Unmet Challenges to Successfully Mitigating Severe Accidents in Multi Unit CANDU Reactors

2020 ◽  
Author(s):  
Sunil Nijhawan
Keyword(s):  
Safety Culture ◽  
Severe Accident ◽  
Reactor Safety ◽  
Candu Reactors ◽  
Safety Critical ◽  
Critical Design ◽  
Nuclear Disaster ◽  
Severe Accidents ◽  
Related Risk ◽  
Safety Systems

Abstract One sees eerie similarities here in Canada to the cozy relationship between regulator and utilities in ‘pre-Fukushima’ Japan. Such ties are hardly limited to Canada though. The chronic degradation of real commitments to continued improvements in reactor safety systems and a decline in overall safety culture that discourages critical design reviews and willfully ignores well justified, safety critical hardware upgrades, has created alarming conditions that are likely inching us towards another nuclear disaster. Operating CANDU reactors are now close to being obsolete but have barely seen any substantive severe accident related risk reduction upgrades nine years after Fukushima, hoopla in Canada around some minor improvements and premature closure of even otherwise sparse and what were really weak regulatory ‘Fukushima Action Items’ notwithstanding.

Characteristics of Pressure Buildup From Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool

2014 ◽  
Author(s):  
Songbai Cheng ◽  
Ken-ichi Matsuba ◽  
Mikio Isozaki ◽  
Kenji Kamiyama ◽  
Tohru Suzuki ◽  
...  
Keyword(s):  
Severe Accident ◽  
Water Volume ◽  
Reactor Safety ◽  
Experimental Conditions ◽  
Melt Temperature ◽  
Pressure Buildup ◽  
Severe Accidents ◽  
Series Of Experiments ◽  
Boiling Mode

Studies on local fuel-coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). To clarify the mechanisms underlying this interaction, in this study a series of experiments was conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Based on the experimental data obtained from a variety of conditions, including difference in water volume, melt temperature and water subcooling, the characteristics of pressure-buildup during local FCIs was investigated. It is found that under our experimental conditions the water volume and melt temperature have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The performed analyses also suggest that the pressurization from local FCIs should be intrinsically limited, due to a suppressing role caused by the increasing of coolant volume entrapped within the pool as well as the transition of boiling mode. Current work, which gives a palette of favorable data for a better understanding and an improved estimation of severe accidents in SFRs, is expected to benefit future analyses and verifications of computer models developed in advanced fast reactor safety analysis codes.

Three Types of a Passive Safety Containment for a Near Future BWR With Active and Passive Safety Systems

2010 ◽  
Author(s):  
Takashi Sato ◽  
Makoto Akinaga ◽  
Yoshihiro Kojima ◽  
Tsunekazu Murakami ◽  
Kenji Hosomi
Keyword(s):  
Severe Accident ◽  
Passive Safety ◽  
Large Power ◽  
Gas Treatment ◽  
The Core ◽  
Containment Vessel ◽  
Passive Safety Systems ◽  
Severe Accidents ◽  
Safety Systems ◽  
Near Future

The paper presents three types of a passive safety containment for a near future BWR. They are tentatively named Mark S+, Mark D and Mark X containments in the paper. They all have a leak tight secondary containment vessel (SCV) in order to meet the reactor site criteria without relying on an active standby gas treatment system at a DBA LOCA. One of their common features is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The containment pressure can be limited within the design pressure. Even if a large amount of hydrogen is generated at a severe accident, it can be released into the SCV. Hydrogen detonation or deflagration is completely prevented without using igniters. Another feature is the capability to submerge the PCV and the RPV above the core level without relying on accident management. The core debris is completely submerged not only ex-vessel but also in-vessel. The third feature is robustness against external events such as a large commercial airplane crash. All the containments have built-in passive safety systems (BIPSS) including a passive containment cooling system (PCCS) and a passive cooling core catcher that has radial cooling channels. The Mark S+ and Mark D containments are applicable to a large power BWR up to 1830 MWe. The SCV is made of steel-concrete composite. The PCV can be vented into the inerted part of the SCV at a severe accident. The Mark X containment has the steel secondary containment vessel (SSCV) and can be cooled by natural convection of outside air. It can accommodate a medium power BWR up to about 1000 MWe and has a permanent grace period without replenishing the PCCS pool. In all cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides in-depth protection against severe accidents and also enables N+2 design. All the three containments coupled with the IDHS can potentially provide an evacuation free plant at a severe accident caused by severe natural disasters such as a giant earthquake, a tsunami, a mega hurricane, and so on.

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.

Enhancement of Plant Availability by Outage Shortening in European Advanced Boiling Water Reactor (EU-ABWR)

2013 ◽  
Author(s):  
Kazuhiro Kamei ◽  
Kazuyoshi Kataoka ◽  
Kazuto Imasaki ◽  
Noboru Saito
Keyword(s):  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Plant Availability ◽  
Control Rod ◽  
Water Reactor ◽  
Spray System ◽  
Maintenance And Inspection ◽  
Reactor Pool ◽  
Safety Systems

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the requirements in the Finnish utility and the safety requirements of Finnish YVL guide. By adopting Scandinavian outage processes, the Plant Availability is aimed to be greater than 95%. ABWRs have an excellent design potential to acheive short outage duration (e.g., shortening of maintenance and inspection duration by applying Fine Motion Control Rod Drive and Reactor Internal Pump). In addition, the EU-ABWR applies following key design improvements to reduce a refueling outage duration; a) Direct Reactor Pressure Vessel (RPV) Head Spray System, b) Self-standing Control Rods and c) Water shielding reactor pool. In this paper, coolability of RPV due to application of the Direct RPV Head Spray System is also verified with numerical evaluations by Computation Fluid Dynamics (CFD) analysis.

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.

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.

scholarly journals Analysis of humanoid robotics for nuclear disaster management incorporated with biomechanics

2019 ◽  
Vol 34 (3) ◽  
pp. 291-298
Author(s):  
Kyung Jang ◽  
Tae Woo
Keyword(s):  
Reactor Core ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Physical Aspect ◽  
Multidisciplinary Research ◽  
Nuclear Disaster ◽  
Collapsed Building ◽  
Very High ◽  
Mechanical Studies

The humanoid is investigated for the mechanical and physical aspect in the nuclear disaster, especially for a severe accident, which includes the core melting. There are some mechanical studies of the leg and hand of the humanoid in which the human mimicking features are described. The management of the task is accomplished by the three regional preparations. The robot is made of the radiation-resistance substance. Therefore, it could work on the normal task of a human for the removal of the broken debris in a collapsed building. However, there is a limitation for the use in the reactor core building due to very high temperature of the nuclear fuel. The regional classification of the site is studied for the practical purposes. The post-accident analysis is accompanied with multidisciplinary research for the humanoid development in the nuclear industry.

scholarly journals Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS injection line using MELCOR code.

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Seung Min Lee ◽  
Nelbia Da Silva Lapa ◽  
Gaianê Sabundjian
Keyword(s):  
Cooling System ◽  
Severe Accident ◽  
Time Intervals ◽  
Severe Accidents ◽  
Independent Analysis ◽  
Accident Management ◽  
Injection Line ◽  
Core Cooling ◽  
Control Volumes ◽  
Physical Phenomena

The aim of this work was to simulate a severe accident at a typical PWR, initiated with a break in Emergency Core Cooling System line of a hot leg, using the MELCOR code. The model of this typical PWR was elaborated by the Global Research for Safety and provided to the CNEN for independent analysis of the severe accidents at Angra 2, which is similar to this typical PWR. Although both of them are not identical, the results obtained of that typical PWR may be valuable because of the lack of officially published simulation of severe accident at Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes, after the break at the hot leg, were calculated as well as degree of core degradation and hydrogen production within the containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management by implementing each measure in this model.

scholarly journals Spreading of Excellence in SARNET Network on Severe Accidents: The Education and Training Programme

2012 ◽  
Vol 2012 ◽  
pp. 1-9 ◽  
Author(s):  
Sandro Paci ◽  
Jean-Pierre Van Dorsselaere
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Training Programme ◽  
Education And Training ◽  
Severe Accident ◽  
Research Network ◽  
Safety Issues ◽  
Severe Accidents ◽  
Management Procedures ◽  
And Training

The SARNET2 (severe accidents Research NETwork of Excellence) project started in April 2009 for 4 years in the 7th Framework Programme (FP7) of the European Commission (EC), following a similar first project in FP6. Forty-seven organisations from 24 countries network their capacities of research in the severe accident (SA) field inside SARNET to resolve the most important remaining uncertainties and safety issues on SA in water-cooled nuclear power plants (NPPs). The network includes a large majority of the European actors involved in SA research plus a few non-European relevant ones. The “Education and Training” programme in SARNET is a series of actions foreseen in this network for the “spreading of excellence.” It is focused on raising the competence level of Master and Ph.D. students and young researchers engaged in SA research and on organizing information/training courses for NPP staff or regulatory authorities (but also for researchers) interested in SA management procedures.

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