Analysis on Impact of Fuel Thermal Conductivity Degradation (TCD) on Large Break Loss of Coolant Accidents of CAP1000

2017 ◽  
Author(s):  
Wang Weiwei ◽  
Lu Lu
Keyword(s):  
Thermal Conductivity ◽  
Nuclear Power ◽  
Power Plants ◽  
Sensitivity Study ◽  
Fuel Pellet ◽  
Steady State Condition ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fuel Storage ◽  
High Burnup

Under high burnup conditions, thermal conductivity of fuel pellet degrades, which is referred to as thermal conductivity degradation (TCD). TCD phenomenon influences fuel average temperature and fuel storage energy under steady state condition before loss of coolant accident (LOCA) and further influences peak cladding temperature (PCT) during large break LOCA process. In this study, sensitivity study on double ended guillotine break of cold leg in CAP1000 at different burnup conditions was performed, using large break LOCA analysis code WCOBRA/TRAC and PCTs under different conditions were obtained. The modified NFI (Nuclear Fuels Institute) TCD model was adopted to model fuel conductivity after degradation in analysis and decrease of peaking factors including FQ and FΔh after 30GWD/MTU was also considered. Sensitivity analysis showed that: after considering the influence of TCD and peaking factor burndown, the PCT limiting case did not occur in low burnup range again, but occurred at burup of about 29GWD/MTU. Compared to other burnup points, the first and second peak values of PCT at that burnup point were all at the highest level. Performing of this study could prefer reference for analysis and estimation of large break LOCA of passive nuclear power plants under high burnup conditions.

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.

Small Break LOCA Analysis of ACR-700 NPP

2006 ◽  
Author(s):  
Limin Zheng ◽  
Sen Shen ◽  
David Wright
Keyword(s):  
Nuclear Power ◽  
Sensitivity Study ◽  
Hydraulic Analysis ◽  
Reactor Outlet ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Subsequent Loss ◽  
Safety Systems ◽  
Limiting Case ◽  
Thermal Hydraulic Analysis

A small break loss of coolant accident (SB-LOCA) analysis to assess a preliminary conceptual design of the ACR-700 PHWR nuclear power plant (NPP) developed by AECL has been performed with CATHENA MOD 3.5d, a PHWR system thermal-hydraulic analysis code. The limiting break size has been found by performing a sensitivity study for three different break locations [i.e. reactor inlet header (RIH), HTS pump suction (PS) pipe and reactor outlet head (ROH)] under the limiting case (i.e. SB-LOCA with subsequent loss of class IV power with all safety systems available). The analysis results indicate that the SB-LOCA acceptance criteria are satisfied.

scholarly journals Simulation of VVER-1000 Guillotine Large Break Loss of Coolant Accident Using RELAP5/SCDAPSIM/MOD3.5

2021 ◽  
Vol 2 (4) ◽  
pp. 516-532
Author(s):  
Fabiano Gibson Daud Thulu ◽  
Ayah Elshahat ◽  
Mohamed H. M. Hassan
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Safety Margin ◽  
Fuel Temperature ◽  
Primary Coolant ◽  
Transient Time ◽  
Steady State Condition ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Safety Systems

The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA.

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.

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