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.

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.

Irradiation Hardening of Austenitic Stainless Steels: From Field Experience to Modeling

2004 ◽  
Author(s):  
C. Pokor ◽  
Y. Thebault ◽  
C. Pujol ◽  
J.-P. Massoud ◽  
D. Loisnard ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Pressurized Water Reactors ◽  
Iterative Approach ◽  
Pressurized Water ◽  
Irradiation Hardening ◽  
The Core ◽  
Water Reactors ◽  
Life Predictions

Materials for the core internals of Pressurized Water Reactors (austenitic stainless steels) are submitted to neutron irradiation. To understand the ageing mechanisms associated to irradiation and propose life predictions of the component, a multi step iterative approach consisting in particular in modeling the evolution of the hardening has been undertaken. Combination of characterization and modeling of simplified situations and field expertise is proposed.

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.

scholarly journals Robustness Study of Electro-Nuclear Scenario under Disruption

2021 ◽  
Vol 2 (1) ◽  
pp. 1-8
Author(s):  
Jiali Liang ◽  
Marc Ernoult ◽  
Xavier Doligez ◽  
Sylvain David ◽  
Léa Tillard ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Pressurized Water Reactors ◽  
Nuclear Wastes ◽  
Fast Reactors ◽  
Pressurized Water ◽  
Static Strategy ◽  
The Future ◽  
Political Choices ◽  
Water Reactors ◽  
Development Objective

As the future of nuclear power is uncertain, only choosing one development objective for the coming decades can be risky; while trying to achieve several possible objectives at the same time may lead to a deadlock due to contradiction among them. In this work, we study a simple scenario to illustrate the newly developed method of robustness study, which considers possible change of objectives. Starting from the current French fleet, two objectives are considered regarding the possible political choices for the future of nuclear power: A. Complete substitution of Pressurized Water Reactors by Sodium-cooled Fast Reactors in 2180; B. Minimization of all potential nuclear wastes without SFR deployment in 2180. To study the robustness of strategies, the disruption of objective is considered: the objective to be pursued is possibly changed abruptly from A into B at unknown time. To minimize the consequence of such uncertainty, the first option is to identify a robust static strategy, which shows the best performance for both objectives A and B in the predisruption situation. The second option is to adapt a trajectory which pursues initially objective A, for objective B in case of the disruption. To identify and to analyze the adaptively robust strategies, outcomes of possible adaptations upon a given trajectory are compared with the robust static optimum. The temporality of adaptive robustness is analyzed by investigating different adaptation times.

scholarly journals Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

2010 ◽  
Author(s):  
Randall O. Gauntt ◽  
Andrew S. Goldmann ◽  
Kenneth C. Wagner ◽  
Dana Auburn Powers ◽  
Scott G. Ashbaugh ◽  
...  
Keyword(s):  
Mixed Oxide ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Source Terms ◽  
Pressurized Water ◽  
Water Reactors ◽  
Enriched Uranium

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.

Modeling and Phenomenological Aspects of Severe Accidents in Integral Pressurized Water Reactors

2011 ◽  
Author(s):  
I. K. Madni ◽  
M. Khatib-Rahbar
Keyword(s):  
Computer Code ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Modeling Framework ◽  
Pressurized Water ◽  
Product Release ◽  
Severe Accidents ◽  
Water Reactors ◽  
Accident Conditions ◽  
Coolant System

This paper focuses on modeling and phenomenological issues relevant to analysis of severe accidents in integral Pressurized Water Reactors (iPWRs). It identifies relevant thermal-hydraulics, melt progression and fission product release and transport phenomena, and discusses the applicability of the MELCOR computer code to modeling of severe accidents in iPWRs. Areas where the current MELCOR severe accident modeling framework has limitations in the representation of phenomenological processes are identified and examples of possible modeling remedies are discussed. The paper identifies modeling and phenomenological issues that contribute to differences in the calculated reactor coolant system and containment response for iPWRs as compared to traditional PWRs under severe accident conditions.

The core design procedure SA V 90 for pressurized water reactors / Das Kernauslegungsverfahren SA V 90 für Druckwasserreaktoren.

Kerntechnik ◽  
1992 ◽  
Vol 57 (1) ◽  
pp. 37-41
Author(s):  
K. Koebke ◽  
G. Ambrosius ◽  
L. Hetzelt ◽  
S. Merk ◽  
H.-J. Winter
Keyword(s):  
Design Procedure ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors
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