EC6 Instrumentation, Control, and Electrical Design for Station Blackouts

2012 ◽  
Author(s):  
Philip Foster ◽  
John Harber ◽  
Steven Ford ◽  
Boris Lekakh ◽  
Chunlei Nie
Keyword(s):  
Critical Role ◽  
Probabilistic Methods ◽  
Severe Accident ◽  
Normal Operation ◽  
Plant Design ◽  
Active Monitoring ◽  
Correct Operation ◽  
Display Devices ◽  
Severe Accidents ◽  
Accident Scenarios

The Enhanced CANDU 6 (EC6) plant design includes complementary design features that assist the goals of preventing severe accident scenarios and mitigating their consequences if they do occur. These features include design or procedural considerations (or both), and are based on a combination of phenomenological models, engineering judgments, and probabilistic methods. This paper provides an overview of the EC6 instrumentation and control (I&C) and electrical design specifically related to prevention of severe accidents during station blackouts (SBOs). The I&C and electrical components and devices may include those used in normal operation, those of the safety systems used during design basis accidents, and those added solely for preventing or mitigating severe accidents. They are evaluated by an equipment survivability assessment to demonstrate a reasonable level of assurance for their operability, and a seismic fragility evaluation in consideration of a seismic initiating event. The requirements for the I&C and electrical design for SBOs are developed by reviewing EC6 event sequences leading to SBOs, identifying the available systems, and identifying the actions needed and whether they are manual or automatic, passive or active. Monitoring plays a critical role in the prevention of severe accidents. The correct operation of instrumentation and associated display devices for key parameters is essential in diagnosing the plant state and safety challenges.

scholarly journals Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

Coatings ◽  
2021 ◽  
Vol 11 (5) ◽  
pp. 557
Author(s):  
Egor Kashkarov ◽  
Bright Afornu ◽  
Dmitrii Sidelev ◽  
Maksim Krinitcyn ◽  
Veronica Gouws ◽  
...  
Keyword(s):  
Protective Coatings ◽  
Zirconium Alloys ◽  
Nuclear Industry ◽  
Normal Operation ◽  
Severe Accidents ◽  
Recent Developments ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Directed Research

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed.

scholarly journals ANALYTICAL SIMULATION OF A PWR USING MELCOR

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Maritza Rodríguez Gual ◽  
Nathalia N. Araújo ◽  
Marcos C. Maturana
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Normal Operation ◽  
Detailed Knowledge ◽  
Pressurized Water ◽  
Severe Accidents ◽  
Reference Plant

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

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.

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.

scholarly journals The EC MUSA Project on Management and Uncertainty of Severe Accidents: Main Pillars and Status

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4473
Author(s):  
Luis Enrique Herranz ◽  
Sara Beck ◽  
Victor Hugo Sánchez-Espinoza ◽  
Fulvio Mascari ◽  
Stephan Brumm ◽  
...  
Keyword(s):  
Source Term ◽  
Initial Boundary ◽  
Severe Accident ◽  
First Year ◽  
Model Parameters ◽  
Spent Fuel ◽  
Horizon 2020 ◽  
Severe Accidents ◽  
Current State ◽  
Accident Scenarios

In the current state of maturity of severe accident codes, the time has come to foster the systematic application of Best Estimate Plus Uncertainties (BEPU) in this domain. The overall objective of the HORIZON-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” is to quantify the uncertainties of severe accident codes (e.g., ASTEC, MAAP, MELCOR, and AC2) when modeling reactor and spent fuel pools accident scenarios of Gen II and Gen III reactor designs for the prediction of the radiological source term. To do so, different Uncertainty Quantification (UQ) methodologies are to be used for the uncertainty and sensitivity analysis. Innovative AM measures will be considered in performing these UQ analyses, in addition to initial/boundary conditions and model parameters, to assess their impact on the source term prediction. This paper synthesizes the major pillars and the overall structure of the MUSA project, as well as the expectations and the progress made over the first year and a half of operation.

scholarly journals Results of Structural Analysis of WWER-1000/320 Containment Behavior under Severe Accidents

2020 ◽  
pp. 18-30
Author(s):  
L. Liashenko ◽  
A. Panchenko ◽  
O. Shugailo ◽  
M. Koliada
Keyword(s):  
Strain State ◽  
Cylindrical Part ◽  
Severe Accident ◽  
Normal Operation ◽  
Stress And Strain ◽  
Severe Accidents ◽  
Design Basis ◽  
Calculation Results ◽  
Design Basis Accident ◽  
Maximum Design

The paper presents the review and evaluation of the containment prestressing system within reinforced concrete structures under seismic loads and severe accidents. Given the complex design of the containment, the detailed finite element model has been developed and used to describe real containment behavior. Containment stress and strain state was calculated by modern LIRA software. The first stage analyzed the results of WWER-1000/320 containment stress and strain state calculation under a combination of loads caused by maximum design basis accident (MDBA) and safe shutdown earthquake (SSE) and defined minimum acceptable tension of tendons. The research determines the minimum acceptable tension of tendons in the containment prestressing system, and evaluates the strength and reliability of containment structures under a combination of loads in normal operation + design-basis accident + maximum design earthquake (NO + DBA + MDE). The verification calculations have been performed using tendon tension of 780 ton-force in the cylindrical part of the containment and 760 ton-force in the containment dome. The second stage covered the analysis of severe accident parameters (pressure and temperature) and the results of calculation. Stress and strain state in ZNPP-1 containment has been calculated, parameters (pressure and temperature) under which the containment can loss its protective and isolation functions have been identified, calculation results have been analysed and conclusions of containment structural integrity and ensuring the implementation of the design confining functions have been made. Based on the calculation results, it can be concluded that strength of the containment cylindrical part during a beyond design-basis accident cannot be ensured under parameters t (temperature) = 120°С, p (pressure) = 0.6 MPa.

Offshore Small Modular Reactor (OSMR): An Innovative Plant Design for Societally Acceptable and Economically Attractive Nuclear Energy in a Post-Fukushima, Post-9/11 World

2014 ◽  
Author(s):  
Jacopo Buongiorno ◽  
Michael Golay ◽  
Neil Todreas ◽  
Angelo Briccetti ◽  
Jake Jurewicz ◽  
...  
Keyword(s):  
Transmission Line ◽  
Deep Water ◽  
Heat Sink ◽  
Nuclear Energy ◽  
Severe Accident ◽  
Plant Design ◽  
Cooling Systems ◽  
Physical Protection ◽  
Severe Accidents ◽  
Small Modular Reactor

The potential for major gains in safety, physical protection and economics of nuclear energy exists through the development of a floating Offshore Small Modular Reactor (OSMR). This is a plant that can be entirely built (and decommissioned) as a floating rig in a shipyard, floated to the operating site (within 8–15 km of the coast), anchored in relatively deep water (i.e., ∼100 m), and connected to the grid via an underwater transmission line. The OSMR design presented here features innovative passive and indefinite emergency core and containment cooling systems that eliminate the loss of ultimate heat sink accident, thus decreasing the likelihood of severe accidents. Furthermore, the OSMR containment design and back-up venting procedures effectively eliminate the threat of serious land contamination, should a severe accident actually occur.

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.

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