Low-Power and Shut-Down Condition Medium-Break Loss-of-Coolant Accident Success Criterion Analysis for a Typical Three-Loop Nuclear Power Plant

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Jing Sun ◽  
Changjiang Yang
Keyword(s):  
Low Power ◽  
Nuclear Power ◽  
Power Plants ◽  
Time Window ◽  
Injection System ◽  
Success Criterion ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Action Time ◽  
Operator Action

Safety equipment demands that the success criterion of useful equipment, operator-action time window, and the damage state of the reactor core can be defined by thermal-hydraulic (T-H) analysis, which makes a basic critical contribution to probabilistic safety assessment (PSA). PSA has been widely used in the safety evaluation and assessment of nuclear power plants (NPPs). A loss-of-coolant accident (LOCA) cannot be controlled without timely safety intervention. Low-power and shut-down (LPSD) conditions of NPPs can be divided into several plant operating states (POSs) in PSA analysis. After the Fukushima nuclear accident, the topic of station black-out (SBO) has drawn widespread concern. However, some LPSD conditions, which result in severe consequences like SBO, have not drawn widespread attention and are thus analyzed and discussed herein. This paper analyzes a medium-break LOCA (MBLOCA) under a certain LPSD condition for a typical three-loop NPP. A simplified method of simulating and selecting operator-action time of MBLOCA for PSA is developed. The proposed method calculates the time windows for both manually opening the high head safety injection system (HHSI) and secondary depressurizing of the system to keep the core undamaged, which could support building PSA model and human reliability analysis.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

scholarly journals SUBSTANTIATION OF MODERNIZED BLACKOUT & LOSS-OF-COOLANT ACCIDENT MANAGEMENT STRATEGY AT NUCLEAR POWER PLANTS WITH WWER

2020 ◽  
Vol 2 (61) ◽  
pp. 70-77
Author(s):  
V. Skalozubov ◽  
◽  
V. Spinov ◽  
D. Spinov ◽  
Т. Gablaya ◽  
...  
Keyword(s):  
Management Strategy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Management Strategies ◽  
Computational Modelling ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Safety Systems

The analysis of the known results of RELAP5/V.3.2 simulation for loss of coolant & blackout accidents at WWER nuclear power plants showed that the design accident management strategies with design passive safety systems do not provide the necessary safety conditions for the maximum permissible temperature of fuel claddings, the minimum permissible level of coolant in the reactor and feed water in the steam generators. A conservative thermohydrodynamic model for a design and modernized blackout & loss-of-coolant accident management strategy at a nuclear power plant with WWER has been developed. Design passive safety systems carry out the design accident management strategy: pressurizer safety valves, secondary steam relief valves, and hydraulic reservoirs of the emergency core cooling system of the reactor. Promising afterheat removal passive systems and the reactor level and steam generator water level control systems carry out the modernized blackout & loss-of-coolant accident management strategy. The main conservative assumptions of the presented model of blackout & loss-of-coolant accidents: complete long-term failure of all electric pumps of active safety systems, the temperature of nuclear fuel in the central part of the fuel matrix is assumed as the maximum allowable one, effect of “run down” flow of a turbine feed pump and the coolant level in pressurizer on accident process is not considered. Computational modelling has found that violations of the safety conditions are over the entire range of leak sizes for the design blackout & loss-of-coolant accident management strategy. For the modernized blackout & loss-of-coolant accident management strategy, safety conditions are provided for 72 hours of the accident and more. The presented results of computational modelling of blackout accident management strategies for nuclear power plants can be used to modernize and improve symptom-informed emergency instructions and guidelines for the severe accident management at nuclear power plants with WWER. Application of the results of computational modelling of blackout accident management strategies is generally not substantiated for other types of reactor facilities. In this case, it is necessary to develop calculated models for blackout accident management taking into account the specifics of the structural and technical characteristics and operating conditions for safety related systems of nuclear power plants.

THE VALIDATION OF COMPREHENSIVE SOFTWARE FOR THE ASSESSMENT OF RADIOLOGICAL IMPACTS ON HUMANS THROUGH BENCHMARK ANALYSIS WITH RADIOLOGICAL CODES

2020 ◽  
Vol 190 (3) ◽  
pp. 250-268
Author(s):  
Ali Haghighi Shad ◽  
Mitra Athari Allaf ◽  
Darioush Masti ◽  
Kamran Sepanloo ◽  
Seyed Amir Hossein Feghhi
Keyword(s):  
Thyroid Gland ◽  
Effective Dose ◽  
Nuclear Power ◽  
Power Plants ◽  
Whole Body ◽  
The Public ◽  
Loss Of Coolant Accident ◽  
Initial Stage ◽  
Loss Of Coolant ◽  
Accident Conditions

Abstract In this paper, a novel domestic code called KIANA was developed for the assessment of radiological impacts on the population in normal and accident conditions including design basis accident (DBA) and beyond DBA (BDBA) for the nuclear power plants. The validation process of the KIANA code was performed using the results of the DOZA_M radiological code, whose results are presented in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant Unit One (BNPP-1). The calculations of KIANA are performed based on the Gaussian diffusion model. The developed KIANA code has the potential of calculating the concentration and radionuclide doses due to the pathways such as airborne, foodstuff, marine (both one and two boxes models), soils, animals, vegetation (with and without tritium) and other pathways without any restriction. In the current research, the individual dose from a cloud to the member of the public in the region of BNPP-1 in normal condition was calculated. Moreover, the total effective dose to the member of the public from the primary to the secondary leakage inside steam generators, large break loss-of-coolant accident (LBLOCA) and small break loss-of-coolant accident was calculated. Thyroid gland equivalent dose for the infant (1–8 years) in the case of LBLOCA at the BNPP in DBA conditions was also evaluated. Finally, the prevented dose at the initial stage for the whole body of adults after BDBA, prevented dose at the initial stage for the thyroid gland of children after BDBA and the effective dose during the first year after the accident (external body irradiation from presence in the area) in the case of BDBA are assessed. The KIANA simulation results showed a good agreement with the FSAR data of BNPP.

Application of the UMAE Methodology for the Uncertainty Evaluation in Support to the EOP Background Analysis

2010 ◽  
Author(s):  
Tomislav Bajs ◽  
Damir Konjarek ◽  
Ilijana Ivekovic´
Keyword(s):  
Time Window ◽  
Uncertainty Evaluation ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Sensitivity Studies ◽  
Operator Action ◽  
Core Cooling ◽  
Reactor Pressure ◽  
Emergency Operating

Validation of EOPs (Emergency Operating Procedures) relies on the best-estimate analysis of the transient scenarios. In order to cover associated uncertainties, usually limited number of sensitivity studies is performed for the development of the EOPs in order to identify possible plant states and associated parameters relevant for operator actions. Recently, developed methodologies for the uncertainty evaluation made it possible to evaluate directly uncertainties with the respect to the scenarios analyzed. UMAE (Uncertainty Methodology based on Accuracy Extrapolation) uncertainty methodology has been applied for development of function restoration EOPs. More specifically, Inadequate Core Cooling (ICC) LOCA (Loss of Coolant Accident) scenario has been analyzed using best estimate transient analysis code RELAP5/SCDAPSIM code. Time window for successful operator action has been evaluated following 4.0″ cold leg break near the Reactor Pressure Vessel (RPV) in a 2-loop PWR plant.

Systematic Approach to Transboundary Radioactivity Monitoring for Accidents in External Nuclear Power Plants

2020 ◽  
Author(s):  
Kampanart Silva ◽  
Piyawan Krisanangkura ◽  
Krirerk Phungsara ◽  
Chatchai Chaiyasaen ◽  
Suchin Udomsomporn
Keyword(s):  
Emergency Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Atmospheric Dispersion ◽  
Nuclear Accidents ◽  
Radioactive Materials ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Radioactivity Monitoring ◽  
Dispersion Calculation

Abstract Past nuclear accidents demonstrated that radioactive materials from an accident in a nuclear power station (NPS) can disperse to other countries or even across the globe. This means all countries need to be prepared to respond to a nuclear power emergency even if they have no nuclear power program. This study aims to propose a structured framework for such a country to perform transboundary atmospheric dispersion assessment of an accidental release in an external NPS with limited calculation resources. A trial calculation of a hypothetical release from an interfacing system loss of coolant accident (ISLOCA) in Unit 1 of Fangchenggang NPS during different representative meteorological scenarios is carried out to demonstrate the usability of the proposed framework. It was found that a relatively large release can reach the border of Thailand within 24 hours when the wind along the dispersion pathway is basically in northeast direction with significant amount of rainfall, though it may not be able to trigger the alarm at the radiation monitoring stations. However, it is highly likely that the release that fulfills the aforementioned conditions be detected by one of the stations within 48 hour-timeframe. As the trial calculation could deliver insightful findings with limited calculation resources, the proposed transboundary atmospheric dispersion calculation framework can be used in other non-nuclear power countries to prepare for emergency response to accidents in external NPSs.

Evaluation of IET Facility Applicability on Simulating SBLOCA in Large-Scale Passive PWR Plant

2018 ◽  
Author(s):  
Haozheng Kong ◽  
Bo Kuang ◽  
Pengfei Liu ◽  
Xia Lu ◽  
Yi Yao ◽  
...  
Keyword(s):  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Scaling Analysis ◽  
Flow Area ◽  
Injection Flow ◽  
Parameter Evaluation ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Focused on the overall experimental verification of important phenomena in small break loss of coolant accident (SBLOCA) of large-scale passive PWR plants, this paper put forward a top-down and bottom-up scaling analysis of SBLOCA process based on H2TS method, by which the scaling conditions of IET facility would be figured out to match with the conditions in real nuclear power plants. Meanwhile, parameter evaluation was conducted between IET facility and real power plants, and finally the results proved that the IET facility could, to a certain degree, represent the conditions in real plants in terms of scaling analysis, and the test data of facility could be applicable for the SBLOCA of real plants. Further, in this paper, scenarios of SBLOCA were simulated and compared between IET facility and AP1000 nuclear power plant by the use of Relap5/MOD3 program. Then, after the conversion of flow area and time ratio, the comparison of results between those two systems was conducted in terms of pressure of reactor coolant system (RCS), injection flow rate of CMT and steam flow rate through ADS-1/2/3/4 valves, which also proved that the IET facility could simulate the SBLOCA of real plants well.

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