Dynamic PRA Based on System Codes Coupling for Passive Safety System in Integral Pressurized Water Reactor

2020 ◽  
Author(s):  
Yi Mi ◽  
Akira Tokuhiro
Keyword(s):  
Natural Circulation ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Active Systems ◽  
Water Reactor ◽  
System Parameters ◽  
Event Ordering ◽  
Passive Safety Systems ◽  
The Difference

Abstract An integral Pressurized Water Reactor (iPWR) type SMR was studied featuring Passive Safety Systems (PSSs). Different from active systems, PSSs are easily influenced by system parameters referred to as phenomenological factors such as heat loss, flow friction, oxidation, non-condensable gases and void fraction due to the low driving force of natural circulation. The system parameters also contribute to the uncertainty and dependency of PSS leading to the system unreliability. Thus, efforts are made to improve the reliability of PSS. A classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering for PSS that dynamic PRA can accommodate. Here we developed and realized coupling between LabVIEW and CAFTA. Isolation Condenser System (ICS) is taken as the benchmark system due to the simple design in single phase without phase change phenomena in order to mainly remove decay heat and secondarily depressurize the reactor pressure vessel (RPV). A classical PRA model of ICS using CAFTA is coupled with real-time simulation of primary loop and ICS in LabVIEW, leading to a dynamic simulation result. The difference in failure probability using dynamic versus classical PRA revealed that for one there are more component demands with different event ordering, such that improved PSS reliability in the iPWR-type SMR designs is possible.

Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor

2014 ◽  
Author(s):  
Shinya Miyata ◽  
Satoru Kamohara ◽  
Wataru Sakuma ◽  
Hiroaki Nishi
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Core Cooling

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.

scholarly journals Analyses of the OSU-MASLWR Experimental Test Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-19 ◽  
Author(s):  
F. Mascari ◽  
G. Vella ◽  
B. G. Woods ◽  
F. D'Auria
Keyword(s):  
Natural Circulation ◽  
Developed Countries ◽  
System Level ◽  
Test Facility ◽  
State University ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Standard Problem ◽  
Oregon State University

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

Simulation Research on Thermal-Hydraulic Performance of a Natural Circulation Integrated Pressurized Water Reactor

2018 ◽  
Author(s):  
Yanan Zhao ◽  
Minjun Peng ◽  
Genglei Xia ◽  
Lianxin Lv
Keyword(s):  
Control Strategy ◽  
Natural Circulation ◽  
Dynamic Performance ◽  
Hydraulic Performance ◽  
Normal Operation ◽  
Outlet Temperature ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Operation Characteristics

As an effective measure to improve the reactor’s inherent safety feature, natural circulation is widely used in current integrated reactor design. The thermal-hydraulic performance of a flashing-driven nature circulation integrated pressurized water reactor (NC-IPWR) is studied, by taking IP100 reactor as reference. The simulation model of the reactor is established by RELAP5 code. A control system is designed based on the operation characteristics of the reactor. Both steady-state and dynamic performance of the reactor are analyzed and the rationality of the control strategy is verified in this work. The results demonstrate the operation characteristics of the IP100 reactor, and the dynamic performance of the reactor during power variation is discussed in detail. The control strategy that keeps the steam pressure and the core outlet temperature constant shows good performance under normal operation conditions. The obtained analysis results are significant for deeper understanding and improving the operation characteristics of the IP100 reactor.

Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4

2010 ◽  
Author(s):  
Xuhua Ye ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Parameter Study ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Pressure Peak ◽  
Residual Heat

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.

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