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.

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.

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%.

scholarly journals Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

2017 ◽  
Vol 2017 ◽  
pp. 1-25 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Light Water Reactors ◽  
Control Rod ◽  
Light Water ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

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.

Requirement and Discussion of Design Extension Conditions

2017 ◽  
Author(s):  
Yan Jinquan ◽  
Chen Song ◽  
Tian Lin ◽  
Wang Minglu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Safety ◽  
Severe Accident ◽  
Plant Design ◽  
Plant Safety ◽  
Severe Accidents ◽  
Safety Code ◽  
Nuclear Power Plant Safety

Nuclear safety especially severe accidents risks are of great concerns of nuclear power plant. Design consideration of severe accident prevention and mitigation is generally required by various nuclear safety authorities worldwide. However, those requirements related to severe accidents consideration are somewhat different from country to country. Recently, the International Atomic Energy Agency (IAEA) updated and published a safety code on Specific Safety Requirement of Nuclear Power Plant Safety: Design (SSR-2/1). Meanwhile, the Chinese National Nuclear Safety Administration (NNSA) also revised and updated the safety code on Requirement of Nuclear Power Plant Safety in Design (HAF102). In these two codes, both IAEA and NNSA established some new requirements, among which two are of great concern. One is Design Extension Conditions (DEC) for consideration of those conditions traditionally called Beyond Design Basis Accidents (BDBA) in design of nuclear power plant, another is requirement of practically eliminating large release of radionuclide. These two new requirements are internally related, somewhat different and more restrict from those related to severe accident requirements set forth by Nuclear Regulatory Committee of United States (USNRC). Up to date, there are no specific guidelines about engineering implementation of those new safety codes. This paper present an overview of those requirements from IAEA, WENRA, NRC and China NNSA, followed by discussion of engineering approach for the implementation of the DEC requirement set forth by safety authorities.

Innovative Practice of Severe Accidents Measures in CAP1400

2017 ◽  
Author(s):  
Likai Fang ◽  
Xin Liu ◽  
Guobao Shi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Power Sources ◽  
Severe Accidents ◽  
Accident Management ◽  
Accident Conditions

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.

scholarly journals Analysis of Fuel Criticality during Severe Accidents

2015 ◽  
pp. 3-8
Author(s):  
I. Bilodid ◽  
J. Duspiva
Keyword(s):  
Severe Accident ◽  
Stress Tests ◽  
Fukushima Accident ◽  
Management Guidelines ◽  
Severe Accidents ◽  
Design Basis ◽  
Accident Management ◽  
The World ◽  
Criticality Safety

Interest in the analysis of beyond design basis accidents, involving a combination of several failures with fuel damage, has increased throughout the world after the Fukushima accident. Stress tests were performed at NPPs, and development of severe accident management guidelines was started. These activities necessitated calculations to analyze the probability of beyond design basis accidents and assess their initiating events and consequences. One of the aspects in analysis of beyond design basis accidents is to determine the potential for re-criticality during such accidents. The paper provides results of some criticality safety calculations for VVER reactors performed, in particular, by ÚJV Řež and SSTC NRS experts. It is shown how criticality can occur in different severe accident phases.

Preliminary Study on the Identification of DECs Without Significant Fuel Degradation in New NPP Designs

2017 ◽  
Author(s):  
Zhiyi Yang ◽  
Yimin Chong ◽  
Chun Li ◽  
Jiajia Zhang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Severe Accident ◽  
Concept Design ◽  
Energy Agency ◽  
Safety Level ◽  
Safety Requirements ◽  
Specific Safety

New nuclear safety objectives and principles are being studied in main nuclear power countries and organizations after Fukushima Dai-ichi nuclear accident, to further improve the safety level of nuclear power plants (NPPs). Based on International Atomic Energy Agency (IAEA) Specific Safety Requirements (No.SSR-2/1), “Safety of Nuclear Power Plants: Design” (HAF102-2016) is issued in China. The concept “design extension condition (DEC)” is put forward, which is intend to enhance the plant’s capability to withstand accidents that are more severe than Design Basis Accidents (DBA). DEC could include conditions without significant fuel degradation (DEC-A in this paper) and conditions with core melting (DEC-B in this paper), e.g. severe accident. In this paper, the DEC-A and its application was discussed preliminarily, firstly, the development and connotation was introduced, then the identification of DEC-A, and the safety analysis principles of DEC-A were mainly described. This study may play a valuable role for implementation of new nuclear safety requirements in China.

Severe Accident Related Vulnerabilities, Potential Design Enhancements and Opportunities for International Cooperation in Risk Reduction in Pressurized Heavy Water Reactors

2015 ◽  
Author(s):  
Sunil Nijhawan
Keyword(s):  
Single Unit ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Mechanical Degradation ◽  
Optimum Number ◽  
Source Terms ◽  
Regulatory Requirement ◽  
Severe Accidents ◽  
Design Basis ◽  
Water Reactors

Operating CANDU PHWRs present significant challenges with respect to their ability to mitigate accidents that are beyond the envelope of design basis drafted over 40 years ago. Today, consideration of severe accidents is a public as well as a regulatory requirement whose implementation begs serious reconsideration in an international coordinated effort. The PHWR enhanced vulnerabilities to accidents such as a sustained loss of AC power, as in Fukushima, arise not only out of the inherent design features but also out of the institutional arrangements that surround their licensing. For one, the reactors will, in absence of a containing pressure vessel as in PWRs, put fission product activity directly into the containment, sport multiple potential containment bypass vulnerabilities and produce copious amounts of flammable gases due to presence of large amounts of Zircaloy in fuel channels and carbon steel in feeders. The relatively thin walled, stepped, welded Calandria vessel into which the disassembled core debris will rest has potential to mechanically fail early, causing explosive and energetic interactions of hot debris with enveloping water. This can catastrophically fail the reactor structures. For severe accidents the containments, well designed for design basis accidents, are either small and weak as in single unit plants or unable to practically take any significant over pressure in negative pressure multi-unit reactor buildings that depend upon a single vacuum building, too small for a multi-unit severe accident. The paper presents analytical arguments in support of these observations, lists conclusions from a series of design reviews and discusses development of ROSHNI, a new generation PHWR dedicated computer code package for simulating an unmitigated station blackout scenario. It does not directly address the institutional issues that handicap a potent reduction of the residual risk posed by continued operation of these reactors without serious design upgrades but discusses the regulatory failures in this regard. It introduces ROSHNI, a newly developed severe accident simulation package that models the reactor core in a greatly enhanced detail necessitated by the variability amongst reactor fuel channels. For a single unit CANDU 6 reactor, the code simulates thermal-mechanical degradation of 4,560 fuel bundles in 380 diverse fuel channels individually (for a total of over 70,000 dissimilar fuel entities) and computes source terms into containment of flammable deuterium gas and fission products. A number of questions are raised about differences between Hydrogen source terms and mitigation measures that are being implemented for light water reactors and Deuterium specific reaction kinetics in generation and mitigation that must be clearly differentiated but ignored so far by PHWR operators. A discussion of effectiveness of certain severe accident specific design upgrade measures that have been implemented at some operating plants is also addressed. For example, potential for a smaller than optimum number (for severe accidents) of PARS units to actually cause Deuterium/Hydrogen explosions as an unintended consequence is discussed. Continued reluctance of CANDU utilities to address a long standing issue of inadequacy of reactor overpressure protection is also detailed.

Systematic Source Term Analyses for Level 3 PSA of a BWR With Mark-II Type Containment With THALES-2 Code

2002 ◽  
Author(s):  
Jun Ishikawa ◽  
Ken Muramatsu ◽  
Toru Sakamoto
Keyword(s):  
Nuclear Power ◽  
Failure Modes ◽  
Radioactive Material ◽  
Severe Accident ◽  
Source Terms ◽  
Energy Research ◽  
Core Damage ◽  
Reactor Building ◽  
Level 3 ◽  
Accident Scenarios

The THALES-2 code is an integrated severe accident analysis code developed at the Japan Atomic Energy Research Institute in order to simulate the accident progression and transport of radioactive material for probabilistic safety assessment (PSA) of a nuclear power plant. As part of a level 3 PSA being performed at JAERI for a 1,100MWe BWR-5 with a Mark-II containment, a series of calculations were performed by THALES-2 to evaluate the source terms for extensive accident scenarios. For some of the containment failure modes not modeled in THALES-2, such as steam explosion, simple models were coupled with the analysis results of THALES-2 to estimate the source terms. This paper presents the methods and insights from the analyses. An insight from the analyses was that the source terms depend more strongly on the differences in the containment function failure scenarios, such as overpressure failure, controlled containment venting, and small leakage to the reactor building, than those core damage sequences.

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