Severe Accident Management Guidelines Research Articles

Severe accident analysis induced by secondary pipeline break in a small modular PWR

The small modular PWR (SMPWR) usually adopts integral design. Under severe accident, the system responses are different from those large PWRs. It is necessary to study the severe accident behavior of the SMPWR. A MELCOR model is developed for SMPWR and its steady-state results are in good agreement with the design values. Severe accidents induced by secondary pipeline break accidents are simulated, and no pressure relief measures are taken to keep the primary loop under high pressure. The mitigation effects of passive containment air cooling system (PAS) and passive cavity injection system (PCIS) are evaluated under different cases. The results show that under high pressure conditions, PCIS can effectively cool the lower head. The earlier the PCIS operates, the more significant the mitigation effect can be. In addition, PAS can effectively reduce the peak pressure and temperature in the containment. This study can provide a reference for the formulation of severe accident management guidelines on SMPWRs.

Validation of EPG/SAG Rev. 4 of the Kuosheng Nuclear Power Plant Using the MAAP5 Code

The Boiling Water Reactor Owner’s Group released Emergency Procedure Guidelines and Severe Accident Guidelines Revision 4 (EPG/SAG Rev. 4) in 2018. The major improvement to EPG/SAG Rev. 4 was Contingency 1 (Alternate Level/ Pressure Control). Contingency 1 coordinates the reactor pressure vessel (RPV) water level and RPV pressure control action to prolong the availability of steam-driven injections and optimize the transfer to motor-driven systems. In this study, the effectiveness of the EPG/SAG Rev. 4 Contingency 1 strategy was compared with those of EPG/SAG Revision 2 Contingency 1 and Specific Major Incident Guidelines (SMI) using the Modular Accident Analysis Program, Version 5 (MAAP5). SMI was developed by the Taiwan Power Company to mitigate a Fukushima-like accident. The surrogate plant that analyzed is the Kuosheng Nuclear Power Plant (NPP). The Kuosheng NPP is BWR-6 Mark-III containment. MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic risk assessments and for verification and validation of mitigation actions specified in severe accident management guidelines. The simulation scenarios were extended loss of alternative-current power and loss of ultimate heat sink. The low-capacity, motor-driven portable pump was the only available system for RPV injection in the first hour of the accident. In this time period, the RPV water level and pressure were controlled by reactor core isolation cooling and safety relief valves. After this study, the strategy of EPG/SAG Rev. 4 Contingency 1 was successfully validated, and the effectiveness of minimum pre-depressurization RPV water level and the low-capacity, motor-driven portable pump were also demonstrated in this study.

CFD analysis of PHWR exposed core under postulated severe accident condition

A two-dimensional steady-state thermal-hydraulic analyses of the exposed core with varying levels of exposure is carried out for heat transfer assessment. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in Pressurised Heavy Water Reactors (PHWRs). Such exposure of fuel channels leads to their heat-up which is governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer. In order to understand such a complex behaviour as well as to plan the severe accident management guidelines, numerical analyses of the exposed core are required. In this paper, one-of-a-kind thermal-hydraulic analyses of the exposed reactor-scale core geometry are presented for row-wise exposures of fuel channels. These numerical analyses are terminated at the onset of fuel channel disassembly i.e., loss of geometrical intactness of the exposed core. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is brought out in this study. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, for the advanced stages of fuel channel exposures, the heat transfer is observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation arising from Zircaloy and steam interaction. Moreover, it is interesting to note that the disassembly of fuel channels is restricted to central regions of the exposed core.

Key elements for proper development and implementation of severe accident management guidance/guidelines – SAMG

Severe Accident Management Guidelines (SAMG) have been developed to support a nuclear power plant (NPP) in case of a severe accident, i.e. an (unlikely) accident where the core and/or the fuel in the Spent Fuel Pool (SFP) have been damaged and there is risk for radioactive releases. Various approaches exist, such as the SAMG developed by the PWR and BWR Owners Groups (PWROG, BWROG), by Electricité de France (EdF) for its fleet of plants, the Candu Owners Group (COG) and by several individual NPPs. Often, the developers of SAMG develop a generic type of SAMG, applicable for a certain type of NPP, and leave it to the individual plants to adapt these to their configuration and Emergency Response Organisation (ERO). That means, that quite some burden remains for the individual plants to finally implement the SAMG. This implementation includes not only technical items, such as defining proper strategies, but also many organisational issues, as handling a severe accident needs an effective ERO, in which the SAMG must be well embedded.Various plants have been reviewed, both the ways they have constructed their plant-specific SAMG, implemented them in their ERO and tested the performance in drills and exercises. Most recently, a review of SAMG at Swedish plants has been performed, after which it was decided to pull the various experiences together and summarize them in one compendium. Some of these also include aspects of the development of the generic SAMG.The result is a number of recommendations (∼30), both on the technical field as well as regarding the ERO. We could also define a number of good practices that deserve wider attention.A few words have been devoted to international support for a stricken plant (Epilogue).

Influence of guide tubes on configuration and heat-up behaviour for PHWR specific suspended debris bed – An analytical assessment

The estimation of accident source term and hydrogen generation is dependent on the core heat-up under postulated accident scenarios. The uncertainty of such estimations increases as the accident severity increases due to complexity in the determination of i) evolving degraded core configurations, and ii) associated suitable physical models for heat-up and degradation simulation. Analysis of postulated severe accidents for PHWR is essential for emergency planning and severe accident management guidelines from a plant licensing perspective. The study presents the order of uncertainty that is present as of today in deciding the size and shape of the suspended debris bed, which in turn decides the exposed core heat up, hydrogen generation as well as fission product release. The study reveals that the location and integrity of reactivity mechanism guide tubes significantly impact the exposed core heat-up, channel disassembly pattern and shape and size of the suspended debris bed. Nevertheless, the experimental study on the guide tube behaviour on asymmetric heat-up condition and their failure will further substantiate the postulation of suspended debris bed configuration.

Effects of elevated temperature steam oxidation and subsequent quenching on light water reactor clad

Following a postulated severe accident, the temperature of fuel pins of light water reactor escalates because of decay heat and devoid of coolant. The presence of steam inside the core causes exothermic oxidation of the cladding surface at elevated temperatures. The in-vessel injection system of severe accident management guidelines injects the cold water inside the hot uncovered core to arrest the escalation of temperature rise and terminate the accident progression. However, the heated clad may be subjected to thermal shock during the quenching process, which inevitably assists in the failure and fragmentation of the severely oxidised fuel pins column.In this paper, the effect of thermal quenching shock on the severely oxidised fuel pins column during the reflooding of the hot uncovered core is studied experimentally. A 2ˣ2 square lattice arrangement of zircaloy-4 clads is used for experiments. First, the clads are oxidised in the presence of steam in the temperature range of 1100–1250 °C, and then the test-section is flooded with water at ambient temperature from its bottom. The violent heat transfer between the clad and cold water has been observed during the reflooding. Furthermore, the clad has failed with a circumferential crack at multiple locations along the clad length during this period; however, longitudinal cracks are not observed on the clad surface. The scanning electron micrographs of the clad samples show a significant oxide layer over the clad surface. At the same time, energy dispersive x-ray spectroscopy reveals significant oxygen diffusion inside the zirconium metal matrix, particularly near the outer tube surface.Additional experiments are performed to study the clad failure mechanism during the reflooding of the severely oxidised clad. The set of experiments reveals that the clad fails under the thermal quench shock (>7.41 °C/s) for severely oxidised cladding (ECR>25%) at an average temperature above 1100 °C. The failed clad is likely to create debris. The average quench front velocity is 0.6 cm/s while the reflood velocity is 1.13 cm/s. Quench velocity below the reflood velocity indicates excessive water evaporation and dispersed type flow ahead of the quench front. The total analytical H2 generated before reflooding is 3.75 g.

Influence of active and passive equipment for advanced pressurized water reactor on thermal hydraulic and source term behavior in severe accidents

Extensive studies have been carried out on the behavior of core degradation and fission products of common pressurized water reactors (PWRs). However, few of them have investigated the relationship between thermal hydraulic and fission product behavior in advanced passive PWRs. Due to the impact of thermal hydraulic behaviors in different accident sequences on the release and transportation of fission products, an integrated severe accident analysis (ISAA) code with highly coupled thermal hydraulic and source term calculations is required to simultaneously analyze thermal hydraulic and source term behavior. For advanced passive PWRs, important safety systems that may affect the behavior of the core and fission products should be considered. It is therefore necessary to simulate the thermal hydraulic and fission product behavior of advanced passive PWRs. In this study, the ISAA code is adopted to simulate the occurrence of a hypothetical double ended cold leg LBLOCA of HPR1000 in three scenarios of equipment failure. The results show that the high-temperature fuel rods and cladding materials exhibit delayed failure at the lower position of the active core, whereas earlier failure at higher position during the reflooding. Active and passive equipment affects fuel temperature, the oxidation conditions of the fuel, the interaction of fission products and structural materials, and the state of the fuel, thereby affecting the release of fission products in the fuel. HPR1000 only relies on passive equipment to relieve the core degradation in severe accidents, realize the in-vessel retention of melt, and eliminate the ex-vessel release possibility of fission product. It is hoped that the results can provide references for HPR1000 to formulate the severe accident management guidelines (SAMG).

Assessment of Filtered Containment Venting Strategies for Mitigating Severe Accident Consequences at a Generic Single-Unit CANDU

Abstract In many jurisdictions, filtered containment venting systems are an integral part of severe accident management guidelines but are normally only used as a last resort, to be manually activated when the containment pressure approaches the containment failure limit. This approach fails to utilize the capability of the filtered venting system to better control containment pressure throughout an accident and to create capacity for possible future pressurization based on the accident progression. In this paper, smart venting strategies are explored, using a generic single-unit Canada Deuterium Uranium (CANDU) (660 MWel) plant model in MAAP-CANDU, as a means to mitigate the consequences of a severe accident. These strategies are able to account for both current conditions as well as possible future pressurization. The use of strategic venting offers an improvement in maintaining containment integrity, as well as reducing I-131 releases to the environment by about 10% to 31% compared to last-resort venting while being feasible for manual operator actions. Overall, this comprehensive analysis demonstrates the capability of a smart venting strategy paired with a filtered containment venting system as an active mitigation tool.

Accident Management under Extreme Events

Most nuclear power plants have extensive sets of Emergency Operating Procedures and Severe Accident Management Guidelines. These offer protection for a large series of events, both inside and outside the licensed design basis of the plant. For Extreme Events, which are characterised by a large destruction on-site and may include loss of command and control, damage to multiple units on-site, loss of communication both on-site and to off-site centres, staff members wounded or killed, such protection may not be enough. Examples of Extreme Events are air plane crash, site flooding, large earthquake plus possible tsunamis, etc. This paper describes what additional procedures, guidelines, hardware and organisational issues are needed to protect a site against such events. It is based on lessons learned from large destructive events in the past, such as the 9/11 attacks in the USA in 2001 and the tsunami at the Fukushima-Daiichi plants in 2011.

Enhancements to PWR SAMG since Fukushima

Plant specific Severe Accident Management Guidelines are provided to give guidance to plant staff to respond to a severe accident, to protect containment, minimise fission product releases and prioritise equipment recovery efforts while bringing the plant to a controlled stable condition. This paper presents and discusses the enhancements that have been made particularly to widely implemented SAMG based on the Westinghouse Owners Group generic model (now part of the PWR Owners Group) since the Fukushima accident. Although U.S. and European responses to the accident may seem at first sight to be different, in fact, the enhancements needed to SAMG in order to address lessons learned were quite consistent. How to deal with loss of d.c. power and / or instrumentation, long term loss of a.c. power (much longer than had previously been considered), multiple unit severe accidents, accidents in the fuel pool, accidents from shutdown initial plant conditions and other issues and conditions that occurred during the event and which we must learn from? This paper presents results of a number of projects aimed at providing SAMG enhancements to address these issues, some originating in the U.S. and some in Europe. Main enhancements addressing lessons learned from Fukushima are described. The paper also summarises and describes an ongoing project to develop a fully revised generic SAMG specifically applicable to PWR plants in Europe (and also applicable to other designs, such as VVER) which benefits from all these efforts and will provide a single integrated and updated basis for plants with mitigation systems typical in Europe (often different from the US) to update their SAMG.

Research on equipment survivability assessment in severe accidents for CANDU6

For the early CANDU6 design, there are no severe accident mitigation countermeasures, and severe accident environment condition has not been taken into account. Along with developing severe accident management guideline (SAMG) for CANDU6, researchers promoted the establishment of severe accident mitigation countermeasures for CANDU6. In this study a method of equipment survivability assessment for CANDU6 is proposed. Four time frames of equipment survivability in severe accidents for CANDU6 are obtained. The relationship between eight high level actions and time frames is clarified. Environment condition during severe accidents is also included for consequence evaluation. Finally, based on the above analysis, severe accident environment condition for each time frame is compared with equipment qualification condition, some key equipments include valves, local air coolers, hydrogen recombiners, penetrations, instrumentations can be successfully quantified.

Information Requirements of a Decision Support System for Severe Accident Management in Nuclear Power Plants

In nuclear power plants, a severe accident is a critical accident involving significant nuclear core damage and it is managed by using a set of Severe Accident Management Guidelines (SAMG). Prepared as a guideline that provides lists of suggestions rather than strict instructions, SAMG’s contents require frequent decision-making by the operators, causing high cognitive load and creating an error-prone situation that is also amplified by the stressful environment during the severe accident mitigation efforts. A decision support system (DSS), designed by considering the human decision-making process and the system’s holistic view, can help the operators in making informed and appropriate decisions. In this study, we aim to identify the information requirements in designing such DSS for severe accident management of nuclear power plants. We combined two methods: Functional Resonance Analysis Method (FRAM) and decision ladder to identify the information requirements. FRAM provides a systematic analysis of the functions involved in severe accident management and decision ladder captures the human decision-making processes. We developed the FRAM model and the decision ladder model based on SAMG’s contents to identify the set of information requirements. The identified information requirements and their implementation suggestions are provided. This study is the first step in designing a decision support system that considers human cognitive load and holistic system concepts. The method used in this study shall contribute to the design and implementation of a DSS capable of supporting the operators in achieving safer decision-making, not only in nuclear power plants’ severe accident management but also in similar safety-critical systems.

Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years

Abstract This article gives an impact analysis of utilization of nuclear power plant full scope simulator on operation parameters, training and education in nuclear power plant Krško (NEK). The Slovenian Nuclear Safety Administration issued their simulator decree to NEK in April 1995. The first training session on the simulator was performed in April 17, 2000 and since then the simulator has been used on daily bases to improve operator knowledges, skills, and performances. At the time, this was the first full scope simulator with the capability to simulate Beyond design basis accidents (severe accidents). The ability to simulate core meltdown and containment breach made it very suitable for emergency preparedness drills. After the 2017 simulator upgrade, fuel meltdown in the spent fuel pool can be simulated using the Modular Accident Analysis Program—MAAP5. This capability is still unique for full scope simulators even today. The simulator is also used for pretesting of plant modifications before their implementation on site or for just-in-time training for infrequent performed evolutions or for procedure development and testing. The Pressurized Water Reactor Owners Group used the NEK simulator in 2018 to develop the new set of the Severe Accident Management Guidelines, incorporated with a completely new usage approach. In all of these years, the simulator has been actively participating in the increased reliability and stability of the electricity production and in achieving NEK's vision to be a worldwide leader in nuclear safety and excellence.

A study of RCS depressurization strategy of CPR1000 SAMG

Abstract CPR1000 severe accident management guideline (SAMG) is a product of the Westinghouse Owners Group (WOG) SAMG and the French SAMG (called GIAG) combination. It adopts the structure of WOG SAMG. However, it follows the approach of GIAG performing reactor coolant system (RCS) depressurization immediately after entering SAMG. This is a special feature of CPR1000 SAMG. The purpose of this paper is to evaluate this RCS depressurization strategy. The simulations of a station blackout (SBO) accident are performed for CPR1000 with RCS depressurization strategy adopted at both core exit temperature (CET) of 650 and 1100 °C together with opening different number of safety relief valves (SRV) using an integral severe accident analysis code. The optimized RCS depressurization strategy is provided to delay the reactor pressure vessel (RPV) failure time. And some comments are provided for future improvement.

Numerical assessment for entry condition of severe accident management guidelines in a Swedish nuclear power plant

The entry conditions of severe accident management guidelines (SAMGs) in pressurized water reactors (PWRs) rely on the indication of core exit temperature (CET). Yet, the setpoints for the CET may be different from plant to plant. Most Westinghouse PWR designs adopt the setpoint of CET at 650℃ as the entry condition of the SAMGs, since this setpoint is an effective indicator of core damage in a wide spectrum of accident sequences. Motivated by the interest in the verification & validation of SAMS after the Fukushima accidents, the present study is conducted numerically to verify the effectiveness of the CET setpoint for the transition from emergency operation procedures (EOPs) to SAMGs in a Swedish nuclear power plant. For this purpose, six representative severe accident sequences covering the main contributors to the core damage frequency (CDF) are analyzed using the MELCOR code. Moreover, different CET readings and alternative entry conditions are also investigated. The simulation results show that the average CET = 650 °C is the effective setpoint as the entry condition of SAMGs, i.e., given this setpoint the transition from EOPs to SAMGs will take place slightly before the occurrence of core degradation, which secures the intended mitigation of SAMGs while keeping EOPs active as long as possible. On the other hand, it is too conservative if the maximum CET = 650 °C is used as the setpoint of entry condition of SAMGs, i.e., it will result in an excessive realization of SAMGs over EOPs. The coolant temperature in the primary circuits, the water level in the RPV and the hydrogen concentration in the containment can also be applied as reference indications of core damage states in the accident management.

Analysis of primary side bleed and feed actions for severe accident management following total loss of feed water in a Swedish PWR

Severe accident management guidelines (SAMGs) have been developed and applied in nuclear power plants (NPPs) to provide the fourth level of defense worldwide since the TMI-2 accident which occurred in 1979. The primary objective of the SAMGs is to protect remaining safety boundaries and to limit release of fission products into the environment by specific severe accident management (SAM) actions with available safety equipment. This paper is concerned with the assessment of the primary side bleed and feed (PBF) actions in the SAMGs of a Swedish pressurized water reactor (PWR). The PBF actions are realized by operations of emergency injection pumps and power-operated relief valves (PORVs), following a severe accident induced by an event of total loss of feed water (TLOFW) and temporary station blackout. In the present study, MELCOR simulations are performed to investigate the effects of the SAM actions on severe accident progression and consequences. Different sets of available equipment and actuation time due to recovery of AC power are simulated. The interest of the MELCOR analysis is focused on how these uncertain parameters of SAM actions affect thermal–hydraulic response, hydrogen generation, core relocation, reactor pressure vessel failure and release of fission products. The results demonstrate the effectiveness of the PBF strategy to prevent RPV failure in the TLOFW accident and the effects of injection timing on accident consequences. The obtained insights are instrumental for validation and improvement of the strategies and actions in the plant specific SAMGs.

Severe accident analysis of the Qinshan Nuclear Power Plant and evaluation of boundary conditions for ex-vessel heat transfer

Severe accident analysis is critical for the operation and licensing of new generation reactor designs as well as for the prevention and management of severe accidents. The Severe Accident Management Guidelines (SAMGs) are developed and implemented for prevention and mitigation measures of a severe accident. Retention of corium in the reactor vessel via ex-vessel cooling is one of the most promising mitigation strategies. Its success requires a thorough analysis of the reactor core following a severe accident to determine the boundary conditions for external cavity flooding. In this study, two loss of coolant accident cases at the Qinshan nuclear power plant are analyzed using RELAP5/SCDAP simulations. Passive accumulators are credited in the first case (Case-1), but not in the second (Case-2). The important severe accident events are studied for the plant and the sequence of events and melt conditions are captured. Hydrogen generation has been estimated to be 120 kg in Case-1 and 108 kg in Case-2. In both instances, a molten pool is formed. The corium reached a maximum temperature of 3016 K and 3006 K, respectively in Case-1 and Case-2. The metallic/oxidic layer properties and heat transfer coefficient of the corium pool as well as the ex-vessel heat transfer boundary conditions are evaluated analytically for the reference plant. The analytically determined temperatures of oxidic and metallic phases in the corium pool are 3006 K and 1919 K respectively. The upward and sideward heat fluxes (vessel wall) are determined to be 298.3 kW/m2 and 730.75 kW/m2 respectively, while the heat transfer coefficients for oxidic and metallic phases are 2.80 W/m2-K and 4.42 W/m2-K respectively.

Experimental study on thermally assisted sagging deflection and interaction of multiple coolant channels

A reactor shutdown would typically initiate with a trip signal following a postulated Loss of Coolant Accident (LOCA). The Emergency Core Cooling System (ECCS) assures fuel cooling by flooding the core and recirculating coolant inside the channels for long-term decay heat removal from the fuel. When the primary flow is interrupted, and simultaneously the ECCS is severely impaired or failed, fuel cooling tends to degrade, leading to the heat-up of fuel and coolant channels. Analysis indicates that further progression of such accidents in Pressurized Heavy Water Reactors (PHWRs) is primarily delayed due to the huge water mass in the calandria vessel around the reactor core. However, the horizontal orientation introduces several phenomenological complexities introducing complexity in formulating the Severe Accident Management Guidelines (SAMGs). For a postulated unmitigated low-pressure accident scenario, the accident progresses with the channels heating up, and creep sagging under thermal load and channel and fuel weight. The sagged channel contacts the channel below, transferring heat and its load. With further progress, it is postulated that the supporting channel would fail, and leading to a cascading core collapse.The present paper discusses the experimental studies carried out on multiple channel sagging and disassembly phenomena envisaged during the mid-phase of a severe accident. It is challenging to perform full-length experiments; hence, scaled-down channels were used in the experiments. The postulated accident condition is a significant break LOCA with loss of ECCS and loss of all heat sinks. The experimental setup consists of three channels placed vertically in a fixed lattice pitch forming three distinct channel rows. The paper compares two possible situations or test cases: 1) all three channels are heated, 2) the top two channels are heated, and the lower channel is unheated. The lower channel is an unheated (dummy) channel supporting the upper deforming channels. Finally, a multi-channel sagging deformation mode has been postulated based on the experimental observations. The contact propagation along the contacting surface is found to be a time-dependent event, indicating a channel's rate-dependent elongation. The middle CT is compressed between the upper and lower channels, compromising its structural support for middle PT.

Framework to model severe accident management guidelines into Level 2 probabilistic safety assessment of a nuclear power plant

Severe accident management guidelines (SAMG) play a significant role in mitigating severe accidents in nuclear power plants (NPPs), and thus the modeling of SAMG into Level 2 probabilistic safety assessment (PSA) has become essential to more realistically quantify the safety of NPPs and the effectiveness of SAMG. This study developed a systematic framework to model SAMG into Level 2 PSA, in which SAMG event trees and fault trees are modeled to formulate SAMG in a probabilistic manner. Application results to the internal events of a real plant, the OPR-1000, showed that the inclusion of SAMG in the Level 2 PSA model led to not only a significant increase in the conditional containment intact probability, from 37 to 70%, but also a considerable decrease in NPP risk, namely by 44% compared to the case without SAMG. The method can be used for the estimation of the effectiveness of each mitigation strategy, as well as for the identification of the risk-significant factors regarding SAMG.

Development of a decision-making support system for the technical support center based on analysis of human performance under severe accident situation

After the Fukushima accident, hardship of group decision making under severe accident situation is one of human factor issues emerged by Fukushima accident. After core damage, the responsibility for plant control is shifted from main control room (MCR) operators to technical support center (TSC) members who need to follow severe accident management guideline (SAMG). Decision making is necessarily required from this time to establish a suitable strategy because SAMG is not a procedure but it is a broad guideline so it cannot give a precise correct answer to TSC members for mitigating accident sequence like EOP.The developed DM support system can shorten the group DM time through a semi automatic algorithm. In addition, it is possible to gather the opinions of all members of TSC by excluding unnecessary obstructive human factors under uncertain situations like lack of information.