Hydrogen Deflagration Analysis in THAI Experiments Using MELCOR Code

2020 ◽  
Author(s):  
Alain Flores y Flores ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Control Volume ◽  
Critical Issue ◽  
Severe Accident ◽  
Test Facility ◽  
Research Centre ◽  
Time Step ◽  
Volume Number ◽  
Fukushima Daiichi

Abstract In order to develop an appropriate knowledge to support the SUJB (State Office of Nuclear Safety), the CVR (Research Centre Rež), in collaboration with SURO (National Radiation Protection Institute) is developing a methodology to simulate nuclear power plants under accidental conditions. A particular effort is focused in the severe accident phenomenology where hydrogen deflagration carries a critical issue for the containment integrity, such as Fukushima Daiichi accident. For this purpose, THAI (Thermal-hydraulics, hydrogen, aerosol and iodine) experimental campaigns are chosen due to the several tests involved in different conditions. THAI containment test facility is used to open questions concerning the behaviour of hydrogen, iodine and aerosols in the containment of water-cooled reactors during severe accidents. The Fukushima Daiichi Accident demonstrates that the hydrogen deflagration could lead to a significant containment damage. For this reason, a particular attention is given to the hydrogen deflagration scenario. All simulations are prepared and modelled in MELCOR 2.1. The results obtained showed a strong influence related with some factors as: the nodalization pattern, control volume number (CV), flow paths number FP and time step. In order to assess the THAI model with the THAI final reports, a sensitivity analysis focused with those parameters was performed.

Causes and Radiological Consequences of the Chernobyl and Fukushima Nuclear Accidents

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
L. Sihver ◽  
N. Yasuda
Keyword(s):  
Local Time ◽  
Emergency Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Nuclear Accidents ◽  
Emergency Response System ◽  
Response System ◽  
Accident Management ◽  
Fukushima Daiichi

In this paper, the causes and the radiological consequences of the explosion of the Chernobyl reactor occurred at 1:23 a.m. (local time) on Apr. 26, 1986, and of the Fukushima Daiichi nuclear disaster following the huge Tsunami caused by the Great East Japan earthquake at 2.46 p.m. (local time) on Mar. 11, 2011 are discussed. The need for better severe accident management (SAM), and severe accident management guidelines (SAMGs), are essential in order to increase the safety of the existing and future operating nuclear power plants (NPPs). In addition to that, stress tests should, on a regular basis, be performed to assess whether the NPPs can withstand the effects of natural disasters and man-made failures and actions. The differences in safety preparations at the Chernobyl and Fukushima Daiichi will therefore be presented, as well as recommendations concerning improvements of safety culture, decontamination, and disaster planning. The need for a high-level national emergency response system in case of nuclear accidents will be discussed. The emergency response system should include fast alarms, communication between nuclear power plants, nuclear power authorities and the public people, as well as well-prepared and well-established evacuation plans and evacuation zones. The experiences of disaster planning and the development of a new improved emergency response system in Japan will also be presented together with the training and education program, which have been established to ensure that professional rescue workers, including medical staff, fire fighters, and police, as well as the normal populations including patients, have sufficient knowledge about ionizing radiation and are informed about the meaning of radiation risks and safety.

Severe Accident Context Quantification for Long-Term Station Blackout in Boiling Water Reactor Nuclear Power Plants

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Gueorgui I. Petkov ◽  
Monica Vela-Garcia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Accident Scenario ◽  
Station Blackout ◽  
Fukushima Daiichi

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the performance evaluation of teamwork (PET) procedure for dynamic context quantification and determination of alternatives, coordination, and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions with the use of thermo-hydraulic (TH) model and severe accident (SA) codes (melcor and maap). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and a hypothetic unmitigated LT SBO at peach bottom #1 boiling water reactor (BWR) reactor nuclear power plants (NPPs). The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and TH calculations made by using maap code at the EC Joint Research Centre.

Study of a Gamma Radiation Impact on Concrete Properties Under Severe Accident Conditions

2020 ◽  
Vol 7 (2) ◽  
Author(s):  
Zbyněk Hlaváč ◽  
Jaroslava Zatloukalová ◽  
Michal Košťál ◽  
Evžen Losa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Radiation Effects ◽  
Experimental Studies ◽  
Life Time ◽  
Severe Accident ◽  
Research Centre ◽  
Concrete Properties ◽  
60Co Source ◽  
Operational Life

Abstract Concrete is an important structural material used in nuclear power plant (NPP) design. Due to relatively high amount of hydrogen as well as the presence of heavier elements, it also acts as a biological shielding. One of the important tasks for prolongation of operational life time is the determination of concrete components' condition after long-term irradiation. The paper aims to present the current activities in the CV Řež institute (Research Centre Řež—CVR) regarding the investigation of ionizing radiation effects on concrete properties. In its first part, the paper deals with experimental identification of the character of mixed neutron and gamma spectra in the concrete part of the VVER-1000 Mock-Up. Using the knowledge, the radiation field character can be scaled up to the commercial power plants with VVER-1000 light water reactor. It also provides justification for usage of the 60Co source for performed irradiation experiments with concrete. The second part of the article describes the experimental studies of the properties of gamma-irradiated concrete samples by strong 60Co source. This irradiation experiment can be understood as the first step in characterizing concrete degradation as gamma flux in biological shielding is significantly higher than that of neutron flux. In order to better understand the concrete properties and the behavior under irradiation, nondestructive as well as destructive testing methods were applied. We found that after 48 days of irradiation by the 60Co source the sample obtained dose from gamma corresponding to approximately 1% of the total during the NPP lifetime operation. Concrete microstructure degraded and the modulus of elasticity slightly decreased within 5%. Conversely, destruction tests prove significant flexural strength decrease by 27% in case of normal test and by 63% at the loss of coolant accident (LOCA) test.

Parametric Study of MISTRA M5 Test Using Lumped Parameter COCOSYS Code

2012 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius ◽  
Sigitas Rimkevičius
Keyword(s):  
Parametric Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Safety Issue ◽  
Severe Accident ◽  
Test Facility ◽  
Lumped Parameter ◽  
Simulation Results ◽  
The Impact ◽  
Parameter Approach

Gas distribution in the containments of nuclear power plants in a case of severe accident currently is a high priority safety issue. One of the topics in this issue is a formation and break up of stratified conditions inside the containment. CEA (France) performed M5 test in their MISTRA test facility, to form a stratified atmosphere, which later was destroyed during MASP tests that followed. During M5 there was formed vertically stratified atmosphere using steam jet release. Application of lumped parameter codes for simulation of jet release is complicated due to limitations inherent in the lumped-parameter approach. However, measures exist which can be used to take these limitations into account when using lumped parameter approach. This paper presents simulations and parametric study of M5 test, i.e. only formation of the stratified atmosphere. Presented simulations of the experiment were performed using lumped parameter code COCOSYS. The aim of the work is to investigate the capability of the code to simulate correctly jet release and formation of the stratified atmosphere in the M5 experiment and the impact of several parameters to the simulation results.

Severe Accident Context Evaluation for BWR NPPS Based on Long-Term Station Blackout

2016 ◽  
Author(s):  
Gueorgui I. Petkov ◽  
Monica Vela-Garcia
Keyword(s):  
Safety Assessment ◽  
Power Plants ◽  
Human Performance ◽  
Severe Accident ◽  
Research Centre ◽  
Joint Research ◽  
Accident Scenario ◽  
Station Blackout ◽  
Fukushima Daiichi

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the Performance Evaluation of Teamwork procedure for dynamic context quantification and determination of alternatives, coordination and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions (timelines) with the use of thermo-hydraulic model and severe accident codes (MELCOR and MAAP). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and an hypothetic unmitigated LT SBO at Peach Bottom #1 Boiling Water Reactor Nuclear Power Plants. The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and thermo-hydraulic calculations made by using MAAP code at the EC Joint Research Centre, Institute for Energy and Transport, Nuclear Reactor Safety Assessment Unit.

Countermeasures to Improve LWR Safety Level in Design and Operation by Accident Analyses at Fukushima Daiichi Nuclear Power Plants

2013 ◽  
Author(s):  
Kazumasa Shimizu ◽  
Yuhei Hamada ◽  
Hiroto Sakashita ◽  
Michitsugu Mori
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Critical Role ◽  
Severe Accident ◽  
Simulation Code ◽  
Safety Level ◽  
Plant Operations ◽  
Fukushima Daiichi ◽  
Water Reactors

The 2011 off the Pacific coast of Tohoku Earthquake occurred on March 11. The earthquake attacked the Fukushima Daiichi nuclear power station with six boiling water reactors (BWRs), three out of which, units 1 through 3 in rated operation except for three reactors of units 4 through 6 in scheduled periodic inspection, automatically shut down in response to the intense seismic motion. Emergency diesel generators started to pump water to cool reactors, and an hour later, the back-up generators lost their all functions by the station blackout resulting from tsunami flooding. In this situation at the unit 1, the isolation condenser system (IC) should have made a critical role to keep the reactor pressure and water level to be safety by removing the decay heat by natural circulation. In fact at the unit 1 during the accident, IC valves were closed by fail-safe and could not have shown the ability of the designed function. An accident report gave general descriptions of the causes and results of accidents, but not the quantitative data indicative of details; therefore, it seems difficult to identify the specific problems in plant operations. Even in this case, if an appropriate analysis code is available for reproducing events based on the reports, it will be possible to determine individual data quantitatively and identify problems in plant operations. In our work, we used the nuclear reactor thermal-hydraulic code RETRAN-3D/MOD4, which has been approved and licensed by U.S. Nuclear Regulatory Commission, to model light water reactors (LWRs) and reproduce the circumstances of the 2011 Fukushima Daiichi nuclear accident as the simulation code. Here, we subjected transition analyses of the process on the core-meltdown accident, and put forward the system to prevent the accident, where the accident analysis report was employed to simulate conditions of the accident. It could enable us to suggest adequate operation procedures suitable for LWR to avoid the severe accident, and to propose countermeasures to improve LWR safety level in design and operation.

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

scholarly journals A decision-support approach to severe accident management in nuclear power plants

2020 ◽  
pp. 1-12
Author(s):  
Marko Bohanec ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
Klemen Debelak ◽  
Luka Štrubelj
Keyword(s):  
Decision Support ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management
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