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.

Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident

2009 ◽  
Author(s):  
S. V. Tsaun ◽  
V. V. Bezlepkin ◽  
A. E. Kiselev ◽  
I. A. Potapov ◽  
V. F. Strizhov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Loss Of Coolant ◽  
Mass Activity

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.

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.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

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.

iB1350: Part 2 — Level1 PRA Considering Optimization of Safety Systems for the iB1350

2018 ◽  
Author(s):  
Tanaka Go ◽  
Sato Takashi ◽  
Komori Yuji ◽  
Matsumoto Keiji
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Lessons Learned ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Active Safety ◽  
Decay Heat ◽  
Active Safety Systems ◽  
Fukushima Daiichi ◽  
Safety Systems

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.

Development of Inherently Safe Technologies for Large Scale BWRs: (4) Hydrogen Explosion Prevention System Using SiC Fuel Claddings

2014 ◽  
Author(s):  
Ryo Ishibashi ◽  
Tomohiko Ikegawa ◽  
Kenji Noshita ◽  
Kazuaki Kitou ◽  
Mamoru Kamoshida
Keyword(s):  
High Temperature ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Hydrogen Gas ◽  
Metallic Layer ◽  
Normal Operation ◽  
Air Cooling ◽  
Composite Substrate ◽  
Prevention System ◽  
High Temperature Steam

In the aftermath of the lessons learned from the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), including a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents. One of inherently safe technologies currently under development is a system to prevent hydrogen explosion during severe accidents (SAs). This hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding of silicon carbide (SiC), and a passive autocatalytic recombiner (PAR). Replacing the zircaloy (Zry) claddings currently used in LWRs with the SiC claddings decreases the hydrogen generation and thus decreases the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B) such as an operation floor. The PAR recombines the leaked hydrogen gas so as to maintain the hydrogen concentration at less than the explosion limit of 4 % in the R/B. The advantages of using SiC claddings in the system were examined through experiments and SA analysis. Results of steam oxidation tests confirmed that SiC was estimated to show 2 to 3 orders of magnitude lower hydrogen generation rates during oxidation in a high temperature steam environment than Zry. Results of SA analysis showed that the total amount of hydrogen generation from fuels was reduced to one fifth or less. Calculation also showed that the lower heat of the oxidation reaction of SiC moderated the steep generation with the temperature increase. We expected this moderated steep generation to reduce the pressure increase in the PCV as well as prevent excess amounts of leaked hydrogen from hydrogen disposal rate capacity using PARs. The SiC cladding under consideration consists of an inner metallic layer, a SiC/SiC composite substrate, and an outer environment barrier coating (EBC). A thin inner metallic layer in combination with a SiC/SiC composite substrate functions as a barrier for fission products. EBC is introduced to have both corrosion resistance in high temperature water environments during normal operation and oxidation resistance in high temperature steam environments during SA. Further reduction of the hydrogen generation rate in high temperature steam by improving the EBC is expected to decrease the total amount of hydrogen generation even more.

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 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.

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.

Analysis of Accident Progression of Fukushima Daiichi NPP by the SAMPSON Code

2014 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hiroaki Suzuki ◽  
Hideo Mizouchi ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Water Injection ◽  
Heat Removal ◽  
Fission Products ◽  
Severe Accident ◽  
Core Material ◽  
Decay Heat ◽  
The Core ◽  
Fukushima Daiichi ◽  
Fire Engine

This paper describes analysis results of the early phase accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Unit 1 by the severe accident analysis code SAMPSON. The isolation condensers were the only devices for decay heat removal at Unit 1, but they stopped after the loss of AC and DC powers. Since there were no decay heat removal for about 14 hours after their termination until the start of alternative water injection into the core by the fire engine, the core melt and the reactor pressure vessel (RPV) bottom failure occurred resulting in large amount of fission products release into the environment. The original SAMPSON was improved by adding new modellings for the phenomena which have been deemed specific to the Fukushima Daiichi NPP: (1) deterioration of SRV gaskets and (2) buckling of in-core-monitor housings which caused the early steam leakage from the core into the drywell, and (3) melt of the in-core-monitor housings in the lower plenum of the RPV. The analysis results showed that (1) 55.3% of UO2 of the initial loading and 66.1% of the core material including UO2, zircaloy, steel and control materials had melted down into the pedestal of the drywell, (2) the amount of Hydrogen generated by Zr-H2O reaction was 686 kg, (3) amount of Cs element released from fuels was 61 kg which was 72% of the total Cs element which was included in fuels at the initiation of the accident, and (4) 18.3% of the corium which fell into the pedestal was one large lump and the 81.7% was particulate corium.

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