An integral effect test of a complete loss of reactor coolant system flow rate for the SMART design using the VISTA-ITL facility and its simulation with the MARS-KS code

Pressurizer is one of the most important components in reactor coolant system of a nuclear power plant, which operates normally at pressure of 15.4 MPa and temperature of 345°C[1]. The main function of pressurizer is to regulate the pressure in the reactor coolant system by either cooling the steam or heating the saturated water in its upper zone. When the pressure in the reactor coolant system increases, it will distribute cold water to decrease its temperature and pressure through atomizing the reactor coolant with swirl spray nozzle in pressurizer. Swirl nozzle is the key part of pressurizer with swirl structure of full cone spray pattern, and the atomization performance include drop size, spray angle and distribution, also it is characterized by huge flow rate and low pressure drop, and its atomization performance decides the quality of pressure control of the reactor coolant system. To enhance the independent design level of both pressurizer and cooling system, it’s necessary to study the atomization performance of swirl nozzle for nuclear reactor pressurizer. Aimed at improving atomization performance of swirl spray nozzle, the structure design methodology of nuclear reactor pressurizer was studied systematically in three aspects including theory design, numerical simulation and test confirm in this thesis. Through designing the swirl nozzle structure according to similar design formula of spray nozzle in theory, especially studying the influence of different structures that mainly include internal swirl structure on internal flow field of swirl nozzles, the primary structure parameters of swirl nozzle were confirmed. Then, through numerical simulation of the internal flow field, flow rate and pressure drop, and swirl core structure of the swirl nozzle (by building physical model and mathematic model according to the spray nozzle structure), the atomization performance of the nozzle was analyzed. On this basis, the typical swirl nozzle was designed and tested, which included spray angle, flow rate as well as pressure drop tests, and spray drop tests, and the applicability of the computational fluid dynamics (CFD) method was verified when it was applied in swirl nozzle design. Finally, the design method of swirl nozzle with deep groove of swirl core for pressurizer was put forward. Through this studying of theoretical calculation, numerical simulating and test, the correlation between the structural parameters of swirl nozzle and atomization performance was achieved, meanwhile design, analysis and test methods of spray nozzle with low pressure drop and huge flow rate were established. It is helpful to realize the independent design of pressurizer’s swirl nozzle and even to put forward the design methodology of pressurizer’s swirl nozzle with our own characteristic.

Download Full-text

Integral Effect Test for Performance Evaluation of the PAFS (Passive Auxiliary Feedwater System) During a SBO (Station Blackout) Transient

Volume 2B: Thermal Hydraulics ◽

10.1115/icone22-30563 ◽

2014 ◽

Author(s):

Byoung-Uhn Bae ◽

Seok Kim ◽

Yu-Sun Park ◽

Yun-Je Cho ◽

Kyoung-Ho Kang

Keyword(s):

Steam Generator ◽

Flow Rate ◽

Water Levels ◽

Experimental Result ◽

Test Facility ◽

Decay Heat ◽

The Core ◽

Integral Effect ◽

Station Blackout ◽

Effect Test

Station blackout (SBO) accident is considered as one of the most significant design extension conditions (DECs), which has been extensively focused after the Fukushima Dai-chi accident. When the SBO accident occurs in the APR+ (Advance Power Reactor Plus), the PAFS (Passive Auxiliary Feedwater System), which is an advanced safety feature adopted in the APR+, should play a significant role to cool down the core decay heat without any operation of active safety systems. This study focuses on validation of the cooling and operational performance for the PAFS during the SBO transient with utilizing an integral effect test facility, ATLAS-PAFS. In order to simulate the SBO transient of the APR+ as realistically as possible, a pertinent scaling approach was taken into account. The initial steady-state conditions and the sequence of event in the SBO scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. In the transient simulation, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate at the return-water line were measured and investigated. Following the reactor trip at the initiation of the transient, the coolant inventory of the secondary system of the steam generator was reduced by the repeated opening and closing of the MSSV. When the collapsed water level reached 25% of wide range, the PAFS was actuated to cool down the primary system by the condensation heat transfer at the PCHX (Passive Condensation Heat Exchanger). The pressure and the temperature of the reactor coolant system continuously decreased during the heat removal by the PAFS operation. It points out that the PAFS can supply auxiliary feedwater to the steam generator and remove the core decay heat without any active system. From the present experimental result, it could be concluded that the APR+ has the capability of coping with the hypothetical SBO scenario with adopting the PAFS and proper set-points of its operation. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS, RELAP5 as well as the SPACE code and to identify any code deficiency for a SBO simulation with an operation of the PAFS.

Download Full-text

RETRAN Analysis of SONGS 2 Reactor Coolant System Flow Measurements

Anticipated and Abnormal Plant Transients in Light Water Reactors ◽

10.1007/978-1-4684-4799-6_51 ◽

1984 ◽

pp. 515-526

Author(s):

Yine-Ping Ting

Keyword(s):

Flow Measurements ◽

Reactor Coolant ◽

System Flow ◽

Coolant System

Download Full-text

Characteristics Analysis of SCWR During Partial Loss of Reactor Coolant Flow Transient

Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems ◽

10.1115/icone20-power2012-54788 ◽

2012 ◽

Author(s):

Juan Chen ◽

Tao Zhou ◽

Zhousen Hou ◽

Canhui Sun

Keyword(s):

Flow Rate ◽

Transient Analysis ◽

Coolant Flow ◽

Partial Loss ◽

Initial Flow ◽

Reactor Coolant ◽

Loss Of Coolant ◽

Characteristics Analysis ◽

And Control ◽

Supercritical Water Cooled Reactor

Partial loss of reactor coolant flow is one of the most important transients for safety analysis of supercritical water-cooled reactor (SCWR). Taking the super LWR concept provided by Japan as research object, transient analysis of partial loss of coolant flow rate is given by coupled neutronics and thermal hydraulics calculation method. The results show that, when 5% partial loss of coolant flow is happening, maximum cladding temperature would increase firstly with the decreasing of fuel channel inlet coolant flow. Then followed with the neutronic feedback and control operation, maximum cladding temperature decreases and finally return to normal. When 50% partial loss of coolant flow is happening, a scram signal will be given to ensure system safety, but the maximum cladding temperature still shows a significant increase early. On this basis, sensitivity analysis is performed considering the influence of core power and main coolant flow. It is found that maximum peaking value increases significantly following the initial flow rate decreasing, but shows a very little increase caused by core power increasing.

Download Full-text