Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

2006 ◽  
Author(s):  
Gert Sdouz
Keyword(s):  
Source Term ◽  
Management Strategies ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Management Measures ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Accident Sequences ◽  
Comprehensive Study

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the untightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the “Station Blackout”-sequence and the “Large Break LOCA”. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a “Large Break LOCA”-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a “Station Blackout”-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors.

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

2017 ◽  
Author(s):  
Jun Ishikawa ◽  
Tomoyuki Sugiyama ◽  
Yu Maruyama
Keyword(s):  
Source Term ◽  
Severe Accident ◽  
Water Pool ◽  
Energy Agency ◽  
Severe Accidents ◽  
Station Blackout ◽  
Carbon Dioxide Co2 ◽  
Accident Sequences ◽  
Term Analysis

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

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.

Assessment of Early-Phase Accident Management Strategies in a Station Blackout Scenario for VVER-1000 Reactors

2010 ◽  
Author(s):  
Polina Tusheva ◽  
Frank Schaefer ◽  
Nils Reinke ◽  
Frank-Peter Weiss
Keyword(s):  
High Pressure ◽  
Early Phase ◽  
Management Strategies ◽  
Feed Water ◽  
Water Tank ◽  
Severe Accidents ◽  
Accident Management ◽  
Station Blackout ◽  
Safety Systems ◽  
Secondary Side

In recent years, many NPPs have developed and implemented severe accident management guidelines (SAMG). It is the primary objective of developing SAMG to prevent or mitigate the consequences of severe accidents by keeping the reactor pressure vessel (RPV) integrity and reducing the load to the containment. In a hypothetical Station Blackout accident all active safety systems are unavailable. Without additional measures this would lead to heating-up of the reactor core with severe core degradation. To avoid or to limit the consequences of a possible core heat up, different accident management strategies can be applied. This paper presents an assessment of early-phase accident management actions for VVER-1000 reactors. In particular Primary Side Depressurization (PSD) is investigated as a basic strategy for managing severe accidents under high pressure conditions. In addition, Secondary Side Depressurization (SSD) is also being investigated. It aims at fast reduction of the secondary pressure and feeding the steam generators’ secondary side with water from the feed water tank or from a different source. In that way, the heat removal from the primary to the secondary side can be significantly enhanced and the core heat-up at high pressure can be delayed. A number of simulations with different criteria for actuation of the PSD procedure and additional SSD were performed using the thermal-hydraulic system code ATHLET. This paper provides a detailed modelling of the reactor coolant system and the required safety systems, analysis of the thermal-hydraulic and safety parameters and description of the physical phenomena. Special attention is given to the possibilities of preventing or at least delaying an extended core heat-up depending on the availability of the operational and safety systems. The effectiveness of the applied accident management measures and the effect on the accident progression were studied in order to assess the maximum response time for operators’ intervention.

Analysis of Potential Hydrogen Risk in the PWR Containment

2006 ◽  
Author(s):  
Deng Jian ◽  
Xuewu Cao
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Management Strategies ◽  
Severe Accident ◽  
Deterministic Approach ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Containment Pressure ◽  
Accident Sequences ◽  
Containment Integrity

Various studies have shown that hydrogen combustion is one of major risk contributors to threaten the integrity of the containment in a nuclear power plant. That hydrogen risk should be considered in severe accident strategies in current and future NPPs has been emphasized in the latest policies issued by the National Nuclear Safety Administration of China (NNSA). According to a deterministic approach, three typical severe accident sequences for a PWR large dry containment, such as the large break loss-of-coolant (LLOCA), the station blackout (SBO), and the small break loss-of-coolant (SLOCA) are analyzed in this paper with MELCOR code. Hydrogen concentrations in different compartments are observed to evaluate the potential hydrogen risk. The results show that there is a great amount of hydrogen released into the containment, which causes the containment pressure to increase and some potential inconsecutive burnings. Therefore, certain hydrogen management strategies should be considered to reduce the risk to threaten the containment integrity.

ASTEC: An Integral Code for Simulation of Severe Light Water Reactor Accidents

2006 ◽  
Author(s):  
N. Reinke ◽  
K. Neu ◽  
H.-J. Allelein
Keyword(s):  
Fission Products ◽  
Feasibility Studies ◽  
Current Status ◽  
Light Water ◽  
Fast Running ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Station Blackout ◽  
Water Reactors ◽  
Reactor Types

The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. Thus, the main fields of ASTEC application are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 studies as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses, as well as a combination of multiple modules for coupled effects testing and integral analyses. Among activities concentrating on the validation of individual ASTEC modules describing specific phenomena, the applicability to reactor cases marks an important step in the development of the code. Feasibility studies on plant applications have been performed for several reactor types such as the German Konvoi PWR 1300, the French PWR 900, and the Russian VVER-1000 and −440 with sequences like station blackout, small- or medium-break loss-of-coolant accident, and loss-of-feedwater transients. Subject of this paper is a short overview on the ASTEC code system and its current status with view to the application to severe accidents sequences at several PWRs, exemplified by selected calculations.

scholarly journals Accident Management Actions in an Upper-Head Small-Break Loss-of-Coolant Accident with High-Pressure Safety Injection Failed

2011 ◽  
Vol 175 (3) ◽  
pp. 572-593 ◽  
Author(s):  
César Queral ◽  
Juan González-Cadelo ◽  
Gonzalo Jimenez ◽  
Ernesto Villalba
Keyword(s):  
High Pressure ◽  
Management Actions ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant

Severe accident management measures for a generic German PWR. Part I: Station blackout

2018 ◽  
Vol 122 ◽  
pp. 217-228 ◽  
Author(s):  
P. Wilhelm ◽  
M. Jobst ◽  
Y. Kozmenkov ◽  
F. Schäfer ◽  
S. Kliem
Keyword(s):  
Severe Accident ◽  
Management Measures ◽  
Accident Management ◽  
Station Blackout
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