Use of PRA in the Design of the Westinghouse AP1000 Plant

2009 ◽  
Author(s):  
Robert J. Lutz ◽  
James H. Scobel ◽  
Richard G. Anderson ◽  
Terry Schulz
Keyword(s):  
Cooling Water ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Pressurized Water ◽  
Core Damage ◽  
Service Water ◽  
Containment Shell ◽  
Water Reactors ◽  
Accident Scenarios

Probabilistic Risk Assessment (PRA) has been an integral part of the Westinghouse AP1000, and the former AP600, development programs from its inception. The design of the AP1000 plant is based on engineering solutions to reduce or eliminate many of the dominant risk contributors found in the existing generation of Pressurized Water Reactors (PWRs). Additional risk reduction features were identified from insights gained from the AP1000 PRA as it evolved with the design of the plant. These engineered solutions include severe accident prevention features that resulted in a significant reduction in the predicted core damage frequency. Examples include the removal of dependencies on electric power (both offsite power and diesel generators) and cooling water (service water and component cooling water), removal of common cause dependencies by using diverse components on parallel trains and reducing dependence on operator actions for key accident scenarios. Engineered solutions to severe accident consequence mitigation were also used in the AP1000 design based on PRA insights. Examples include in-vessel retention of molten core debris to eliminate the potential for ex-vessel phenomena challenges to containment integrity and passive containment heat removal through the containment shell to eliminate the potential for containment failure due to steam overpressure. Additionally, because the accident prevention and mitigation features of the AP1000 are engineered solutions, the traditional uncertainties associated with the core damage and release frequency are directly addressed.

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

2020 ◽  
Vol 7 (3) ◽  
Author(s):  
Andrea Bachrata ◽  
Fréderic Bertrand ◽  
Nathalie Marie ◽  
Fréderic Serre
Keyword(s):  
Mitigation Strategies ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Physical Parameters ◽  
Fast Reactors ◽  
Pressurized Water ◽  
The Core ◽  
Sodium Fast Reactors ◽  
Water Reactors ◽  
Accident Scenarios

Abstract The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in pressurized water reactors (PWR) compared to those potentially occurring in future GenIV-type sodium fast reactors (SFR). First, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. This paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III, and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

scholarly journals Experimental investigation of PWR accident scenarios at the PKL test facility

2019 ◽  
Vol 137 ◽  
pp. 01016 ◽  
Author(s):  
Rafał Bryk ◽  
Lars Dennhardt ◽  
Simon Schollenberger
Keyword(s):  
Severe Accident ◽  
Test Facility ◽  
Pressurized Water ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Core Cooling ◽  
Safety Systems ◽  
Reactor Pressure ◽  
Additional Loss

PKL is the only test facility in Europe that replicates the entire primary side and the most important parts of the secondary side of western-type Pressurized Water Reactors (PWR) in the scale of 1:1 in heights. It is also worldwide the only test facility with 4 identical reactor coolant loops arranged symmetrically around the Reactor Pressure Vessel (RPV) for simulation of nonsymmetrical boundary conditions between the reactor loops. Thermal-hydraulic phenomena observed in PWRs are simulated in the PKL test facility for over 40 years. The analyses carried out in these years encompass a large spectrum of accident scenario simulations and corresponding cool-down procedures. The overall goal of the PKL experiments is to show that under accident conditions - even for extreme and highly unlikely assumptions as additional loss of safety systems - the core cooling can be maintained or re-established by automatic or operator- performed procedures and that a severe accident e.g. a core melt-down can be avoided under all circumstances. Another goal of the tests performed in the PKL facility is the provision of data for validation of thermal-hydraulic system codes. This paper presents recent modifications of the PKL facility, applied in order to adapt the facility to the latest western-type designs currently built in the world. The paper discusses also important results obtained in the last years.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

Investigations on Catalyst Systems for Catalytic Recombiners

2002 ◽  
Author(s):  
Ernst-Arndt Reinecke ◽  
Peter Broeckerhoff ◽  
Inga M. Tragsdorf
Keyword(s):  
Catalyst Activity ◽  
Heat Removal ◽  
Gaseous Mixture ◽  
Severe Accident ◽  
Experimental Investigations ◽  
Light Water Reactors ◽  
Reaction Heat ◽  
Conversion Rates ◽  
Water Reactors ◽  
Catalyst Systems

Passive Autocatalytic Recombiners (PARs) are used for hydrogen removal in the containments of Light Water Reactors after a severe accident. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces already at low temperatures generating steam and heat. One major concern is the fact that existing recombiners bear the risk of ignition of the gaseous mixture by overheated catalytic substrates or parts of the casing, since the heat generated is not removed by cooling systems. Overheating may occur due to insufficient heat removal. Experimental investigations on existing systems show that the highest temperatures appear near the leading edges of the catalyst sheets. Furthermore, local conversion rates are too high not allowing sufficient reaction heat removal by convection. Possible countermeasures are additional cooling or limiting local conversion rates. At FZJ investigations are made on adapting the catalyst activity according to the requirements by using electro-plating technology instead of washcoat coatings, allowing well defined coating densities. Substrates with corresponding coatings have been tested, proving their ability in mixtures up to oxygen limitation. Different substrate materials and pre-treating measures are investigated for optimizing the surface properties. SEM-studies give insight in the surface structure and allow detailed analysis of the catalyst activity.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

Risk-Informed Analysis for Accident Tolerant Fuel in Pressurized Water Reactors

2020 ◽  
Author(s):  
Zhegang Ma ◽  
Carlo Parisi ◽  
Cliff Davis ◽  
Sai Zhang ◽  
Hongbin Zhang
Keyword(s):  
System Analysis ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Damage ◽  
Systems Research ◽  
Research Activities ◽  
Near Term ◽  
Accident Scenarios

Abstract This paper presents the research activities performed by Idaho National Laboratory (INL) for the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research, using the probabilistic risk assessment (PRA) tool SAPHIRE and the deterministic best estimate tool RELAP5-3D for risk-informed analysis. The ERP research supports DOE and industry initiatives by developing Accident Tolerant Fuel (ATF), the Diverse and Flexible Coping Strategy (FLEX), and passive cooling system designs to enhance existing reactors’ safety features (both active and passive) and to substantially reduce operating costs of nuclear power plants (NPPs) through risk-informed approaches to analyze the plant enhancements and their characterization. The risk-informed analysis used SAPHIRE and RELAP5-3D to evaluate the risk impacts from near-term ATF (FeCrAl and Chromium-coated clads) on a generic Westinghouse three-loop pressurized water reactor (PWR) under the following accident scenarios: station blackout (SBO), loss of feedwater (LOFW), steam generator tube rupture (SGTR), loss-of-coolant accidents (LOCAs), locked rotor transient, turbine trip transient, anticipated transient without scram (ATWS), and main steam line break (MSLB). The RELAP5-3D simulations included the time to core damage, time to 0.5 kilograms hydrogen generation, and total hydrogen generation. The simulation results show there are modest gains of coping time (delay of time to core damage) due to efficacy of the near-term ATF designs in various accident scenarios. The risk benefits on behalf of the core damage frequency (CDF) brought by the ATF designs would be small for most of the scenarios. However, results revealing much less hydrogen being produced at the time of core damage show a clear benefit in adopting ATFs.

Assessment of Radiological Source Term Releases for Potential Severe Accident Scenarios in a PWR SFP

2018 ◽  
Author(s):  
Kwang-Il Ahn ◽  
Jae-Uk Shin
Keyword(s):  
Water Injection ◽  
Severe Accident ◽  
Base Case ◽  
Source Terms ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Makeup Water ◽  
Accident Scenarios ◽  
The Impact

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.

The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-Stress

2009 ◽  
Author(s):  
Mark T. EricksonKirk ◽  
Terry L. Dickson
Keyword(s):  
Structural Integrity ◽  
The United States ◽  
Integrated Analysis ◽  
Total Risk ◽  
Pressurized Water ◽  
Best Estimate ◽  
Water Reactors ◽  
Using Data ◽  
Accident Scenarios ◽  
Reactor Operating

Warm pre-stress, or WPS, is a phenomenon by which the apparent fracture toughness of ferritic steel can be elevated in the fracture mode transition if crack is first “pre-stressed” at an elevated temperature. Taking proper account of WPS is important to the accurate modeling of the postulated accident scenarios that, collectively, are referred to as pressurized thermal shock, and to the accurate modeling of routine cool-down transients. For both accident and routine cool-downs the transients begin at the reactor operating temperature (approximately 290°C for pressurized water reactors in the United States) and proceed to colder temperatures as time advances. The probabilistic fracture mechanics code FAVOR, which is being used by the NRC to provide the technical basis for risk-informed revisions of 10 CFR 50.61 and 10 CFR 50 Appendix G, adopts a model of WPS as part of its fracture driving force module. In this paper we assess the conservatism inherent to the FAVOR WPS model relative to a best-estimate WPS model constructed using data recently produced by the European Commission “SMILE” project and published by Moinereau and colleagues. Assessments of the conservatisms inherent to the so-called “conservative principle” WPS model, and also to a classic LEFM model that does not credit WPS are also made. The data presented herein demonstrate that, for an integrated analysis of PTS risk, the FAVOR and conservative principle WPS models both over-estimate the vessel failure risk by a factor of between 2 and 3× relative to the best estimate model. Our examination of the effect of WPS models on the predictions of individual transients reveals that for the severe transients that dominate risk there is little difference (usually less than 4×) between the conditional probabilities of crack initiation and of through wall cracking predicted by the different WPS models. There are considerable differences in the predictions of the various WPS and non-WPS models for low severity transients, however, the contribution of these transients to the total risk of vessel failure is small.

PRELIMINARY METHODOLOGY FOR THE ANALYSIS OF THE NATIONAL RESEARCH UNIVERSAL REACTOR USING INTEGRATED SEVERE ACCIDENT MODELLING CODES

2015 ◽  
Vol 4 (1) ◽  
pp. 53-65 ◽  
Author(s):  
A.C. Morreale ◽  
M.J. Brown ◽  
S.M. Petoukhov
Keyword(s):  
Large Scale ◽  
Aluminum Matrix ◽  
Heat Sinks ◽  
Severe Accident ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Fuel Design ◽  
Water Reactors ◽  
Accident Case ◽  
Accident Modelling

The National Research Universal (NRU) Reactor is a multi-purpose research reactor located at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories. The severe accident case for the NRU has been explored through deterministic and probabilistic safety analysis (PSA) including multi-level PSAs that detail the progression and consequences of a severe accident in the NRU. These previous calculations lack the interconnected and comprehensive features of a full severe accident modelling code that is now the standard for severe accident analysis of power reactors. It was of interest within AECL to evaluate modern severe accident modelling codes to the NRU reactor case to enhance the understanding of accident progression and predict the system damage and radiation release consequences of a severe accident, which is a very low probability event. The NRU is smaller and operates at a lower power than the large scale power reactors (e.g., pressurized heavy water reactors, pressurized water reactors, and boiling water reactors) that these codes were designed to analyze. Additionally, the NRU has a unique design different from the power reactors and several features relevant to severe accidents including filtered venting, large passive heat sinks, and a dispersion fuel design of uranium-silicide in an aluminum matrix. The major severe accident analysis codes available to AECL and their applicability to the NRU are explored in this paper. In addition, a preliminary strategy for employing the most applicable codes to the NRU for the purposes of severe accident modelling is proposed.

Sign in / Sign up

Export Citation Format

Share Document