Study on Technical Scheme and Frame of Extreme Hazards Mitigation Guideline for CPR1000 NPP

2017 ◽  
Author(s):  
Juanhua Zhang ◽  
Shouyang Zhai ◽  
YeHong Liao
Keyword(s):  
Large Scale ◽  
Command And Control ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Water Storage Tank ◽  
Extreme Hazards ◽  
And Control ◽  
Extensive Damage

The extreme hazards such as large floods, earthquakes, fires or explosion may be lead to an extensive damage or large scale incident in NPP, for instance, loss of command and control, loss of most instruments or equipment which are used for accident mitigation. The typical situations are loss of control room and alternate shutdown capabilities, or loss of AC and DC power, or all of them. Extreme Hazards Mitigation Guideline (EHMG) is developed as the coping strategy for the above extensive damage situations according to the requirements of the characteristic of CPR1000 NPP and the modifications implemented after Fukushima accident. The technical scheme and frame of EHMG includes two modules: EH-IRs (initial response) and EH-MGs (mitigation guideline). And then based on the typical accident analysis, the critical diagnostic parameters and mitigation strategies were made for the EHMG. EH-IRs contains the following contents: Rebuilding the command and control, Off-Site and On-Site Communications, Initial Operational Response Actions, Initial Damage Assessment. EH-MGs contains the following strategies: Manually operating auxiliary turbo pump, Manually depressurizing SGs and injecting into SG, Alternate makeup to RCS, Controlling containment conditions, Alternate Makeup to Spent fuel pool, Alternate Makeup to water storage tank, Containment flooding and so on. The strategies in EHMG can cover both prevention and mitigation phases of severe accident. The first EHMG in China has been validated and implemented in Hong Yan He NPP in September, 2016.

Study on Hydrogen Risk of Spent Fuel Compartment Induced by Containment Venting

2017 ◽  
Author(s):  
Yabing Li ◽  
Xuewu Cao
Keyword(s):  
Control System ◽  
Heat Removal ◽  
Compartment Model ◽  
Severe Accident ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Spent Fuel Pool ◽  
Heat Removal System ◽  
And Control ◽  
Removal System

Hydrogen risk in the spent fuel compartment becomes a matter of concern after the Fukushima accident. However, researches are mainly focused on the hydrogen generated by spent fuels due to lack of cooling. As a severe accident management strategy, one of the containment venting paths is to vent the containment through the normal residual heat removal system (RNS) to the spent fuel compartment, which will cause hydrogen build up in it. Therefore, the hydrogen risk induced by containment venting for the spent fuel compartment is studied for advanced passive PWR in this paper. The spent fuel pool compartment model is built and analyzed with integral accident analysis code couple with the containment analysis. Hydrogen risk in the spent fuel pool compartment is evaluated combining with containment venting. Since the containment venting is mainly implemented in two different strategies, containment depressurization and control hydrogen flammability, these two strategies are analyzed in this paper to evaluated the hydrogen risk in the spent fuel compartment. Result shows that there will not be significate hydrogen built up with the hydrogen control system available in the containment. However, if the hydrogen control system is not available, venting into the spent fuel pool compartment will cause a certain level of hydrogen risk there. Besides, suggestions are made for containment venting strategy considering hydrogen risk in spent fuel pool compartment.

Spent Fuel Pool Under Severe Accident Conditions

2014 ◽  
Author(s):  
Bernd Jaeckel ◽  
Jonathan Birchley ◽  
Leticia Fernandez-Moguel
Keyword(s):  
Large Scale ◽  
Natural Circulation ◽  
Severe Accident ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Flow Parameters ◽  
Loss Of Coolant ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Accident Conditions

The possibility of a spent fuel severe accident has received increasing attention in the last decade, and in particular following the Fukushima accident. Several large scale experiments and also separate effect tests have been conducted to obtain a data base for model development and code validation. The outcome of the Sandia BWR Fuel Project was used to define the flow parameters adjusted for the low pressure and the increased flow resistance due to the presence of the spent fuel racks which resulted in reduced buoyancy driven natural circulation flow compared with reactor geometry. The possibility of a zirconium fire, using the flow parameters obtained from the spent fuel experiments, is investigated in the present work. The important outcome of the study is that spent fuel will degrade if temperatures above 800 K are reached. In partial loss of coolant accidents the flow through the lower bottom nozzle is blocked and because there is no cross flow possible due to the spent fuel racks the coolant flow in the upper dry part of the spent fuel is limited by the steam production in the lower still wetted part of the fuel. This accident scenario leads to the fastest heat up in a postulated spent fuel accident. The influence of different kind of spent fuel storage (hot neighbour and cold neighbour) is investigated. An important factor in these calculations is the radial heat transfer to the neighbouring fuel assemblies. In the present work limits of the spent fuel storage under accident conditions (minimum allowed water levelin the pool) and total loss of coolant (maximum coolable decay heat per fuel assembly) are shown and explained.

scholarly journals Safest Roadmap for Corium Experimental Research in Europe

2017 ◽  
Vol 4 (3) ◽  
Author(s):  
Christophe Journeau ◽  
Viviane Bouyer ◽  
Nathalie Cassiaut-Louis ◽  
Pascal Fouquart ◽  
Pascal Piluso ◽  
...  
Keyword(s):  
Experimental Research ◽  
Nuclear Power ◽  
Large Scale ◽  
Reactor Core ◽  
Severe Accident ◽  
Material System ◽  
Stress Tests ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Spent Fuel Pool

Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) severe accident (SA) with reactor core melting and formation of hazardous material system known as corium. The main objective of the project is to establish coordinated activities, enabling the development of a common vision and severe accident research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on severe accident experimental research has been developed to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. The roadmap takes into account different SA phenomena and issues identified and prioritized in the analyses of severe accidents at commercial NPPs and in the results of the recent European stress tests carried out after the Fukushima accident. Nineteen relevant issues related to reactor core meltdown accidents have been selected during these efforts. These issues have been compared to a survey of the European SA research experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. The comparison shows certain important lacks in SA research infrastructures in Europe, especially in the domains of core late reflooding impact on source term, reactor pressure vessel failure and molten core release modes, spent fuel pool (SFP) accidents, as well as the need for a large-scale experimental facility operating with up to 500 kg of chemically prototypic corium melt.

scholarly journals Leading in a VUCA World

2021 ◽  
Vol 20 (1) ◽  
pp. 89-111
Author(s):  
R Ramakrishnan
Keyword(s):  
Critical Period ◽  
Large Scale ◽  
Command And Control ◽  
Insufficient Information ◽  
Control Structures ◽  
The Future ◽  
Entire World ◽  
To Come ◽  
And Control ◽  
Crisis Leaders

The current COVID-19 virus has put the entire world in lockdown, creating one of the worst times of a VUCA world. The changes that are happening because of the pandemic are large scale and occur suddenly. There is a shortage of leadership everywhere. Leaders are unprepared to lead effectively. In this fast-changing and disruptive environment, command and control structures fail. Leaders are expected to act on incomplete or insufficient information. They do not know where to start to drive change as increased complexity makes it difficult. Leaders lack time to reflect and end up acting too quickly or acting too late as they get stuck in analysis paralysis. They are far removed from the source and are forced to act with a limited understanding of events and their meanings. The role and type of leadership are being tested as we are trying to come out of this crisis. Leaders cannot predict the future but need to make sense of it in order to thrive. This paper would analyse challenges that are being faced by leaders in this critical period and how these can be converted into opportunities like a vaccine for the virus.

Assessment of Radiological Source Term Releases for Potential Severe Accident Scenarios in a PWR SFP

2018 ◽  
Author(s):  
Kwang-Il Ahn ◽  
Jae-Uk Shin
Keyword(s):  
Water Injection ◽  
Severe Accident ◽  
Base Case ◽  
Source Terms ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Makeup Water ◽  
Accident Scenarios ◽  
The Impact

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.

The Importance of Reliable Indications for Accident Management

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
Bill T. Williamson
Keyword(s):  
Severe Accident ◽  
Spent Fuel ◽  
Critical Importance ◽  
Dc Power ◽  
The Status ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
And Control ◽  
Parameter Values ◽  
Support Guidelines

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. There is evidence that the failure of key instrumentation to provide reliable information to the control room licensed operators contributed to the severity of the accident at both TMI and Fuskushima Daiichi. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data and yet have to make urgent decisions. While progress in these areas has been made since TMI-2, the accident at Fukushima suggests there may still be some potential for further improvement in critical plant instrumentation. As a result, several approaches are being employed to provide better information to emergency response personnel during a severe accident. The first approach being taken by the PWROG and BWROG is the identification of methods to obtain information related to key plant parameters when there is a loss of dc power for instrumentation and control. The FLEX guidance in NEI 12-06 requires that reliable instrumentation be available to ensure core, containment and spent fuel pool cooling is maintained for the beyond design basis events for which FLEX was intended. For the most part, this instrumentation that is important for FLEX is the same instrumentation that is used for diagnosis of severe accident conditions and challenges to fission product barriers. Generic FLEX Support Guidelines have been developed to provide a uniform basis for plants to meet the NEI 12-06 requirements that includes methods to obtain key parameter values in the event of a loss of all dc instrument power. The PWROG and the BWROG have also taken a complimentary approach to provide Technical Support Guidance (TSG) for instrumentation during a severe accident. This approach identifies the primary instrumentation as well as alternate instrumentation and other tools to validate the indications from the primary instrumentation. The validation consists of: a) comparing the primary instrument indications to the alternate instrumentation, b) comparing instrument indications to related instrumentation, c) comparing instrument indications and trends to expected trends based on the accident progression and actions already implemented, and d) comparing instrument indications to information in calculational aids.

scholarly journals ESFR SMART PROJECT CONCEPTUAL DESIGN OF IN-VESSEL CORE CATCHER

2021 ◽  
Vol 247 ◽  
pp. 01002
Author(s):  
Joel Guidez ◽  
Antoine Gerschenfeld ◽  
Janos Bodi ◽  
Konstantin Mikityuk ◽  
Francisco Alvarez-Velarde ◽  
...  
Keyword(s):  
Natural Convection ◽  
Fast Reactor ◽  
Passive Control ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Fukushima Accident ◽  
The Core ◽  
Core Catcher ◽  
Thermal Calculations ◽  
Core Meltdown

Even before Fukushima accident occurred, the safety authorities have required that new power plant designs must take into account beyond design-basis accidents including possible core meltdown. Among the mitigation strategies, the corium retention must be ensured, so a core catcher is implemented in the design of the Generation IV Sodium-cooled Fast Reactor. An internal core catcher within the vessel (in-vessel retention) is the option chosen for the European Sodium-cooled Fast Reactor investigated in the H2020 ESFR-SMART project. The new core investigated in ESFR SMART with lower void effect has a better behavior in case of severe accident. The use of passive control rods is also an improvement for prevention of severe accident. Moreover, we have in the ESFR SMART core dedicated tubes for corium discharge that should allow discharging quickly the melted materials and should help to prevent large criticality. Calculations show that after several seconds, these discharge tubes begin to open, and the corium arrives by this preferential way on the core catcher, quicker and in limited quantities at the beginning of the accident. However, the core catcher is designed to be able to retain the whole core meltdown. Its design allows good possibilities of cooling by natural convection of sodium. Some thermal calculations were provided with a multi-layer concept but the global mechanical conception seems difficult. So a one layer core catcher in molybdenum, material compatible with sodium and used on the core catcher of the last SFR, started in 2016: BN 800, is investigated. Explanations are given on the choice of this material proposed for the catcher and used for thermal calculations. With the proposed design, the corium is spread on the core catcher and the residual power of the corium can be dispelled by natural convection by the sodium circulating around and above the core catcher without boiling of sodium if the melted core is less than about 25% of whole core. In case of bigger quantities of melted core, boiling of sodium could appear under the core catcher. Further less conservative calculations would be necessary to better know the limit.

Command and Control for Large-Scale Hybrid Warfare Systems

2015 ◽  
Vol 03 (01) ◽  
pp. 1-15 ◽  
Author(s):  
Chee Khiang Pang ◽  
Gregory R. Hudas ◽  
Dariusz G. Mikulski ◽  
Cao Vinh Le ◽  
Frank L. Lewis
Keyword(s):  
Real Time ◽  
Large Scale ◽  
Discrete Event ◽  
Command And Control ◽  
Armed Forces ◽  
Networked Computing ◽  
Discrete Event Control ◽  
Team Decision ◽  
And Control ◽  
Multiple Missions

Emerging hybrid threats in large-scale warfare systems require networked teams to perform in a reliable manner under changing mission tactics and reconfiguration of mission tasks and force resources. In this paper, a formal Command and Control (C2) structure is presented that allows for computer-aided execution of the networked team decision-making process, real-time tactic selection, and reliable mission reconfiguration. A mathematically justified networked computing environment is provided called the Augmented Discrete Event Control (ADEC) framework. ADEC is portable and has the ability to provide logical connectivity among all team participants including mission commander, field commanders, war-fighters, and robotic platforms. The proposed C2 structure is developed and demonstrated on a simulation study involving Singapore Armed Forces team with three realistic symmetrical, asymmetrical, and hybrid attack missions. Extensive simulation results show that the tasks and resources of multiple missions are fairly sequenced, mission tactics are correctly selected, and missions and resources are reliably reconfigured in real time.

scholarly journals SAFEST Roadmap for Corium Experimental Research in Europe

2016 ◽  
Author(s):  
Christophe Journeau ◽  
Viviane Bouyer ◽  
Nathalie Cassiaut-Louis ◽  
Pascal Fouquart ◽  
Pascal Piluso ◽  
...  
Keyword(s):  
Experimental Research ◽  
Research Priorities ◽  
Large Mass ◽  
Severe Accident ◽  
Stress Tests ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Global Vision ◽  
Spent Fuel Pool ◽  
Experimental Facilities

SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on corium experimental research has been written to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. It is based on the research priorities determined by SARNET SARP group as well as those from the recently formulated in the NUGENIA Roadmap for severe accidents and the recently published NUGENIA Global Vision report. It also takes into account issues identified in the analysis of the European stress tests and from the interpretation of the Fukushima accident. 19 relevant issues related to corium have been selected during these prioritization efforts. These issues have been compared to a survey of the European corium experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. It shows a few lacks in EU corium infrastructures, especially in the domains of core late reflooding impact on source term, Reactor Pressure Vessel failure and corium release, Spent Fuel Pool accidents, as well as the need for a large mass (100–500 kg) prototypic corium facility.

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