Survey of Research Activities for Passive Safety System of AC600

1999 ◽  
Author(s):  
Bingde Chen ◽  
Zhumao Yang ◽  
Fuyun Ji

Abstract The use of passive safety system in AC600, the Chinese advanced 600 MWe PWR proposed by NPIC, together with other improvements, such as simplification and advanced I&C etc., makes the plant more safe, economic and reliable. The core damage frequency (CDF) decreases from less than 10−4 of conventional PWR to less than 10−5 to 10−6 and the plant available factor increases to ∼90%. The passive safety system of AC600 consists of three complete independent systems. They are passive containment cooling system (passive CC system), passive core residual heat removal system (passive CRHR system) and passive safety injection system (CMT). To verify and demonstrate the AC600’s innovative passive safety features and to obtain an experimental database for system design modification and optimizing, and for computer code development and assessment, the experimental studies on these systems were finished in NPIC during the eighth national Five Year period under the national support. In this paper, the experimental research activities on passive containment cooling system, passive CRHR system and CMT injection system, including test rigs and main results are summarized. These experiments proved the design of all these passive systems are feasible and reliable and can meet basically the required safety functions. Some undesired thermal hydraulic phenomena, for example, “water hammer”, which may have bad impacts on its safety functions and to which high attention should be given, was found and identified in these studies. All data obtained have already been used in the design improvement and next R&D program planning.

Author(s):  
Jie Zou ◽  
Lili Tong ◽  
Xuewu Cao

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.


Author(s):  
Guohua Yan ◽  
Chen Ye

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.


Author(s):  
Mian Xing ◽  
Linsen Li ◽  
Feng Shen ◽  
Xiao Hu ◽  
Zhan Liu ◽  
...  

This paper gives a brief introduction of the Compact Small Reactor (CSR). It is a simplified two-loop reactor with thermal power of 660MW and with compact primary system and passive safety feature. Preliminary safety analysis of the CSR is conducted to evaluate and further optimize the design of passive safety system, especially the passive core cooling system. Large Break Loss Of Coolant Accident (LBLOCA) and Steam Generator Tube Rupture (SGTR) are selected as two reference accidental scenarios. Each scenario is modeled and computed by RELAP5/MOD3.4. For the LBLOCA analysis, a guillotine break happens in the cold leg of the loop containing the core makeup tanks balance lines. The results show certain safety margins from the guideline values, and the passive safety system could supply enough cooling of the core. For the SGTR analysis, the results show the robustness of the design from the safety perspective. It is concluded that the safety systems are capable of mitigating the accidents and protecting the reactor core from severe damage.


Author(s):  
Ye Cheng ◽  
Wang Minglu ◽  
Qiu Zhongming ◽  
Wang Yong

With the demand for nuclear power increasing, the first choice of almost all countries who want to build a new nuclear power plant is to use generation III technology, primarily because the safety of generation III technology is greatly improved compared with that of generation II and II + technology. The passive safety technology was introduced by the AP1000 and is one of the best applications of generation III technologies. In this study, the representative passive containment cooling system of the CAP1400 (developed based on AP1000) and the containment spray system of a generation II nuclear power plant are compared and analyzed using the Probabilistic Safety Assessment method. The reasons why a passive safety system has comparative advantages are determined by concrete calculations.


2014 ◽  
Vol 74 ◽  
pp. 136-142 ◽  
Author(s):  
J. Lim ◽  
J. Yang ◽  
S.W. Choi ◽  
D.Y. Lee ◽  
S. Rassame ◽  
...  

Author(s):  
Wei Li ◽  
Shuhong Du ◽  
Weiquan Gu ◽  
Nan Zhang ◽  
Ming Ding ◽  
...  

Abstract HPR1000 is an advanced nuclear power plant with the significant feature of an active and passive safety design philosophy, developed by the China National Nuclear Corporation. It is based on the large accumulated knowledge from the design, construction as well as operations experience of nuclear power plants in China. The passive containment cooling system (PCS) of HPR1000 is an important and innovative passive safety system to suppress the pressure in the containment during LOCA. In this paper, the detailed design process of PCS is reviewed, and an integrated experiment facility for the study on the coupling behavior between PCS and thermal hydraulic characteristics in the containment is described, and arrangement of measuring points including temperature, pressure, gas composition and so on are introduced in detailed. Also, the experimental energy released and energy vent to ensure the similarity of containment pressure response, thermal stratification and PCS heat removal is introduced. According to this versatile experiment facility can conduct real-engineering system test which is designed to support the PCS development. In addition, this valuable experience in the design and manufacture of integrated experiment facility can provide important technical support and guidance for the China next generation advanced PWR as well as safety related system.


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