Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 Mark-I Station Blackout: Part II – Uncertainty Analysis

2021 ◽  
Vol 383 ◽  
pp. 111398
Author(s):  
Lucas I. Albright ◽  
Nathan Andrews ◽  
Larry L. Humphries ◽  
David L. Luxat ◽  
Tatjana Jevremovic
Keyword(s):  
Uncertainty Analysis ◽  
Severe Accidents ◽  
Station Blackout

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.

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.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

2018 ◽  
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
System Analysis ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Thermal Behaviors ◽  
Severe Accidents ◽  
Nuclear Plants ◽  
Station Blackout ◽  
Water Reactors ◽  
Transient Thermal

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

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.

LBLOCA Initiated Emergency Condition Analysis for a China Three-Loop PWR

2018 ◽  
Author(s):  
Wang Ning ◽  
Chen Lei ◽  
Zhang Jiangang ◽  
Yang Yapeng ◽  
Xu Xiaoxiao ◽  
...  
Keyword(s):  
Severe Accident ◽  
Feed Water ◽  
Fukushima Accident ◽  
Flow Paths ◽  
The Core ◽  
Emergency Condition ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Station Blackout ◽  
Control Volumes

Great interest in severe accident has been motivated since Fukushima accident, which indicates that the probability of severe accident exists even though it is extremely small. Emergency condition is important in decision making in case of severe accident in NPP. Although many studies have been conducted for severe accident, there was necessary to investigate emergency condition of severe accidents that could possibly happen and haven’t been sufficiently analyzed. Since station blackout (SBO) happened in Fukushima accident, a number of studies in severe accidents initiated by SBO have been carried out. Off-site power is assumed to be lost during large break loss of coolant accident (LBLOCA), but there is few study to find out emergency condition during LBLOCA if both of off-site and on-site power are lost. A hypothetical severe accident initiated by LBLOCA along with SBO in a China three-loop PWR was simulated in the paper using MELCOR code. Emergency condition was obtained including start of core uncover, start of zirconium-water reaction, failure of fuel cladding and failure of the lower head. Thermal-hydraulic response of the core during the accident was also analyzed in the paper. The model for this study consists of 46 control volumes (27 in primary loop, 17 in secondary loop, 1 in containment and 1 in environment) and 52 flow paths. High pressure safety injection (HPSI) and low pressure safety injection (LPSI) are lost because of loss of on-site and off-site power, and simultaneously main feed water and auxiliary feed water of the steam generators are lost for the same reason. The accumulator can inject water into the core since it is passive and doesn’t need any power. Results of the study will be useful in gaining an insight into detailed severe accident emergency condition that could happen in a China three-loop PWR and may provide basis for severe accident mitigation.

scholarly journals SOARCA uncertainty analysis of a short-term station blackout accident at the Sequoyah nuclear power plant

2020 ◽  
Vol 145 ◽  
pp. 107495
Author(s):  
N.E. Bixler ◽  
M. Dennis ◽  
K. Ross ◽  
D.M. Osborn ◽  
R.O. Gauntt ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Uncertainty Analysis ◽  
Nuclear Power ◽  
Short Term ◽  
Station Blackout

The Development of a Mechanistic Model of MCCI Under Severe Accidents

2013 ◽  
Author(s):  
Wei Wei ◽  
Kelin Qi ◽  
Fuchang Shan ◽  
Yanfang Chen ◽  
Fude Guo
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Mechanistic Model ◽  
Severe Accident ◽  
Severe Accidents ◽  
Large Scale Analysis ◽  
Loss Of Coolant ◽  
Calculation Results ◽  
Station Blackout ◽  
Core Concrete

This paper describes a mechanistic model of the molten core-concrete interaction (MCCI) process under severe accidents, and selects the Daya-Bay nuclear power plant as the research object to calculate and analyze the process of the MCCI when the station blackout (SBO), or loss of coolant (LOCA) severe accident serial is happened. The calculation results of this procedure are compared with the large-scale analysis programs MELCOR to verify the reasonableness and correctness of the model. The results indicate that the model present in this paper can simulate the MCCI process correctly and reasonably under multi-serial severe accidents.

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.

Sign in / Sign up

Export Citation Format

Share Document