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%.

Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 38-53
Author(s):  
M. J. Leotlela ◽  
I. Petr ◽  
A. Mathye
Keyword(s):  
Gradual Increase ◽  
Primary Objective ◽  
Fuel Pellet ◽  
Main Body ◽  
Spent Fuel ◽  
Reactivity Effect ◽  
Water Ingress ◽  
Accident Scenarios ◽  
The Impact ◽  
Gain Access

Abstract An essential component of safety analyses is the investigation of accident scenarios. In this paper water ingress scenarios of spent fuel containers, as they may occur during transport or storage, are examined. In the main body of this paper, a number of paths are studied through which water can gain access to the spent fuel cask and eventually reach the fuel pellet, potentially resulting in an increase in reactivity as a result of over-moderation. The primary objective of this project was to perform an assessment of what, in the unlikely event of a Fukushima- type accident, the impact would be on the reactivity of the cask by analyzing a gradual increase in water level in the spent fuel casks. In addition, the way the keff of the system responds to such an increase is discussed. The paper also provides the results of an assessment of the reactivity effect of water ingress via various pathways/channels.

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

Simulation of a Steam Generator Tube Rupture Accident in a Lead-Cooled Accelerator Driven System

2010 ◽  
Author(s):  
Michael Flad ◽  
Shisheng Wang ◽  
Werner Maschek
Keyword(s):  
Steam Generator ◽  
Liquid Water ◽  
Water Injection ◽  
Spent Fuel ◽  
Steam Generator Tube ◽  
Damage Propagation ◽  
Core Damage ◽  
Driven System ◽  
Safety Studies ◽  
The Impact

The European Facility for Industrial Transmutation (EFIT) is developed to transmute long-lived actinides from spent fuel on an industrial scale. In this lead-cooled reactor an intermediate loop is eliminated for economic reasons. Within the framework of design and safety studies the impact of a steam generator tube rupture accident has been investigated. In this postulated event high-pressured liquid water blasts into the lead pool which could trigger various transients. As a major concern steam could be dragged into the core featuring a positive void worth. A thermal lead/water interaction could lead to in-core damage propagation; it could initiate a sloshing of the lead coolant and trigger voiding processes. Furthermore the pressurization of the cover gas needs to be considered. To prove the feasibility of the proposed design these risks are investigated and assessed. Numerical simulations are performed using the advanced safety analysis code SIMMER-III [2]. For the important issue of thermal lead/water interactions the SIMMER code has been validated against Japanese heavy-liquid/water injection experiments.

Integration of Direct/Indirect Influences of Severe Accidents for Improvements of Nuclear Safety

2012 ◽  
Author(s):  
Kampanart Silva ◽  
Yuki Ishiwatari ◽  
Shogo Takahara
Keyword(s):  
Assessment Tool ◽  
Nuclear Safety ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Source Terms ◽  
Fukushima Accident ◽  
Energy Agency ◽  
Severe Accidents ◽  
Level 3 ◽  
The Cost

Risk evaluation is an important assessment tool of nuclear safety, and a common index of direct/indirect influences of severe accidents as a compound of risk is necessary then. In this research, various influences of severe accidents are converted to monetary value and integrated. The integrated influence is calculated in a unit of “cost per severe accident” and “cost per kWh”. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. To calculate the “cost per severe accident” and the “cost per kWh”, typical sequences of severe accidents are picked-up first. Containment failure frequency (CFF) and source terms of each sequence are taken from the results of level 2 probabilistic risk assessment (PRA). The source terms of each sequence is input into the level 3 PRA code OSCAAR which was developed by Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the results presented in this study are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated. It consists of various costs and other influences converted into monetary values. This methodology is applied to a virtual 1,100 MWe BWR-5 plant. Seismic events are considered as the initiating events. The data obtained from the open documents on the Fukushima Accident are utilized as much as possible. Sensitivity analyses are carried out to identify the dominant influences, sensitive assumptions/parameters to the cost per accident or per kWh. Based on these findings, optimization of radiation protection countermeasures is recommended. Also, the effects of sever accident management are investigated.

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

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.

Study on Hydrogen Risk of Spent Fuel Compartment Induced by Containment Venting

2017 ◽  
Author(s):  
Yabing Li ◽  
Xuewu Cao
Keyword(s):  
Control System ◽  
Heat Removal ◽  
Compartment Model ◽  
Severe Accident ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Spent Fuel Pool ◽  
Heat Removal System ◽  
And Control ◽  
Removal System

Hydrogen risk in the spent fuel compartment becomes a matter of concern after the Fukushima accident. However, researches are mainly focused on the hydrogen generated by spent fuels due to lack of cooling. As a severe accident management strategy, one of the containment venting paths is to vent the containment through the normal residual heat removal system (RNS) to the spent fuel compartment, which will cause hydrogen build up in it. Therefore, the hydrogen risk induced by containment venting for the spent fuel compartment is studied for advanced passive PWR in this paper. The spent fuel pool compartment model is built and analyzed with integral accident analysis code couple with the containment analysis. Hydrogen risk in the spent fuel pool compartment is evaluated combining with containment venting. Since the containment venting is mainly implemented in two different strategies, containment depressurization and control hydrogen flammability, these two strategies are analyzed in this paper to evaluated the hydrogen risk in the spent fuel compartment. Result shows that there will not be significate hydrogen built up with the hydrogen control system available in the containment. However, if the hydrogen control system is not available, venting into the spent fuel pool compartment will cause a certain level of hydrogen risk there. Besides, suggestions are made for containment venting strategy considering hydrogen risk in spent fuel pool compartment.

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.

Numerical Two-Phase Flows Simulation and Analysis of the Evolution of the Local Hydrogen Concentration in a PWR Nuclear Containment in the Event of a Severe Accident

2016 ◽  
Author(s):  
Alexandre Zanchetti ◽  
Mickael Hassanaly ◽  
Hervé Cordier ◽  
Antonio Sanna ◽  
Namane Mechitoua ◽  
...  
Keyword(s):  
Hydrogen Concentration ◽  
Nuclear Power ◽  
Cfd Simulation ◽  
Severe Accident ◽  
Water Spray ◽  
Hydrogen Distribution ◽  
Two Phase ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Spray System

The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a water cooled reactor building. Some mitigation means against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and hydrogen combustion are necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on CFD simulation of hydrogen distribution in penalized conditions. After dealing with cases for which the water spray system was assumed to be unavailable, and so treated with single-phase CFD code [1] [2], the present paper content is now about simulation and analysis of the local hydrogen concentration in the case of a severe accident for which the water spray system is available. Numerical developments of a multi-phase CFD code (Neptune_CFD) and code validation lead to consistent simulations. The numerical simulation performed by EDF confirms the favorable safety impact of water spray on pressure and temperature for a LOCA scenario occurring on a 1300 MWe Pressurized Water Reactor. Nevertheless, CFD results show that the activation of the spray system before hydrogen injection gives greater hydrogen concentration. So, in the future, to better assess hydrogen risk, EDF will perform computations at CFD taking into account the interaction between combustion and water sprays.

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

2013 ◽  
Vol 2013 ◽  
pp. 1-15 ◽  
Author(s):  
Andrej Prošek ◽  
Leon Cizelj
Keyword(s):  
Management Strategies ◽  
Severe Accident ◽  
Stress Tests ◽  
Pressurized Water ◽  
Makeup Water ◽  
Reactor Coolant ◽  
The Core ◽  
Station Blackout ◽  
Accident Sequences

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.

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