Two-loop Pressurized Water Reactor Research Articles

Total Loss of Feedwater Analysis of Pressurized Water Reactor Using RELAP5

Abstract In Europe, the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW is multiple failures in which, in addition to the loss of main feedwater, the auxiliary feedwater is also lost. The scenarios without DEC safety features and the scenarios with DEC safety feature assumed have been simulated. The results showed that after TLOFW event initiation, it is very important to trip the reactor as soon as possible. In case of loss of offsite power, the reactor coolant pumps stop, and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed, the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety features assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited in the first 30 min of event. When primary side bleed and feed is used, less time is available for operator actions.

RELAP5/MOD3.3 Analyses of Core Heatup Prevention Strategy during Extended Station Blackout in PWR

The accident at the Fukushima Dai-ichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power for several days, so called extended station blackout (SBO). A set of measures have been proposed and implemented in response of the accident at the Fukushima Dai-ichi nuclear power plant. The purpose of the study was to investigate the application of the deterministic safety analysis for core heatup prevention strategy of the extended SBO in pressurized water reactor, lasting 72 h. The prevention strategy selected was water injection into steam generators using turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump. Method for assessment of the necessary pump injection flowrate is developed and presented. The necessary injection flowrate to the steam generators is determined from the calculated cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed method allows assessment of the necessary injection flowrates of pump, TD-AFW or portable, for different plant configurations and number of flowrate changes. The RELAP5/MOD3.3 Patch04 computer code and input model of a two-loop pressurized water reactor is used for analyses, assuming different injection start times, flowrates and reactor coolant system losses. Three different reactor coolant system (RCS) coolant loss pathways, with corresponding leakage rate, can be expected in the pressurized water reactor (PWR) during the extended SBO: normal system leakage, reactor coolant pump seal leakage, and RCS coolant loss through letdown relief valve unless automatically isolated or until isolation is procedurally directed. Depressurization of RCS was also considered. In total, six types of RCS coolant loss scenarios were considered. Two cases were defined regarding the operation of the emergency diesel generators. Different delays of the pump injection start following the station blackout were assumed and analysed. For each scenario, two kinds of SBO calculations for two-loop PWR were performed, base and verification. Base calculations were needed to determine necessary minimum flowrate for steam generators feeding in such a way that they are not overfilled or emptied. Namely, it was assumed that instrumentation is not available during extended SBO. The verification calculations have been then performed to verify if the determined minimum flowrates are sufficient to prevent the core heatup. The calculated results show effectiveness of the proposed extended SBO prevention strategy provided that the water injection is available in the first two hours after SBO occurring at full power. If diesel generator is running after loss of offsite power for some time, e.g. one hour, the available time for steam generator water injection is significantly longer. The obtained results demonstrate the need for assessment of the pump injection flowrates before the utilization of the pump for mitigation of the event. The applicability of the developed method for assessment of the required pump injection flowrate has been validated on PWR.

Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor

A comprehensive uncertainty analysis in the event of a severe accident in a two-loop pressurized water reactor is conducted using an uncertainty package integrated in the ASYST code. The plant model is based on the nuclear power plant (NPP) Krško, a Westinghouse-type power plant. The station blackout scenario with a small break loss of coolant accident is analyzed, and all processes of the in-vessel phase are covered. A best estimate plus uncertainty (BEPU) methodology with probabilistic propagation of input uncertainty is used. The uncertain parameters are selected based on their impact on the safety criteria, the operation of the NPP safety systems and to describe uncertainties in the initial and boundary conditions. The number of required calculations is determined by the Wilks formula from the desired percentile and confidence level, and the values of the uncertain parameters are randomly sampled according to appropriate distribution functions. Results showing the thermal hydraulic behaviour of the primary system and the propagation of core degradation are presented for 124 successful calculations, which is the minimum number of required calculations to estimate a 95/95 tolerance limit at the 3rd order of the Wilks formula application. A statistical analysis of the dispersion of results is performed afterwards. Calculation of the influence measures shows a strong correlation between the decay heat and the representative output quantities, which are the mass of hydrogen produced during the oxidation and the height of molten material in the lower head. As the decay heat increases, an increase in the production of hydrogen and the amount of molten material is clearly observed. The correlation is weak for other input uncertain parameters representing physical phenomena, initial and boundary conditions. The influence of the order of the Wilks formula is investigated and it is found that increasing the number of calculations does not significantly change the bounding values or the distribution of results for this particular application.

Numerical study of blowdown period of double-end break loss of coolant accident in PWR based on RELAP5

Loss of coolant accident (LOCA) is one of the most severe accidents of nuclear power plants (NPP). During the early blowdown period, the system is in a high-pressure gradient and high-velocity gradient state. Therefore, the hydrodynamic forces and safety analysis are highly related to calculating thermal-hydraulic parameters in the flow field. This paper studied the factors and modeling methods that affect the calculation of thermal-hydraulic characteristics under large break LOCA conditions. Firstly, the reactor core and primary system of a typical two-loop pressurized water reactor (PWR) NPP is modeled using the system analysis code RELAP5 Mod4.0. The normal operation state is obtained after the steady calculation. Then, transient calculation of double-end break LOCA is implemented considering the effects of break opening time and the pressurizer. The influence of the results on the calculation of hydrodynamic force has also been analyzed. Finally, the modeling method of reactor pressurized vessel (RPV) downcomer is presented. It is found that the flow field can be better simulated when the downcomer is modeled with parallel channels and several transverse junctions. Thus, this paper provides meaningful suggestions for modeling and simulation in the early blowdown period under LOCA conditions and contributes to further hydrodynamic load calculation and reactor safety analysis.

Analysis of loss of residual heat removal system during mid-loop operation of a 300 MWe PWR

Thermal-hydraulic response of a two-loop pressurized water reactor is studied on loss of residual heat removal system (RHRS) during mid-loop operation. Decay heat during mid-loop operation is removed by RHRS. On loss of RHRS, no cooling system is available to cool the core. Transient thermal-hydraulic analysis of the plant is performed using RELAP/SCDAP on loss of RHRS. For this purpose, first steady-state is achieved at mid-loop conditions. System obtains steady-state at primary coolant inventory of 58000 kg, water level of 7.8m in the reactor pressure vessel (RPV) and upper plenum pressure of 104.41 kPa. On loss of RHRS, the core uncovery time, maximum reactor system pressure and cladding temperature are determined under four accidental scenarios in absence of any mitigation measure. The results show that the core uncovery time for the open case of reactor pressure vessel top vent valve (RPV top vent valve), pressurizer relief valve (PZR valve), steam generator man way and pressurizer man way are 423.1, 214.4, 50.1, 67.2 min respectively. The effect of reflux condensation for the above accidental configurations has also been studied. It is found that reflux condensation is a prominent phenomenon to increases core uncovery time for open case of RPV top vent valve and PZR valve.

Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

Integrity analysis of a reactor pressure vessel subjected to a realistic pressurized thermal shock considering the cooling plume and constraint effects

The fracture mechanic analysis of a reactor pressure vessel subjected to Pressurized Thermal Shock (PTS) loading is one of the most important issues for the assessment of life time extension of a nuclear power plant. The most severe scenario occurs during cold water injection in the cold leg due to a Loss-Of-Coolant Accident (LOCA). Two hypothetical LOCAs are assumed for an adopted reference design of a two-loop pressurized water reactor. Boundary conditions obtained from the RELAP5 code are used as input for three-dimensional computational fluid dynamics which provides the accurate description of the transient including the plume cooling effect. The safety assessment considers the comparison of the stress intensity factor for the deepest point of a surface crack front with the fracture toughness of the material. However, the fracture toughness of the materials is influenced by crack constraints determined by the T-stress. Therefore, safety margin of the reactor pressure vessel should be based on the K-T instead of on the K approach. The existence of a cooling plume affects the stress intensity factor and T-stress. Even if the cooling plume affects the T-stress, it is not translated into an increase of the safety margin. Therefore, in a PTS event with pronounced plume cooling a detailed model of the plume has to be considered in the structural analysis.

Investigation of a Hydrogen Mitigation System During Large Break Loss-Of-Coolant Accident for a Two-Loop Pressurized Water Reactor

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

Coupled RELAP5, 3D CFD and FEM analysis of postulated cracks in RPVs subjected to PTS loading

The fracture mechanic analysis of a reactor pressure vessel subjected to pressurized thermal shock loading is one of the most important issues for the assessment of life time extension of a nuclear power plant. The most severe scenario occurs during cold water injection in the cold leg due to a Loss-Of-Coolant Accident (LOCA). In the present study a comprehensive fracture mechanics analysis is performed. Two hypothetical LOCAs are assumed for an adopted reference design of a two-loop Pressurized Water Reactor. Boundary conditions obtained from the RELAP5 code are used as input for Computational Fluid Dynamics (CFD) simulations. For the structural integrity analysis, submodeling technique and the eXtended Finite Element Method (XFEM) based on temperatures calculated by CFD are applied. The results from the 3D FEM calculations are compared to those from a simplified axisymmetric model based on axisymmetric thermal hydraulic model results. The analysis identifies the worst crack orientation and location. It also proves that a complete model is needed for a correct analysis as the simplified model is not conservative and fails to describe accurately the local plume effect.

Computational fluid dynamics study of pressurized thermal shock phenomena in the reactor pressure vessel

One of the most important issues for assessment of life time extension of a nuclear power plant is the Pressurized Thermal Shock (PTS) phenomenon. The most critical scenario takes places during cold water injection in the cold legs after a postulated Loss of Coolant Accident (LOCA). In the present study, a hypothetical medium break LOCA is assumed in one of the hot legs for an adopted reference design of a two-loop Pressurized Water Reactor (PWR). The boundary conditions obtained from RELAP5 code are used as input for the Computational Fluid Dynamics (CFD) simulations in order to study the details of the flow behavior in the downcomer and to provide three-dimensional temperature distributions for the structural mechanics analysis. The calculations showed good mesh convergence results adopting unsteady two-equation modeling approaches for turbulence. Both symmetric and asymmetric injection conditions of the emergency cooling water were analyzed. The results showed highly three-dimensional behavior of flow field and temperature distributions in the downcomer. Large temperature gradients and the maximum von Mises stresses are predicted at the inlet nozzle and at the plume affected areas of the reactor pressure vessel.

Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation

This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal–mechanical stress analysis were performed for typical thermal–mechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in US-NRC report: NUREG-6909.

RELAP5/MOD3.3 Analyses for Prevention Strategy of Extended Station Blackout

The purpose of the study was to investigate the application of deterministic safety analysis for a prevention strategy of the extended station blackout (SBO). A method for the assessment of portable pump flow rates for a steam generator (SG) makeup is proposed. The RELAP5/MOD3.3 computer code and input model of a two-loop pressurized water reactor (PWR) is used for analyses, assuming different injection start times, flow rates, and reactor coolant system (RCS) losses. The calculated results show the effectiveness of the proposed extended SBO prevention strategy. The results demonstrate the need for mitigation of the common cause failure (CCF) potential of the on-site and off-site power sources.

Comparison of PTS analyses of RPVs based on 3D-CFD and RELAP5

The integrity of a reactor pressure vessel related to pressurized thermal shocks is one of the most important issues for the assessment of life time extension of a nuclear power plant. The most critical scenario occurs during cold water injection through the cold leg due to a Loss-Of-Coolant Accident (LOCA). Due to the difficulties associated with the crack modeling with the three-dimensional finite element method (FEM), simple geometries and crack configurations are usually employed.In the present study, a hypothetical medium break LOCA is assumed in one of the hot legs for an adopted reference design of a two-loop pressurized water reactor. The boundary conditions obtained from RELAP5 calculations are used as input for the three-dimensional computational fluid dynamics simulations in order to provide three-dimensional temperature distribution for the structural mechanics analysis in which submodeling and eXtended FEM (XFEM) are applied. The results from these three-dimensional computations are compared with those from simplified axisymmetric models based on reference data temperatures.

A Stylized Three-Dimensional Pressurized Water Reactor Benchmark Problem with UO2 and MOX Fuel

This paper fully describes a whole-core benchmark problem based on a small two-loop pressurized water reactor composed of UO2 and MOX fuel assemblies. The specification includes heterogeneity at both the assembly and the core levels. The geometry and material compositions are fully described, and multigroup material cross-section libraries are provided in the two-group, four-group, and eight-group formats. Detailed Monte Carlo reference solutions including core eigenvalue, assembly-averaged fission distribution, and selected fuel pin fission density distributions are presented for benchmarking diffusion and transport methods. Three different core configurations are presented in this paper, namely, all-rods-out, all-rods-in, and some-rods-in.

Advanced progressive real coded genetic algorithm for nuclear system availability optimization through preventive maintenance scheduling

One of the main goals of reliability-centered maintenance programs is to find optimal maintenance strategies. Availability of emergency systems through preventive maintenance scheduling has an important role in Probabilistic Safety Analysis (PSA). Preventive maintenances can be replaced with reloading time maintenance and reduce unavailability of the system, especially in the last days of reload cycle. Of course, the emergency core cooling systems connect to circuit in accidents and control the intensity of the unavailability. For reducing unavailability of standby emergency systems, preventive maintenance scheduling method is offered. In this paper, an Advanced Progressive Real Coded Genetic Algorithm (APRCGA) is applied to optimize the availability of standby systems with preventive maintenance scheduling. Using APRCGA in nuclear maintenance systems and considering appropriate objective functions, the most suitable conditions are obtained. The preventive maintenance scheduling keeps unavailability of systems within safe and reliable conditions. In order to demonstrate the effectiveness of the proposed method, it is applied for two nuclear power plant emergency systems. First, for the nuclear emergency core cooling system of a two-loop pressurized water reactor, most available maintenance scheduling is presented. In the second case, for a four-loop pressurized water reactor, the maintenance scheduling for emergency core cooling system is proposed. The results are compared with those obtained by some standard maintenance policies, and previously published papers.

Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs) leaks assumed) to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS). For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

The plant-specific uncertainty analysis for an ex-vessel steam explosion-induced pressure load using a TEXAS–SAUNA coupled system

An ex-vessel steam explosion has been considered as one of the most challenging severe accident phenomena to the integrity of the reactor cavity and containment of a nuclear power plant, owing to its rapid and dynamic characteristics. The purpose of this paper is to provide plant-specific uncertainty analysis results on the ex-vessel steam explosion-induced pressure loads, which can be used as key input to assess the conditional failure probability if the fragility structures of interest is provided. The APR1400, a two-loop pressurized water reactor, has been selected as a reference plant for an uncertainty analysis. For this purpose, a comprehensive uncertainty analysis has been performed for key thermal–hydraulic conditions of the reactor pressure vessel and cavity, which can highly influence on these pressure loads, with the help of a coupling of a steam explosion analysis code (TEXAS-V) with a sampling-based uncertainty quantification code (SAUNA). To get a more robust conclusion based on the analysis results, various sensitivity analyses have been applied to both probability types (e.g., normal and uniform PDF) and sampling schemes (e.g., random and Latin Hypercube sampling). Key contributors (i.e., physical and model parameters) to the underlying pressure loads have been determined by assessing the six currently available types of importance measures.

Modular 3-D solid finite element model for fatigue analyses of a PWR coolant system

Abstract The extension of operational licenses of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of the affected components. Analyses, based upon the in-service transient loads should be compared to the loads analyzed at the design stage. The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements (FE). The FE mesh density is crucial for both the accuracy and the cost of the analysis. The main goal of the paper is to propose a set of computational tools which assist a user in a deployment of modular spatial FE model of main components of a typical reactor coolant system, e.g., pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in a system. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. For optimal accuracy, all components are meshed with hexahedral elements with quadratic interpolation. The performance of the model is demonstrated with simulations performed with a complete two-loop PWR coolant system (RCS). Heat transfer analysis and stress analysis for a complete loading and unloading cycle of the RCS are performed. The main results include expected ranges of fatigue loading for the pipe lines and coolant pump components under the given conditions.

The AP1000TM Reactor: Passive Safety and Modular Design

Abstract Our world is ever growing, there will be higher demands on electricity, and fossil fuels cannot satisfy this demand without further harming the environment. Likewise, renewable energy sources such as solar and winds are still in their infancy and, when used alone, are not realistic solutions to meet this demand. Westinghouse Electric Company is ready to address higher electricity demand with the only Generation III + reactor to receive Design Certification from the United States Nuclear Regulatory Commission, the AP1000™ pressurized water reactor (PWR). Westinghouse Electric Company once again sets a new industry standard with the AP1000™ reactor. The AP1000™ is a two-loopPressurized Water Reactor (PWR) with passive safety features and extensive plant simplifications that enhance its construction, operation, maintenance and safety. The paper discusses the unique design features of AP 1000.