigates the thermal-hydraulic interaction between CMT and SIT during sequential injections of coolant from these two tanks to a high-temperature and high-pressure reactor pressure vessel using an integral effect test facility of SMART-ITL (System-Integrated Modular Advanced ReacTor-Integral Test Loop). Both CMT and SIT were connected to the reactor pressure vessel by a pressure balance line (PBL) and injection line (IL). A steady-state condition was maintained for 1,000 seconds before the start of the injection. The major parameters agreed well with the target value. After one of safety injection system line was simulated to be broken, a transient injection test was conducted according to the small-break loss-of-coolant accident (SBLOCA) scenario. Coolant injections from a CMT and SIT were started sequentially by opening quick-opening valves installed on the IL and PBL piping, respectively. Several thermal-hydraulic phenomena such as direct contact condensation, thermal stratification, and coupling effects between the CMT and SIT were locally observed during the SBLOCA scenario. The results show that the adopted passive safety injection system functions well as an emergency core cooling system.


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