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.

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 EXPERIMENTAL STUDY OF DAMAGED CR-COATED FUEL CLADDING IN POST-ACCIDENT CONDITIONS

2020 ◽  
Vol 28 ◽  
pp. 1-7
Author(s):  
Petr Červenka ◽  
Jakub Krejčí ◽  
Ladislav Cvrček ◽  
Vojtěch Rozkošný ◽  
František Manoch ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Experimental Methodology ◽  
Fuel Cladding ◽  
Loss Of Coolant Accident ◽  
Ring Compression ◽  
Loss Of Coolant ◽  
Accident Conditions ◽  
Cr Coating ◽  
The Impact ◽  
Positive Effect

To enhance the safety of nuclear power, the focus of researchers all around the world has recently mainly objected on the development of Accident Tolerant Fuels. Especially the Chromium coating of current Zirconium based cladding has been widely suggested and discussed for its immense positive effect on overall cladding properties. Nevertheless, it was observed that during the first stage of the Loss of Coolant Accident, cracks appear in the Cr coating due to its inability to tolerate higher plastic strain. Therefore, experimental methodology used in this article focuses on testing fuel cladding with damaged Cr coating after the high-temperature transient. The impact of cracks on degradation of cladding mechanical properties was observed using optical microscopy, ring compression test, microhardness, and evaluating hydrogen content and weight gain.

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.

scholarly journals Study on Airborne Radionuclide Dispersion in Floating Nuclear Power Plant under the Loss-of-Coolant Accident

2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Shuliang Zou ◽  
Na Liu ◽  
Binhai Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Effective Dose ◽  
Nuclear Power ◽  
Severe Accident ◽  
Emergency Plan ◽  
Dose Limit ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Dynamics Method

Floating nuclear power plant is a kind of nuclear power plant on a barge moored specifically in an area of the sea. In order to study the factors influencing airborne radionuclide dispersion induced by the loss-of-coolant accident in floating nuclear power plant, the floating nuclear power plant platform was taken as the research object, and the dispersion of airborne radionuclide under combined conditions of platform positions, wind directions, and break directions (north, south, west, and east) was simulated by the CFD (computational fluid dynamics) method. The results show that northern and southern breaks have less dangerous island area than western and eastern ones but have more platform dangerous area than the western and eastern ones. The risk of the southern break is the greatest, and that of the western break is the least. Rotating the floating nuclear power plant platform in a certain angle can reduce the damage of loss-of-coolant accident. The effects of the dose received by the personnel under the condition of the severe accident were evaluated based on previous research, showing that the inhalation effective dose and the effective dose of plume immersion exposure were less than the radiation dose limit of 0.25 Sv within two hours in the accident. The results of the study can provide reference for the design of floating nuclear power plant platform and the formulation of emergency plan.

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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