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.

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 Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

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.

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.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

The application efficiency of intumescent coatings for power cables of safety systems of nuclear power plants under fire conditions

2021 ◽  
Vol 30 (4) ◽  
pp. 36-47
Author(s):  
O. S. Lebedchenko ◽  
S. V. Puzach ◽  
V. I. Zykov
Keyword(s):  
Outer Surface ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fire Retardant ◽  
Insulation Material ◽  
Power Cables ◽  
Fire Retardants ◽  
Safety Systems ◽  
Fire Conditions

Introduction. The reliable operation of safety systems, that allows for the failure of no more than one safety system component, entails the safe shutdown and cool-down of an NPP reactor in the event of fire. However, the co-authors have not assessed the loss of performance by an insulating material, treated by intumescent compositions and used in the power cables of the above safety systems exposed to the simultaneous effect of various modes of fire and current loads.Goals and objectives. The purpose of the article is the theoretical assessment of the application efficiency of intumescent fire-retardant coatings in power cables used in the safety systems of nuclear power plants having water-cooled and water-moderated reactors under fire conditions. To achieve this goal, the temperature of the outer surface of the insulation and the intumescent fire-retardant coating was analyzed depending on the mode of fire. Theoretical foundations. A non-stationary one-dimensional heat transfer equation is solved to identify the temperature distribution inside the multilayered insulation and the fire-protection layer of a conductive core.Results and their discussion. The co-authors have identified dependences between the temperature of the outer surface of the insulation and the fire retarding composition of the three-core cable VVGng (A)-LS 3x2.5-0.66, on the one hand, and the temperature of the indoor gas environment for three standard modes of fire and one real fire mode. It is found that before the initiation of the process of destruction of the insulation material, the intumescence of the fire-retardant coating occurs only in case of a hydrocarbon fire. Under real fire conditions, the maximal insulation melting time before the initiation of intumescence of the fire-retardant coating at the minimal temperature of intumescence is 4.75 minutes, while the maximal time period from the initiation of destruction of the insulation material to the moment of the insulation melting is 6.0 minutes.Conclusions. An experimental or theoretical substantiation of parameters of intumescent fire retardants, performed using standard modes of fire, has proven the potential loss of operational properties by insulating materials of power cables, used in the safety systems of nuclear power plants, in case of a real fire. Therefore, it is necessary to establish a scientific rationale for the efficient use of fire retardants in the above cables with regard for the conditions of a real fire.

scholarly journals PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE

2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono
Keyword(s):  
Cooling System ◽  
Fuel Temperature ◽  
Blow Down ◽  
Beam Tube ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Dry Out

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

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