scholarly journals RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

2020 ◽  
Vol 2020 ◽  
pp. 1-14
Author(s):  
Lin Sun ◽  
Xuesong Wang ◽  
Jun Wang ◽  
Meiru Liu ◽  
Genglei Xia
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Injection System ◽  
Passive Safety ◽  
Main Stream ◽  
Hydraulic Analysis ◽  
Pressurized Water ◽  
System Pressure ◽  
Thermal Hydraulic Analysis

Qinshan nuclear power plant is the first large Chinese-designed nuclear power station based on pressurized water reactor, and the second generation main stream active safety injection system is adopted for Qinshan nuclear power plant. In this paper, a novel passive safety injection system (PSIS) has been proposed for ocean-based Qinshan Phase One nuclear power plant to replace the original active one. The PSIS contains high-pressure, medium-pressure, and lower-pressure safety injection systems, and a two-stage automatic depressurization system. To evaluate the system performance, small-break LOCA has been investigated using RELAP5. Various break sizes and locations including 2-inch, 10-inch cold leg break, and double-ended direct vessel injection line break were analyzed. Key safety parameters such as safe injection mass flow rates, coolant level of the core, system pressure, and fuel cladding temperature were monitored during the accident process. The results illustrate that the performance of the safety injection system can guarantee the effective core cooling and submerged under different LOCA even with only half of the safety injection system, which can fulfill the single failure criteria. The thermal-hydraulic analysis for the Qinshan passive safety injection system is significant to master the related technologies for Chinese engineer and develop the Chinese-designed third-generation nuclear power plants, and the PSIS can guarantee the reactor submerged under LOCA even plus the station block out accident.

Scaling Application of a Hot Leg SBLOCA Scenario to a Nuclear Power Plant

2014 ◽  
Author(s):  
Andrea Querol ◽  
Sergio Gallardo ◽  
Gumersindo Verdú
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Experimental Conditions ◽  
Integral Test ◽  
Pressurized Water ◽  
Choked Flow

Several experimental facilities, such as the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA), have been built to reproduce some accidental scenarios because full-scale testing is usually impossible to perform. One of the objectives of these Integral Test Facilities (ITFs) is to obtain measured data to be compared to simulations in order to test the capability of the thermalhydraulic codes to reproduce experimental conditions. The applicability of these experimental results to a full-size power plant system depends on the scaling criteria adopted. The present paper is focused on the simulation and the scaling of the Test 1-2 in the frame of the OECD/NEA ROSA Project to a Nuclear Power Plant (NPP). This test simulates a hot leg 1% Small Break Loss-Of-Coolant Accident (SBLOCA) in a Pressurized Water Reactor (PWR) under the actuation of High Pressure Injection (HPI) system and Accumulator Injection System (AIS). A scaled-up NPP TRACE5 input has been developed from a LSTF TRACE5 model validated by authors in previous works. The scaled-up model has been developed conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. Lengths and diameters of hot legs have been scaled from LSTF model trying to conserve Froude number. A comparison between both TRACE5 models (LSTF and scaled-up NPP) is performed (system pressures, discharged inventory and collapsed liquid levels). Special TRACE5 models such as Choked flow model and OFFTAKE model have been tested. A 3D VESSEL component has been tested in comparison to 1D TEE component to simulate the hot leg where the SBLOCA is located and varying the break orientation (downwards and upwards). Finally, a sensitivity analysis has been made to determine the effect of the break size in the SBLOCA range.

Preliminary Study of Function Determination and Capacity of ACP600’s Reactor Emergency Borating System

2016 ◽  
Author(s):  
Shu Zhang ◽  
Xiaohua Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
System Capacity ◽  
Injection System ◽  
Third Generation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Preliminary Study ◽  
Generation Technology

ACP600 is a 600WMe power, self-reliance developed pressurized water reactor design of third generation technology, which is the China National Nuclear Corporation’s (CNNC) achievement of continuous R&D and engineering feedback started with M310 series and 600WMe designs. In general technical project of ACP600, there is a plan to cancel Boron Injection Tank (BIT) in safety injection system and set Reactor Emergency Borating (REB) system in order to enhance safety of nuclear power plant. This paper discusses function determination and system capacity of ACP600’s REB system based on the investigations of both international and domestic advanced third generation nuclear power plant emergency borating system’s safety functions and according to ACP600’s top design and general project along with operation and typical accident analysis, bringing forward to preliminary suggestions about function determination and system capacity of ACP600’s REB system as well as further aspect and content for investigations. This provides concrete information for further design of ACP600’s REB system.

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

Safety Study of Primary Loop System of Civil Marine Nuclear Power Plant

2014 ◽  
Vol 541-542 ◽  
pp. 916-921 ◽  
Author(s):  
Li Xu ◽  
Ru Chao Deng ◽  
Chu Xu ◽  
Di Zhang ◽  
Chen Xing Sheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Comprehensive Evaluation ◽  
Environmental Parameters ◽  
Loop System ◽  
Injection System ◽  
Safety Study ◽  
Primary Loop ◽  
Control Rods

For evaluate the risk of civil marine nuclear power plant, through the searching related standards for ship, external environmental parameters that the nuclear ship should be suited was found. Based on the characteristics of power plant of civil nuclear-powered ship, the hierarchy system of primary loop system was established and corresponding indicator marking criteria were formulated for the risk assessment. The result shows that the Reactor Safety Injection System (RIS), the Reactor Boron and the Water Supply System (REA), the Control Rods and the Hull of Fuel Canning are the key risk factors in the primary loop system. Finally, the comprehensive evaluation was carried out for collision, stranding and swing of multi-degree of freedom, and put forward relative countermeasures to cope with the possible risks based on the comprehensive evaluation and combined with the literatures.

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