Offsite Power Research Articles

Probabilistic Safety Assessment of Off-site Power System under Typhoon Considering Failure Correlation between Transmission Towers

High-intensity typhoons can cause failure to the structures, systems, and components of nuclear power plants (NPPs) and off-site power systems. Instances have occurred where failure of the off-site power system caused by typhoons has affected NPPs. The off-site power system has been in cases where a typhoon failed multiple transmission towers. In Korea, transmission towers have similar design criteria and material properties and experience similar wind characteristics, leading to a correlation in their failure probabilities. Assuming independent or totally dependent failure probabilities among the transmission towers is unrealistic. For a realistic PSA of the off-site power system due to typhoon-induced high winds, it is necessary to consider an appropriate failure correlation. This study performed a PSA of an off-site power system due to typhoon-induced high winds, considering the failure correlation between transmission towers. The failure correlation between transmission towers based on distance was calculated. The wind fragility of the off-site power system was convolved with the typhoon-induced high-wind hazard at the Kori NPP to calculate the probability of the NPP losing its power supply. Thus, applying the failure correlation of typhoon-induced high winds to the PSA of the off-site power system is considered realistic.

Methodology for Generating Wildfire Hazard Map for Safety Assessment of Off-Site Power Systems Against Wildfires

Methodology for Generating Wildfire Hazard Map for Safety Assessment of Off-Site Power Systems Against Wildfires

Dynamic Seismic Probabilistic Risk Assessment of Nuclear Power Plants Using Advanced Structural Methodologies

The conditional core damage probability (CCDP) of a representative nuclear power plant is estimated for a beyond design basis earthquake (BDBE) using state-of-the-art structural models within a dynamic probabilistic risk assessment (DPRA) framework. Randomness of seismic excitation and uncertainty of structural parameters are considered using a simulation-based approach. Finite element models are developed for the structure and a liquid container (hydro-accumulator), and fluid–structure interaction is considered with the arbitrary Lagrangian-Eulerian method. The CCDP is evaluated through a time-dependent event tree. Also, correlation among multiple hydro-accumulators is investigated. The results show that operator action and implementation time of mobile safety equipment are critical under BDBE in case of loss of offsite power following an earthquake. It was also found necessary to adopt a DPRA approach to determine the available time for action and core damage timing probabilistically.

NPP accident prevention: Integrated neural network for coupled multivariate time series prediction based on PSO and its application under uncertainty analysis for NPP data

Due to the requirement of a cleaner, more sustainable form of energy production and the rapid development of DT, nuclear energy needs to be digital transformed. This paper aims at the time series prediction module of DT in NPP. Despite the widespread application of time series forecasting in various domains, the inherent uncertainties within the data and the selection of neural network hyperparameters pose a formidable challenge to accurate predictions. This study proposes a coupled multivariate prediction method and conducts a comparative analysis of three popular time series forecasting models: LSTM, CNN-LSTM, and Transformer. PSO hyperparameter optimization is also employed, accompanied by multiple error calculation metrics for model evaluation. It is noteworthy that the experimental data employed in this study are derived from the real operation process of a Gen II + reactor in the Chinese mainland, which is in the off-site power loss accident condition. Simultaneously, adopting a smaller training dataset proportion contributes to striking a balance among resource utilization, model performance, and experimental efficiency in the research. Experimental results unequivocally demonstrate that among these models, Transformer excels in multi-input multi-output time series forecasting. Moreover, uncertainty analysis using Bayesian generates a forecast band, proving the robustness of Transformer.

Methodology for estimating the contribution of forest fires in loss of offsite power events

Forest fires are disasters caused by natural or human factors and have environmental as well as economic impacts such as loss of biodiversity, agricultural damage, and property damage. According to statistics, the number of forest fires and the area damaged are gradually increasing in Korea. In addition, there are cases where the units 1, 2, and 6 of the Hanul Nuclear Power Plant (NPP) have already been affected by forest fires. Korea has recently initiated research on forest fires under multi-unit probabilistic safety assessment (PSA) project. This paper deals with the impact on the loss of offsite power (LOOP), which is one of the NPP incidents and accidents affected by forest fires. To estimate the frequency of LOOP induced by forest fires, the PSA methodology for earthquakes, a representative external event, was referred.

Loss of power accident analysis for the HCCR blanket

The blanket system is a crucial element in nuclear fusion power generation, tasked with tritium production and the efficient dissipation of the high thermal flux from the tokamak. Korea is in the process of developing the Helium Cooled Ceramic Reflector (HCCR) blanket, which stands as the forefront design among the nation's blanket systems. This paper presents accident analysis results of LOOP (Loss Of Offsite Power) accident for the HCCR blanket. Loss of power accident is one of the postulated events in the nuclear fusion system. This event may be caused by common failure of the external electrical grids or electrical faults in the site. The objective of this safety analysis is to assess the impact of the activation of an additional power supply within 90 seconds after an accident on the system temperature, and to evaluate the heat load that can be removed by the stagnated coolant in the CCWS (Component Cooling Water System) connected to the heat exchanger of the blanket system.

MUPSA-based evaluation of portable equipment utilization for risk reduction in nuclear power plants: Framework and case study

Based on the lessons from the Fukushima accident, new strategies to mitigate the spread of accidents in nuclear power plants have been established that consider portable equipment. The diverse and flexible coping strategies (FLEX) and multi-barrier accident coping strategy (MACST) based on the utilization of portable equipment are being adopted in case of an extended loss of all AC power and loss of ultimate heat sink. In this context, the establishment of specific strategies for utilizing portable equipment has emerged as an important issue. This study proposes a multi-unit probabilistic safety assessment (MUPSA)-based evaluation framework consisting of several steps to quantitatively assess the reduction in risk from the use of portable equipment at a multi-unit site, which can provide a basis for establishing optimal portable equipment strategies. According to the framework, after first defining the scope of the evaluation, risk-significant types of portable equipment are selected and the scenarios are analyzed. Risk reduction is then evaluated using modified MUPSA models according to the deployment of the portable equipment. A case study was conducted for an internal loss of off-site power (LOOP) and seismic LOOP focusing on how to evaluate the effectiveness of portable equipment through the developed framework.

Multi-unit PSA based risk evaluation framework for utilizing cross-tie systems for nuclear power plants

The Fukushima accident showed that the safety of multiple nuclear power plants (NPPs) at the same site could be jeopardized simultaneously. Since then, many studies have focused on developing strategies to prevent the spread of multi-unit accidents, with numerous countries establishing strategies to use mobile equipment. However, mobile equipment strategies are inherently accompanied by a high degree of uncertainty regarding operation success and duration because multiple organizations and personnel interact in various ways during multi-unit accident situations. Furthermore, supplementing current fixed equipment with additional mobile equipment requires extra resources. Therefore, cross-tie strategies that use currently installed fixed equipment can provide additional means to manage site risk with relatively few additional costs. This study proposes a multi-unit probabilistic safety assessment–based risk evaluation framework for utilizing cross-tie systems in NPPs and a modeling methodology to quantify the effectiveness of the cross-tie strategies. A case study was conducted to evaluate the risk reduction from using cross-tie strategies for emergency diesel generators and alternate AC diesel generators, which are power systems utilized in multi-unit loss of offsite power initiating events. It is expected that the developed framework and methodology can be utilized for other types of cross-tie strategies as well.

An Off-Site Power Purchase Agreement (PPA) as a Tool to Protect against Electricity Price Spikes: Developing a Framework for Risk Assessment and Mitigation

Significant price spikes occurred as early as 2021, initially driven by low gas storage levels, a post-pandemic economic rebound and then exacerbated by the Russian invasion of Ukraine. The situation had a range of wide-ranging consequences, from rising inflation, increasing energy poverty, food insecurity, business bankruptcies and recession. A well-known tool to protect energy consumers from energy price spikes, while at the same time contributing to the development of sustainable technologies, is Power Purchase Agreements. PPAs are long-term bilateral contracts for the purchase and sale of a certain amount of electricity, usually generated from renewable sources. The primary goal of this paper is to assess how the risk associated with PPAs has evolved between 2020 and 2023. It aims to examine whether, after the events in 2022, PPAs remain a robust solution that protects the off-taker from energy price spikes, ensures greater energy budget stability and enables savings. To achieve this, the probability of PPA prices being higher than market prices is evaluated, considering the changing market landscape. Furthermore, this paper intends to gain a thorough understanding of each risk related to PPAs and the best strategies for mitigating it, to maximize the protection of the off-taker.

Preliminary accident analysis of the loss of vacuum in vacuum vessel for the European DEMO HCPB blanket concept

Design basis accidents are investigated continuously for the European DEMO reactor accompanying its development. One selected postulated initial event (PIE) is a loss of vacuum (LOVA) in vacuum vessel (VV) with large ingress of air induced by rupture in a VV penetration. It has been investigated for the helium cooled pebble bed (HCPB) blanket concept according to the DEMO baseline 2017. The associated primary heat transfer system (PHTS) and the related systems in the tokamak building, from the VV to the PHTS vault and galleries, are considered for the investigation. The LOVA is postulated to occur at a port seal of the electron cyclotron equatorial port plug on the side of the closure plate with (i) a small leak of 1.0 × 10−3 m2, or (ii) a large break size of 1.0 × 10−2 m2. Air ingress from one port cell into the VV leads to the VV pressurization and the fusion power termination followed by an unmitigated plasma disruption. A loss of off-site power for 32 h is assumed to coincide with the disruption. An in-vessel loss of coolant accident (LOCA) is considered as a consequence if the affected first wall (FW) reaches the defined temperature of 1000 °C. The radioactive inventories in the VV (tritium, W-dust) can mobilize towards the VVPSS, the affected systems in the building and the environment due to pressurization, venting and leak conditions. MELCOR 1.8.6 for fusion is applied for this deterministic safety analysis. The resulting releases of radioactivity to the environment are then provided for dose calculation using the computer systems UFOTRI and COSYMA. Outcomes of this LOVA analysis are critically discussed: the transient evolutions of different cases are compared; hydrogen production is detected in case of aggravating FW failure; the source terms (tritium, W-dust) are transported to the connected systems; and the dose results from the environmental releases are provided.

Transient Analysis of LBLOCA with Subsequent Loss of Offsite Power and Loss of All AC Power in VVER-1200 Reactor

The generation III+ VVER-1200 (AES-2006) reactor is equipped with improved safety features with component redundancy compared to the previous models. This paper is concerned with the performance of active and passive safety features during the transient of a major beyond design basis accident (BDBA) scenario of the reactor. The study is done considering large break loss of coolant accident (LB LOCA) with loss of offsite power (LOOP) and LB LOCA with loss of all AC power (LOACP) sources namely station blackout (SBO) using VVER-1200 NPP simulator. It was found that the reactor is capable of mitigating such BDBA scenarios by the effectiveness of the active and passive safety systems. Moreover, LB LOCA with LOACP is observed to be more severe compared to LBLOCA with LOOP. This comparative analysis is compatible with IAEA safety guidelines.

Addressing the development gap in net-zero energy buildings: A comparative study of China, India, and the United States

In recent years, the urgency to address climate challenges has driven the need to achieve net-zero carbon goals by 2050. The construction industry, as a major contributor to carbon emissions, plays a vital role in achieving this target. While numerous net-zero energy building approaches have been proposed globally, there is limited comparative research between developed and developing countries. This study aims to address this research gap by selecting the top three carbon-emitting countries in the world - China, India, and the United States (US) - as the research subjects. Through a comparison of their net-zero energy building approaches and typical cases, it is evident that there exists a significant disparity between developed and developing nations. Specifically, differences are observed in the following aspects: net-zero building approaches in the United States have an earlier start and involve collaborations between the government and multiple organizations, while China and India started relatively later and rely primarily on government initiatives. China's consideration of applicable building types in net-zero approaches is not inclusive enough, particularly lacking coverage on industrial buildings. It is encouraging to observe that China and India pay more attention to embodied carbon emissions compared, while their decarbonization strategies for embodied carbon emissions are not as comprehensive as those in the United States when it comes to actual building applications. Regarding carbon offset options, the United States and India demonstrate more comprehensive considerations, while China currently lacks consideration for off-site power generation and negative carbon technologies. The United States boasts a larger number of net-zero buildings, most of which have been certified through the country's own net-zero building approaches. In contrast, China and India have fewer certified net-zero buildings, and only a small percentage have been certified using their own methods, especially in China. The technologies employed in zero‑carbon buildings in the United States are more advanced compared to China and India. These findings underline how the disparities in economic levels between developed and developing countries are reflected in net-zero building approaches. Based on these insights, professionals in the field can develop tailored strategies to bridge the gap in net-zero building development between developed and developing nations. This study provides a foundation and recommendations for achieving carbon neutrality in the construction industry for the world's top three economies by 2030.

Total Loss of Feedwater Analysis of Pressurized Water Reactor Using RELAP5

Abstract In Europe, the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW is multiple failures in which, in addition to the loss of main feedwater, the auxiliary feedwater is also lost. The scenarios without DEC safety features and the scenarios with DEC safety feature assumed have been simulated. The results showed that after TLOFW event initiation, it is very important to trip the reactor as soon as possible. In case of loss of offsite power, the reactor coolant pumps stop, and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed, the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety features assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited in the first 30 min of event. When primary side bleed and feed is used, less time is available for operator actions.

Numerical Simulation of the Loss of Offsite Power to the VVER-1200 Nuclear Power Unit Auxiliaries

Numerical Simulation of the Loss of Offsite Power to the VVER-1200 Nuclear Power Unit Auxiliaries

A mechanistic model of a PWR-based nuclear power plant in response to external hazard-induced station blackout accidents

Natural hazard-induced nuclear accidents, such as the Fukushima Daiichi Accident that occurred in Japan in 2011, have significantly increased reactor safety studies in understanding nuclear power plant (NPP) responses to external hazard events such as earthquakes and floods. Natural hazards could cause the loss of offsite power in nuclear power plants, potentially leading to a Station Blackout (SBO) accident that significantly contributes to the overall risk of nuclear power plant accidents. Despite the fact that extensive research has been conducted on the station blackout accident for nuclear power plant, further understanding of these events is needed, particularly in the context of the dynamic nature of external hazards such as external flooding. This paper estimates the progression of station blackout events for a generic pressurized water reactor (PWR) in response to external flooding events. The original RELAP5-3D model of the Westinghouse four-loop design pressurized water reactor was adopted and modified to simulate the external flood-induced station blackout accident, including the short-term and long-term station blackout scenarios. A sensitivity analysis of long-term station blackout, examining reactor operation times and analyzing key parameters over time, was also conducted in this work. The results of the analyses, especially the critical timing parameters of key event sequences, provide useful insights about the time during the external flooding event, which is important for plant operators to make timely decisions to prevent potential core damage. This paper represents significant progress toward developing an integrated risk assessment framework for further identifying and assessing the effects of the critical sources of uncertainties of nuclear power plant under external hazard-induced events.

Technical feasibility of powering U.S. manufacturing with rooftop solar PV

The use of renewable electricity is vital for the decarbonization of industry. Industrial firms source renewables through off-site power purchase agreements or on-site installations, though the latter currently supplies <0.1% of industrial electricity demand in the U.S. Manufacturing buildings typically have large, flat rooftops that are ideal for solar photovoltaic (PV) arrays. This study investigates the feasibility of using rooftop solar PV to cover the net annual electricity needs of industry across all U.S. states and manufacturing sectors. Modeled electricity supply intensity for solar PV arrays is compared with the electricity demand per unit of floorspace for average manufacturing buildings derived from the U.S. Department of Energy Manufacturing Energy Consumption Survey. Results show that, depending on geographical location, rooftop solar PV can completely fulfill the electricity requirements of 5%–35% of manufacturing sectors considered on a net annual basis (assuming net metering). Furniture, textiles, and apparel manufacturing can be powered through on-site means in nearly every location, representing 2% of U.S. manufacturing electricity use and 6% of floorspace. Considering seasonal potential during summer months expands the list of feasible sites, particularly in the U.S. Southwest. Compared to off-site sourcing of renewable energy, pursuing on-site PV installations can also enable manufacturers to maintain limited operations during periods of grid disruption, especially when coupled with on-site energy storage. Overall, the results indicate a substantial physical opportunity for industrial firms to expand rooftop solar PV from currently low levels to help meet decarbonization goals.

The Method of Calculating the Frequency of the Initiating Event in a Dual-Unit Site with the Example of LOOP Events

In a nuclear power plant, the consequences of a multi-unit event occurring concurrently are more serious than those of a single-unit event. The first step in the probabilistic safety analysis of multi-units is to analyze the initiating events and calculate the frequency of initiating events for simultaneous events of multiple units. The difficulty in using the fault tree model is that the known data are all frequency data from a single unit and cannot be logically multiplied. In this paper, taking a dual unit as an example, we used the formula to convert the probability of failure of the second unit within 72 h and then build a fault tree model. After analyzing the results of the dual unit, the most frequent cut set was the common cause of failure of the main transformer and of the switching failure of the main and auxiliary external power. The final calculation of the frequency of simultaneous loss of off-site power events for the dual units within 72 h was 3.22 × 10−4/year. After comparing with the single-unit results, it was found that the common cause failure of each unit’s independent equipment was the main reason for the occurrence of a loss of off-site power. Shared equipment in a single unit was ranked low in all the cut sets (such as the stability of the external grid for the main and auxiliary power systems) but was ranked high in multiple units. The calculation results of the frequency of initiating events of double units were two orders of magnitude lower than those of a single unit. However, the consequences of simultaneous events of multiple units were higher than those of single reactors. Therefore, attention should be paid to the risk of a simultaneous loss of off-site power event of multiple units.

Transient analysis of a locked rotor/shaft seizure accident involving the EU-DEMO WCLL Breeding Blanket primary cooling circuits

The EU-DEMO Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) main subsystems to be cooled are the Breeder Zone (BZ) and the First Wall (FW). Each subsystem will be equipped with an independent Primary Heat Transfer System (PHTS). Within the framework of the EUROfusion Work Package Breeding Blanket research program, several accidents belonging to the category of “Decrease in Coolant System Flow Rate” were studied. The activity was aimed at evaluating the blanket and primary cooling systems thermal-hydraulic performances during such transient conditions. A complete model including the BB and related PHTS circuits has been developed at Sapienza University of Rome. A modified version of RELAP5/Mod3.3 system code has been used to perform the calculations. The simulation results showed that a locked rotor/shaft seizure of a BZ or a FW main coolant pump is the most challenging scenario. BZ and FW system behavior has been analyzed following this initiating event with the goal of the design improvement and to individuate the need for preventive measures. The influence of loss of off-site power on the accident evolution has also been investigated. Moreover, management strategies have been proposed for different reactor components. Calculations demonstrate that the current blanket and PHTS design is appropriate to cope with these kinds of accident scenarios.

Verification of the NACAC atmospheric dispersion calculation using a hypothetical accident in a neighboring nuclear power plant

Calculating atmospheric dispersion in a hypothetical nuclear accident plays an essential role as a predictive tool in nuclear emergency planning. The Nuclear Accident Consequence Analysis Code (NACAC) has been developed as an in-house code by the Thailand Institute of Nuclear Technology to develop its capability for understanding the transboundary radiological effect of a hypothetical nuclear accident. To verify the performance of the NACAC, a comparison between the calculated results using the NACAC with those from the well-developed Java-based Real-time On-line Decision Support (JRODOS) code is performed with representative cases of all seasons in a year. Two hypothetical severe accidents at the Fangchenggang pressurized water reactor, namely, (i) a loss of off-site power and (ii) a large-break loss of coolant, are utilized as the initial boundary conditions. The air concentration characteristic maps obtained from both codes are analyzed, revealing that the dominant radionuclide dispersion pathways are similar in all study cases. The air concentration, ground concentration, and total effective dose rate at several locations up to 500 km away from the release point are directly compared. Typically, the calculated results from the two codes are in reasonable agreement. However, for specific locations and directions, the NACAC predictions of the concentrations and total effective dose rate are lower than those from the JRODOS. These differences may be influenced by the different trajectory calculations between the two codes, which resulted in varying plume densities in certain areas. However, it does not significantly affect the radiation effect assessment. The NACAC still provides a total effective dose rate corresponding to the JRODOS, lower than one mSv throughout the main dispersion direction.

Advanced Safety Programmable Logic Controller and Operational Programmable Logic Controller for Data Acquisition Control System of SADHANA Sodium Loop

The SAfety Decay Heat Analysis in NAtrium (SADHANA) sodium loop is designed with two PLC controllers. All safety control interlocks are monitored by a Safety programmable logic controller (SPLC), and temperature control, including averaging ON/OFF heaters, is managed by an Operational PLC (OPLC).The decay heat exchange system of the SADHANA sodium loop is evaluated using these combined safety and operational PLCs. In the case of reactor loss of offsite and onsite power, decay heat generated in the reactor core after shutdown is removed by four Safety Grade Decay Heat Removal (SGDHR) systems [1,2,3,4,5]. Each exchanger (AHX) is located outside the reactor containment building, and a sodium loop circulation is created by natural convection [6,7,8,9,10]. The purpose of construction of SADHANA loop, with simulated models of DHX and AHX to study the performance of the SGDHR system, has been planned [11,12,13,14,15]. The SADHANA sodium loop has been commissioned, operating at isothermal temperatures of 300°C and hot temperatures of 500°C/& 550°C. The smart instrumentation and control system comprising Loop Operational Data Acquisition and Control System (LODACS) and Experimental Data Acquisition System (EDAS) are used in the SADHANA sodium loop. During the tests, LODACS personal computer and Instrumentation panel monitor loop parameters such as sodium level, leak, and pressure. In SADHANA loop DHX and AHX temperature decay heat changes monitoring by EDAS. The temperature of the pool in Test vessel 4 (TV-4) is controlled by immersion type sodium heaters via PID controllers. Currently, the loop consists of 64 surface heaters, 30 immersion type sodium heaters, and 205 thermocouples (84 Nos. of operational thermocouples and 121 Nos. of experimental thermocouples). According to the experimental thermocouples are being monitored by a separate fast data acquisition (scan rate of 10 seconds) by in-house developed software. This technical paper also describes the implementation of PLCs in different data acquisition control systems.