Offsite Power Research Articles

A site risk assessment for internal events: A case study

Site risk assessment (SRA) is to estimate the risk of accidents that are probable to occur on the site. The group of accidents consists of not only the initiating events inducing impact on a single unit, but also the events causing simultaneous malfunctions, such as reactor trip on more than 2 units by common causes, called common-cause initiators (CCIs). These events need to be distinguished at the initiating event analysis and the accident progress should be analyzed separately to conduct the SRA. Furthermore, inter-unit dependencies such as inter-unit common-cause failure (CCF) and unavailability of shared components should be treated appropriately to model the accident sequences with CCI events. In this study, in order to perform the SRA of a reference site, the accident history of nuclear power plants in Korea is investigated to identify single-unit initiators (SUIs) and CCIs for SRA and to calculate the frequency of each initiator. The Level 1 SRA model is developed implementing inter-unit dependencies, unavailability of alternate AC diesel generator and inter-unit CCF. In addition, re-evaluation of failure probability of off-site power recovery is performed. Using these models, site core damage frequency (SCDF) and the site risk indicators identified in this study are calculated to deduce the impact of the multi-unit factors considered in SRA. It is expected that the SRA model developed in this study can be applied to perform the site risk assessment for the other site and contributes to improve the safety from the site risk perspective.

Numerical simulation on transient thermal and hydraulic characteristics in sodium pool of CEFR under OPT-SLOOP and OPT-RHROSL conditions

One Pump Trip (OPT) may cause asymmetric phenomena in China Experimental Fast Reactor (CEFR) primary system. Not only two loops asymmetric coast-down processes and failure pump backflow, but superposition loss of off-site power (SLOOP) or residual heat removal of single loop (RHROSL) with OPT may make the situation more complicated. In order to investigate transient thermal hydraulic characteristics in sodium pool under OPT-SLOOP and OPT-RHROSL conditions, a full-scale numerical 3D model of sodium pool in CEFR was established, in which the transient processes of coast-down and backflowwere simulated. The simulation results revealed that, in the case of OPT-SLOOP, the intact loop overall temperature was significantly higher than the failure loop overall temperature in the cold sodium pool, and the maximum IHX outlet temperature difference between the two loops reached 125.7℃, while in the case of OPT-RHROSL, the maximum IHX outlet temperature difference was limited to 15℃.

Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category

The Breeding Blanket (BB) is one of the key components of the European Demonstration (EU-DEMO) fusion reactor. Its main subsystems, the Breeder Zone (BZ) and the First Wall (FW), are cooled by two independent cooling circuits, called Primary Heat Transfer Systems (PHTS). Evaluating the BB PHTS performances in anticipated transient and accident conditions is a relevant issue for the design of these cooling systems. Within the framework of the EUROfusion Work Package Breeding Blanket, it was performed a thermal-hydraulic analysis of the PHTS during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”, by using Reactor Excursion Leak Analysis Program (RELAP5) Mod3.3. The BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat Exchangers were included in the study. Selected transients consist in partial and complete Loss of Flow Accident (LOFA) involving either the BZ or the FW PHTS Main Coolant Pumps (MCPs). The influence of the loss of off-site power, combined with the accident occurrence, was also investigated. The transient analysis was performed with the aim of design improvement. The current practice of a standard Pressurized Water Reactor (PWR) was adopted to propose and study actuation logics related to each accidental scenario. The appropriateness of the current PHTS design was demonstrated by simulation outcomes.

An Extended Thermosyphon Cooling System for APR-1400 Reactor Design

Advanced Power Reactor (APR-1400) is a Generation III + Pressurized Water Reactor (PWRs) and has gained popularity among energy mix community. APR-1400 features enhanced safety limits to prevent a “Fukushima-type” accident scenario for a duration up to 8 h. This work proposes a system that enhances APR 1400 passive cooling (Thermosyphon Cooling System, TCS) capabilities to guarantee at least double the coping time before core meltdown. TCS provides enough cooling to the reactor core in the case of loss of offsite power, known as Station Black-Out (SBO) accident. This work invites to use the In-containment Refueling Water Storage Tank (IRWST) to further extend the APR1400 capability to passively cool the reactor core after an SBO accident. This work is designed and tested by 3Keymaster simulator platform developed by Western Service Corporation. Simulation has modified the existing model of APR1400 in the simulator by adding pipes, valves and heat exchanger. Results show the system will cool the reactor core for almost 18 h after the SBO accident. The system works initially in parallel with steam generators then it continues cooling the reactor core for extra 18 h resulting in providing more time for restoring the on-site power sources and preventing core meltdown.

Presenting new approach for optimal placement of nuclear power plant connected to the grid after the trip

This study presents a combination of optimal placement and power system development with the aim of supplying electricity for Nuclear Power Plant after the trip of power plant. Power supply to the internal loads of power plant by the off-site power system is one of the main fields of research in achieving the safety Nuclear Power Plant. One of the main purposes in this article is to introduce a suitable and safe place for the construction and connection of a Nuclear Power Plant to the power system. These locations are identified by the power plants on-site loads and the average of the lowest number of relay protection after the Nuclear Power Plant trip, based on electricity considerations. Along with the optimal placement in this paper, the power system development, including the generation and transmission development in order to provide electricity with higher reliability to the Nuclear Power Plant after the trip is also presented. Monte Carlo and Latin Hypercube Sampling probabilistic methods are proposed for locating the site of a Nuclear Power Plant and algorithms of Genetic and Particle Swarm Optimization for locating and developing power generation and transmission systems. The simulation results are implemented on the IEEE RTS 24-bus system, and finally suitable locations for the construction of the Nuclear Power Plant and the generation and transmission development with the aim of feeding the power plant from the off-site power system and sufficient assurance that the reactor core does not melt after the trip, are determined.

VVER-1000/V446 spent fuel pool risk assessment and support through portable mitigating equipment

The Spent Fuel Pool (SFP) accident at the Fukushima Daiichi unit 4 demonstrated the need for identifying and analyzing possible accident sequences and subsequent application of severe accident mitigating systems in LWR (Light Water Reactor) SFPs. In this study, Level 1 Probabilistic Safety Assessment (PSA) is applied to evaluate the Fuel Damage Frequency (FDF) in SFP of Bushehr Nuclear Power Plant unit-1 (BNPP-1). This paper specifically focuses on assessing the SFP risk using Portable Mitigating Equipment (PME) including a reliable robust Portable Diesel Generator (PDG), a Portable Diesel Pump (PDP), and an SFP-spray system proposed to maintain steady decay heat removal. PME is the recommended equipment proposed by IAEA -following the Fukushima disaster- to mitigate accident consequences. In this study, the fault/event trees have been applied to conduct PSA using the SAPHIRE code. The results indicate that the most dominant Initiating Events (IEs) in the reference plant are the Loss of Offsite Power (LOOP) and Loss of Service Water (LOSW), both simultaneously affecting the reactor core and the SFP. There is a significant decrease in calculated FDF – after implementing PDG and PDP- in respect of the reference plant from 3.086E to 6 to 2.166E−7 in the normal storage mode. Heavy load drop, which is a low probability/high consequence IE, becomes crucially important in cases where PME is employed as it directly leads to fuel damage. It is expected that the analyses set forth here would provide efficient guidelines and insights into the strengths and weaknesses inherent in the design and operation of SFP in VVER-1000/V-446 nuclear power plant.

Probabilistic Safety Analysis for Loss of Offsite Power Accident in Dual-units Nuclear Power Plant

In order to explore the risk assessment method of the multi-unit nuclear power plant site, this paper selects the dual-unit plant nuclear site to analyze lose off-site power accident. By combining and improving the single-unit ET/FT model, to establish the dual-unit ET/FT model. From the analysis of the accident sequence, it can be concluded that the common cause failure of equipment is the main challenge faced by the dual-units. Especially the RPC sub-channel in the reactor protection system and the failure of emergency diesel engine circuit breaker. As can be seen from the high proportion of core CD occurring simultaneously in both uints, it has a great significance to study the risk of mult-units sites.

Simulation of a Station Blackout Accident for the SMART Using the CINEMA Code

SMART is a small-sized integral type PWR containing major components within a single reactor pressure vessel. Advanced design features implemented into SMART have been proven or qualified through experience, testing, or analysis according to the applicable approved standards. After Fukushima accident, a rising attention is posed on the strategy to cope with a Station Blackout (SBO) accident, which is one of the representative severe accidents related to the nuclear power plants. The SBO is initiated by a loss of all offsite power with a concurrent failure of both emergency diesel generators. With no alternate current power source, most of the active safety systems that perform safety functions are not available. The purpose of SBO analysis in this paper is to show that the integrity of the containment can be maintained during a SBO accident in the SMART (System-integrated Modular Advanced ReacTor). Therefore, the accident sequence during a SBO accident was simulated using the CINEMA-SMART (Code for INtegrated severe accidEnt Management and Analysis-SMART) code to evaluate the transient scenario inside the reactor vessel after an initiating event, core heating and melting by core uncovery, relocation of debris, reactor vessel failure, discharge of molten core, and pressurization of the containment. It is shown that the integrity of the containment can be maintained during a SBO accident in the SMART reactor. It has to be mentioned that the assumptions used in this analysis are extremely conservative that the passive safety systems of PSIS and PRHRS were not credited. In addition, as ANS73 decay heat with 1.2 multiplier was used in this analysis, actual progression of the accident would be much slow and amount of hydrogen generation will be much less.

Development of a hybrid method to assess grid-related LOOP scenarios for an NPP

Loss of Offsite Power (LOOP) is the most dominant initiating event in the Core Damage Frequency (CDF) of Nuclear Power Plants (NPPs). Since the frequency of occurrence of LOOP depends on the grid configuration, outage schedule, load flow, etc., even a relevant generic frequency may result in significant uncertainty.In this study, an analytical hybrid methodology, which combines the probabilistic and deterministic modules, is introduced to systematically estimate the frequency of Grid-related LOOP. The proposed probabilistic module develops the relevant event-trees and fault-trees in SAPHIRE7 to identify the LOOP sequences upon the response of the distance protection system to a fault on the transmission lines. Meanwhile, the identified scenarios are confirmed based on the simulation results from the deterministic module in DIgSILENT-PowerFactory15.1.15-out-of-48 scenarios (100-out-of-300 sequences) are identified as grid-related LOOP upon faults on the four transmission lines of the test grid, and result in 1.76E-06/yr for the grid-related LOOP frequency. The developed methodology enables us to trace a LOOP chain, from a primitive presumed fault through the progress of the cascade. The grid-related LOOP scenarios in a certain network can be ranked probabilistically, with the desired level of details in both deterministic and probabilistic modules.

Dependency analysis method for human failure events in multi-unit probabilistic safety assessments

Dependency between human failure events (HFEs) is often analyzed as a part of the conventional human reliability analysis (HRA) process for nuclear power plants (NPPs). None of the existing methods have considered an application to multi-unit (MU) scenarios. In this study, we introduce a novel approach to evaluate the dependencies between HFEs in cases of MU event scenarios. We focus on developing a set of MU HFE dependency evaluation elements and their evaluation criteria based on the framework of the single-unit evaluation elements that have been utilized in the HRA practice for NPPs. The unique MU HFE dependency evaluation elements are defined, and we present a new dependency evaluation tree for qualitative analysis. In addition, important cases are provided to demonstrate the application of the proposed method, including four MU initiating events: MU loss of offsite power, MU loss of condenser vacuum, MU general transient, and MU loss of ultimate heat sink. Quantitative analysis with several illustrative HFEs in MU probabilistic safety assessment (PSA) cutsets are also presented. This method fills the gap of a HFE dependency analysis method in the case of MU PSA.

High Reliability Microgrid for a Nuclear Facility Emergency Power Supply

Abstract This paper investigates the use of a high reliability microgrid to supply emergency electrical power to a nuclear facility following loss of offsite power accident. Majority of the nuclear facilities around the world utilize diesel generators (DGs) and battery banks as emergency power to back up the grid power supply. This power supply configuration represents the concept of the microgrid system. The research work proposes reliability improvement of the emergency power supply by introducing distributed energy sources and large-scale energy storage systems. The proposed system was modeled with reliability block diagrams (RBDs) and analyzed with Monte Carlo simulation (MCS) to evaluate the system's reliability in terms of loss of power supply probability (LPSP). The analysis takes into consideration the failure probability of the components combined with the power unavailability due to the intermittent energy sources. The results show that the proposed system reduces the LPSP.

The risk analysis of the electrical Cross-tie for extended Station Blackout in Multi-Unit site

The SBO is defined as the events of Loss of Offsite Power (LOOP) combined by loss of dedicated Emergency Diesel Generators(EDGs) for each unit. The SBO can be also extended when the shared Diesel Generator (AAC-DG) is lost additionally. The Electrical Cross-tie (ECT) is the concept of sharing other unit’s dedicated EDG to the unit under extended SBO in the multi-unit NPP site. We re-evaluate the risk of the extended SBO with the ECT option in Single Unit PRA considering Multi-Unit Features. To get a more realistic estimation of Core Damage Frequency (CDF), we developed the KU-PRA model using the AIMS-PSA code. We examined the factor of redundancy of the AC sources on the CDF by adding additional EDGs to the reference KU-PRA Model. We developed methodology and modeling of the Electrical Cross-tie for Multi-units PRA considering Common Cause Failure, Human Reliability Analyses, Battery load shedding, and Occupancy factor.

Preventing cascading failure of electric power protection systems in nuclear power plant

Cascading failure is the main cause of large blackouts in electrical power systems; this paper analyzes a cascading failure in Hanbit nuclear power plant unit two (2) caused by a circuit breaker (CB) operation failure. This malfunction has been expanded to the loss of offsite power (LOOP). In this study, current practices are reviewed and then the methodologies of how to prevent cascading failures in protection power systems are introduced. An overview on the implementation of IEC61850 GOOSE messaging-based zone selective interlocking (ZSI) scheme as key solution is proposed. In consideration of ZSI blocking time, all influencing factors such as circuit breaker opening time, relay I/O response time and messages travelling time in the communication network should be taken into account. The purpose of this paper is to elaborate on the effect of cascading failure in NPP electrical power protection system and propose preventive actions for this failures. Finally, the expected advantages and challenges are elaborated.

A framework for verifying Dynamic Probabilistic Risk Assessment models

Recent development of more powerful computational and technological resources has led to significant improvements in the utilization of dynamic methodologies for the Probabilistic Risk Assessment (PRA) of nuclear power plants. These methodologies integrate deterministic and probabilistic analyses and are generally referred to as Dynamic PRA (DPRA) methods. DPRA is performed through the generation and simulation of possibly thousands of different accident scenarios. To ensure the quality and the correctness of the results, DPRA models should be verified. Since DPRA generates large amount of data, a visual inspection of results to verify the correctness of the model used is neither practical nor reliable. As one of the steps for DPRA analysis, a framework is proposed to systematically explore the DPRA model prior to its simulation using statecharts which provide a graphical notation for describing dynamic aspects of system behavior. The application of the framework is illustrated using two case studies: (i) performance assessment of a heated room using the PyCATSHOO DPRA tool, and, (ii) DPRA performed with RAVEN-MAAP5-EDF codes for loss of off-site power as the initiating event in a pressurized water reactor.

Innovative passive management of the Station Black-Out accident for VVER1000/V446 NPP and development of its emergency operating procedures

Loss Of Offsite Power (LOOP) accident is one of the most important Design Basis Accidents (DBAs) that originate due to loss of the house-load demands which make serious concerns for each NPP. Moreover, in the case of loss of all diesel generators (due to earthquake, etc.), the accident forwards to the Station Black-Out(SBO). Following and learning the Fukushima accident, the cares have been focused on using internal and/or external water resources for the passive management of the SBO accident. In the current study, passive strategies in the VVER-1000/V446 NPP are proposed using a tank in the secondary loop (called Deaerator) for the SBO management. After RELAP5 modeling benchmarking in the steady-state as well as standard SBO management (which are mentioned in the NPP FSAR), two operational SBO management are proposed herein including “Feeds and Bleeds” (in the primary and secondary loops) and also including a Deaerator tank in the secondary side. The results show that using the Deaerator-water tank of the secondary side as well as feed and bleed, the time available for operators before the core heat-up (i.e. fuel cladding temperature of 1200 °C); extends to 35000 s. Obviously; it gives a good opportunity for plant operators to access further time for using other facilities/methods to avoid core damage.

Loss of Offsite Power Recovery Modeling in United States Nuclear Power Plants

ABSTRACTLoss of offsite power (LOOP) can have a major, adverse impact on a nuclear power plant's (NPP) ability to achieve and maintain safe shutdown conditions. The time required for subsequent restoration of offsite power after a LOOP event occurred and the probabilities of LOOP events exceeding various durations (or LOOP non‐recovery probabilities) are important inputs to NPP probabilistic risk assessments (PRA). This paper reviews the analysis of LOOP events at United States commercial NPPs conducted by Idaho National Laboratory (INL) for the US Nuclear Regulatory Commission (NRC). This paper presents the current LOOP recovery modeling that estimates probabilities of LOOP events exceeding various durations (or LOOP non‐recovery probabilities) based on operating experience. NPPs use these LOOP results in PRA models for various risk‐informed activities. Finally, this paper provides the LOOP non‐recovery probability results for the four LOOP categories: plant‐centered, switchyard‐centered, grid‐related, and weather‐related with LOOP recovery data from 1988 to 2018.

Effects of house load operation on PSA based on operational experiences in Korea

House load operation (HLO) occurs when the generator supplies power to the house load without triggering reactor trips during grid disturbances. In Korea, the HLO capability of optimized power reactor 1000 (OPR1000) plants has prevented several reactor trips.Operational experiences demonstrate the difference in the reactor trip incidence due to grid disturbances between OPR1000 plants and Westinghouse plants in Korea, attributable to the availability of the HLO capability. However, probabilistic safety assessments (PSAs) for OPR1000 plants have not considered their specific design features in the initiating event analyses.In an at-power PSA, the HLO capability can affect the initiating event frequencies of general transients (GTRN) and loss of offsite power (LOOP), resulting from transients within the grid system. The initiating event frequencies of GTRN and LOOP for an OPR1000 plant are reduced by 17.7% and 78.7%, respectively, compared to the Korean industry-average initiating event frequencies, and its core damage frequency from internal events is reduced by 15.2%. The explicit consideration of the HLO capability in initiating event analyses makes significant changes in the risk contributions of the initiating events. Consequently, for more realistic at-power PSAs in Korea, we recommend incorporating plant-specific HLO-related design features when estimating initiating event frequencies.

A development of methodology for assessing the inter-unit common cause failure in multi-unit PSA model

The inter-unit common cause failure (CCF) is an event that multiple components are failed simultaneously by common factors may existent in nuclear power plants on a site. The inter-unit CCF can be identified as one of the inter-unit dependencies that has an impact on the multi-unit probabilistic safety assessment (MUPSA). For instance, it may adversely affect the accident mitigation in accident scenarios of multiple units. Therefore, the MUPSA model should be developed considering the inter-unit CCF.In this study, we proposed two methods with the Swain's dependency model and the CCF data in the single-unit PSA (SUPSA) to calculate the inter-unit CCF. The Swain's dependency model is used to calculate the conditional probability considering the dependency level between each CCF event in two different units. To perform the case study, the inter-unit CCF events of emergency diesel generators (EDGs) and alternate AC diesel generators (AAC DGs) in 4 units of different pressurized water reactor (PWR) types were analysed in the case of the multi-unit loss of off-site power (MULOOP). It is expected that methodologies in this study will contribute to the multi-unit level 1 PSA model development as well as evaluating the site risk in the future.

Risk Reduction Assessment for Installing RCP Shutdown Seal

As one of the lessons learned from the Fukushima Daiichi accident, long-term station blackout (SBO) and subsequent loss of ultimate heat sinks have prompted discussion on this topic in the nuclear industry. The SBO sequences for a Westinghouse three-loop pressurized water reactor (PWR) have been under investigation for a long period. To cope with the long-term SBO issue, many nuclear power plants have decided to replace the reactor coolant pump (RCP) seal by the new passive thermal shutdown seal (PSDS). The PSDS is a fail-safe protection device that will significantly reduce leakage from the RCP seal in case of loss of cooling. This makes a seal loss-of-coolant accident no longer a risk-significant event, and the relevant probabilistic risk assessment (PRA) models need to be modified to reflect the associated plant change. The PRA model of a Westinghouse PWR plant has been reviewed to reflect this more strongly; i.e. loss of component cooling water (CCW), loss of 4-kV vital alternating-current power, and loss of off-site power are revised for their sequences. According to the Westinghouse analysis, the PSDS temperature must be maintained below 104°C, and the operators have to control the RCP speed. In this paper, those factors are incorporated into the event tree structure revision of the loss of CCW event (TC) and loss of power either off-site (TP) or vital power A train (TAPB). In another case, LOOP initiating events need to consider the time span that the blackout conditions would affect the RCP seal integrity. Because the RCP will be tripped automatically, the limitation associated with RCP speed will not be applicable. Compared with the TC, TP, and TAPB event tree base cases, RCP speed slowdown or available time span is introduced into the PSDS model. The relevant part in the PRA model is subject to review and modification. The risk reduction associated with the PSDS is found to be significant.

Study on high wind hazard probability risk assessment methods of nuclear power plant

China’s nuclear power plants were built in coastal areas. High wind such as hurricane would make nuclear power plant transmission lines, structure and component to fail, leading to the transient or loss of off-site power accidents, which have an impact on the safety of nuclear power plants. This paper focuses on high wind hazards analysis of the operational nuclear power plants. We study the high wind probability safety assessment methods and the main technical elements. The following five tasks were performed to complete the high wind PSA models: Site Specific High wind hazard analysis, High wind Walk-downs and equipment list screen out, High wind Fragility Analysis, Integrate High wind Impacts into PSA Model, and Risk Quantification of High wind PSA Model. The methodology can be used to addresses high wind hazard probability risk analysis of nuclear power plant.