scholarly journals EVALUATION METRICS APPLIED TO ACCIDENT TOLERANT FUEL CLADDING CONCEPTS FOR VVER REACTORS

2016 ◽  
Vol 4 ◽  
pp. 89 ◽  
Author(s):  
Martin Sevecek ◽  
Mojmir Valach
Keyword(s):  
Severe Accident ◽  
Evaluation Metrics ◽  
Development Phase ◽  
Fuel System ◽  
Fuel Cladding ◽  
High Interest ◽  
Severe Accidents ◽  
Time Period ◽  
Fukushima Daiichi ◽  
Reactor Types

Enhancing the accident tolerance of LWRs became a topic of high interest in many countries after the accidents at Fukushima-Daiichi. Fuel systems that can tolerate a severe accident for a longer time period are referred as Accident Tolerant Fuels (ATF). Development of a new ATF fuel system requires evaluation, characterization and prioritization since many concepts have been investigated during the first development phase. For that reason, evaluation metrics have to be defined, constraints and attributes of each ATF concept have to be studied and finally rating of concepts presented. This paper summarizes evaluation metrics for ATF cladding with a focus on VVER reactor types. Fundamental attributes and evaluation baseline was defined together with illustrative scenarios of severe accidents for modeling purposes and differences between PWR design and VVER design.

Review on Water Radiolysis in the Fukushima Daiichi Accident: Potential Cause of Hydrogen Generation and Explosion

2014 ◽  
Author(s):  
Genn Saji
Keyword(s):  
Hydrogen Generation ◽  
Hydrogen Gas ◽  
Severe Accident ◽  
Normal Operation ◽  
Water Radiolysis ◽  
Chain Reactions ◽  
Decay Heat ◽  
Severe Accidents ◽  
Heat Curve ◽  
Fukushima Daiichi

Although the water radiolysis, decomposition of water by radiation, is a well-known phenomenon the exact mechanism is not well characterized especially for severe accidents. The author first reviewed the water radiolysis phenomena in LWRs during normal operation to severe accidents (e.g., TMI- and Chernobyl accidents) and performed a scoping estimation of the amount of radiological hydrogen generation, accumulation and release for the Fukushima Daiichi accident. The estimation incorporates the decay heat curve after a reactor trip combined with G-values. As much as 450 cubic meters-STP of accumulated hydrogen gas is estimated to be located inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1. When a set of radiological chain reactions are incorporated the resultant reverse reactions substantially reduce the hydrogen generation, even when removal of molecular products (i.e., oxygen and hydrogen) is assumed stripped rapidly from boiling water through bubbles. Even in the most favorable configuration a typical amount of hydrogen gas reduces to only several tens of cubic meters. Finally, the author tested a new mechanism, “radiation-induced electrolysis,” which had been applied to his corrosion studies for last several years. His theory has been verified with the published in-pile test data, although he has never tried to apply this to his severe accident study. The predicted results indicated that the total inventory of hydrogen gas inside RPV may reach as much as 1000 cubic meters in just 3 hours during the SBO due to a high decay heat soon after the reactor trip through this process.

scholarly journals Analysis of Severe Accidents in Spent Fuel Pool of Fukushima Daiichi NPP Using MELCOR 1.8.6 Computer Code

2016 ◽  
pp. 13-20
Author(s):  
O. Kotsuba ◽  
Yu. Vorobyov ◽  
O. Zhabin ◽  
D. Gumenyk
Keyword(s):  
Computer Analysis ◽  
Computer Code ◽  
Computer Models ◽  
Reliable Data ◽  
Severe Accident ◽  
Spent Fuel ◽  
Severe Accidents ◽  
Spent Fuel Pool ◽  
Fukushima Daiichi ◽  
To Receive

The paper presents specific approaches to modeling the spent fuel pool (SFP) of Fukushima Daiichi NPP and results of thermohydraulic calculations of severe accidents in SFP using MELCOR 1.8.6 computer code. The dynamics of main processes accompanying severe accident progression in SFP of such a type was defined based on computer analysis. Obtained results may be used to improve available SFP computer models to receive more reliable data on the progression of emergency processes in NPP SFPs.

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.

Chinese Public and Nuclear Experts May Have Different Perceptions Of Severe Nuclear Accidents

2021 ◽  
Author(s):  
Hsingtzu Wu ◽  
Leyao Huang
Keyword(s):  
Nuclear Power ◽  
Nuclear Accident ◽  
Severe Accident ◽  
The Public ◽  
Severe Accidents ◽  
White Collar Workers ◽  
Questionnaire Surveys ◽  
Definition Of ◽  
Chinese Public ◽  
Severe Nuclear Accident

Abstract Nuclear power has been a controversial social issue, and societal acceptance is critical to its development and future. In addition, risk informed rules and regulations rely on the public’s understanding. However, there seems a communication gap about nuclear safety between nuclear experts and the public in China, and three questionnaire surveys were conducted to better understand Chinese public’s perceptions of a severe nuclear accident. The sample sizes were 117, 280 and 1071. Most of the respondents were students or white-collar workers born after 1990. In these three surveys, we found that more than 85% of respondents consider a less severe accident as a severe nuclear accident, and most respondents considered an incident to constitute a severe nuclear accident. The results demonstrate that nuclear experts and Chinese public may have different definitions of a severe nuclear accident. Therefore, we suggest that the definition of severe accidents should be better explained to the public to benefit the communication about risk informed rules and regulations. In addition, our three different surveys yielded a similar result, and we anticipate that a questionnaire survey with a larger sample size would do the same.

VINSAP Tool for Visualization of Severe Accident Progress

2017 ◽  
Author(s):  
Martin Kropik ◽  
Jiri Duspiva
Keyword(s):  
Czech Republic ◽  
Staff Training ◽  
Data Transformation ◽  
Severe Accident ◽  
First Year ◽  
The Czech Republic ◽  
Severe Accidents ◽  
Technical University

The contribution provides information about the development of a system for visualization of NPP severe accident progress. This visualization is under development in cooperation of UJV Rez, a.s. and Czech Technical University in Prague. The project is supported by the Technology Agency of the Czech Republic and is planned to be solved from 2015 to 2017. The visualization uses results of an analytical code MELCOR for evaluation of the NPP severe accident progress. The visualization firstly reads MELCOR results, transforms them to a suitable format for quick processing and provides graphical screens with reactor components that could demonstrate the progress of the evaluated severe accident. The visualization can even provide parallel presentation of more different scenarios of the severe accident. The system is planned to be used for training of NPP staff to handle severe accidents. In the first year of the project solution (2015), the software for MELCOR data transformation, next for providing information about transformed data were developed. In the following year (2016), software for creation of graphical screens with reactor components and software for severe accident progress presentation is creating. In the final year of the project (2017), thorough testing is going to be carried out, and the applicability of the visualization for a practical use during a NPP staff training is going to be verified.

scholarly journals Investigation of Corrosion Resistance of Alloys with Potential Application in Supercritical Water-cooled Nuclear Reactors

2019 ◽  
Vol 63 (2) ◽  
pp. 328-332 ◽  
Author(s):  
Ákos Horváth ◽  
Attila R. Imre ◽  
György Jákli
Keyword(s):  
Corrosion Resistance ◽  
Supercritical Water ◽  
Power Plants ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Water Reactors ◽  
Reactor Types ◽  
Materials Used ◽  
Supercritical Water Cooled Reactor ◽  
Nuclear Applications

The Supercritical Water Cooled Reactor (SCWR) is one of the Generation IV reactor types, which has improved safety and economics, compared to the present fleet of pressurized water reactors. For nuclear applications, most of the traditional materials used for power plants are not applicable, therefore new types of materials have to be developed. For this purpose corrosion tests were designed and performed in a supercritical pressure autoclave in order to get data for the design of an in-pile high temperature and high-pressure corrosion loop. Here, we are presenting some results, related to corrosion resistance of some potential structural and fuel cladding materials.

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.

Sign in / Sign up

Export Citation Format

Share Document