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Improper operation leading to an emergency shutdown in a nuclear power plant can result in significant energy loss and have a substantial impact on the economy. This study focuses on a sodium-cooled fast reactor (SFR) nuclear power plant, analyzing system characteristics under different accidents through feedwater system simulation and improving control system performance. The results show that the single module accident will affect the whole feedwater system, and the influence of faulty SG on the loop signal should be eliminated. The improved control system reduces sodium temperature fluctuation by 3%–41.5%, shortens adjustment time by 57.9%, and ensures standby pump activation within 10 s after the feedwater pump trips or when the system stabilizes. Although the feedforward of mass flowrate ratio can improve the response of the system to flow disturbance, it resists the temperature feedback. However, considering inlet temperature's impact on feedwater demand, this confrontation can be effectively avoided, and a good control effect can be maintained during accidents. • The third circle accident of sodium-cooled fast reactor is simulated. • An improved feedwater control system, based on energy balance, is proposed. • Recommendations for addressing different types of accidents in the third circuit are provided. • The performance of feedwater system under accident has been significantly improved. [ABSTRACT FROM AUTHOR]

Regulators around the world have devised practical methods for evaluating the safety of nuclear power plants (NPPs) and increasing public awareness of plant safety. One essential approach is the regulatory safety-related inspection, which plays a crucial role in confirming the safety of power plants to prevent and mitigate accidents. In recent years, many developed countries have adopted a risk-informed methodology, which, similar to the conventional deterministic approach, prioritizes various aspects of regulatory inspection which relies on PSA insights for decision-making, rather than being rooted in a deterministic approach. In the United Arab Emirates, where three nuclear units are in operation, the regulatory inspection focuses on reducing the likelihood of incidents through a defense-in-depth (DiD) concept and the insights gained from risk analysis. To strengthen accident prevention oversight, both deterministic and probabilistic approaches are utilized, including probabilistic risk assessments (PRA) aimed at identifying safety improvements. This study aims to develop a preliminary safety inspection methodology using AIMS-PSA as an analysis tool. It involves constructing event trees and fault trees to understand how to secure the success path of Loss of Coolant Accidents (LOCAs) transients, block potential failure paths, and assess their impact on Core Damage Frequency (CDF) by analyzing the generated cut sets. This study will offer quantifiable insights into key risk factors, improving our capacity to anticipate and prevent accidents, maintain high reliability, and conduct effective risk assessments for the continued operability of safety systems. While previous investigations have explored nuclear accident frequencies, often with limitations, this research aims to provide a more comprehensive perspective on nuclear power plant safety, addressing a wide range of accident scenarios and their implications. • Regulatory safety inspections are crucial for verifying nuclear power plant safety. • The UAE's focus on minimizing incident probability using a defense-in-depth approach. • AIMS-PSA is used to analyze LOCA transient success and failure paths. • The study identifies core contributors to risk and aids in accident anticipation. • The methodology enhances safety, reliability, and long-term accident mitigation. [ABSTRACT FROM AUTHOR]

In order to further clarify the fire safety evaluation of nuclear grade activated carbon, thermal analysis combined with pore structure and phase analysis was adopted for coal-based activated carbon and coconut shell activated carbon used in nuclear power plants. On the basis of establishing the relative fire safety evaluation model by using four important parameters: activation energy (E), maximum mass loss rate (DTG p), maximum mass loss rate temperature (T p) and rapid reaction percentage (A 12), the O/C ratio and enthalpy (H) were further introduced into the evaluation model to calculate the relative fire safety factor (RFSF). The results show that O/C ratio and enthalpy have greater influence on the fire safety of activated carbon than porosity characteristics. The RFSF values of coal-based and coconut shell activated carbon are 1.33 and 1.08, indicating that fire safety is higher than that of impregnated coconut shell activated carbon. • A relatively perfect fire safety evaluation model of nuclear grade activated carbon was successfully established. • The O/C ratio and enthalpy have greater influence on the fire safety of activated carbon than porosity characteristics. • RFSF of CBAC and CSAC are 1.33 and 1.08, indicating that fire safety is higher than that of ICSAC. • The introduction of O/C ratio and enthalpy will further improve the evaluation model of relative fire risk. [ABSTRACT FROM AUTHOR]

The Fukushima nuclear accident poses a new challenge for Probabilistic Risk Assessment (PRA) to consider the integrated risk of a nuclear power plant having more than one units. That calls for new technologies in the area of accident sequence modelling which traditionally deals with the accident in one reactor only. It has been shown that multi-unit event trees tend to be huge and consume a lot of manpower. This paper presents a rapid approach to generate multi-unit event trees based on tree combinational calculation, which leverage the existing event trees that have been built in single unit PRA. The combinatorial calculation is driven from the vector representation of event trees and can screen out the invalid combinations automatically by applying the proposed principles. A case study is presented in this paper with a nuclear power plant having 2 reactor modules under the Loss-Of-Offsite-Power initiating event to demonstrate the feasibility of the approach. The proposed approach has been realized by the software MUETBASE which can significantly improve the working efficiency and correctness of multi-unit event tree modelling. • The proposed method can help to evaluate the risk of plants with multiple units. • A multi-unit event tree generation method by combining single-unit ones is proposed. • The combinatorial calculation is realized via a vector representation of event tree. • Principles for screening valid combinations are provided. • A case study is also presented to demonstrate the feasibility of this method. [ABSTRACT FROM AUTHOR]

To improve the emergency disposal capacity of radioactive sewage entering into the ocean, concentration distribution and hazardous area were observed and calculated. Haiyang nuclear power station and its surrounding coastal waters were selected as the research object. Based on Euler method, calculation model of concentration distribution and hazardous area with radioactive sewage entering into the ocean was established with a resolution of 1′ × 1'. Radionuclide concentration distribution and hazardous area were calculated under an assumed nuclear leakage accident and an assumed radioactive sewage discharge event with a plan. The study show that, under the given conditions of an assumed nuclear leakage accident, radionuclides of 137Cs are mainly distributed within about 81 km east, 39 km south and 33 km west of the nuclear power station after fourteen days. While under the given conditions of an assumed radioactive sewage discharge event with a plan, radionuclides of 3H are mainly distributed within about 58 km east, 14 km south and 23 km west of the nuclear power station after fourteen days. Compared with the total amount of source term, the high concentration is reduced by about 8 orders of magnitude. Area with radioactivity lower than 100 Bq/L is considered to be safe area, that more than 100 Bq/L（105Bq/m3）to be hazardous area, then the hazardous area account for about 50% of the main radionuclide distribution area. The hydrodynamic result was verified, and the maximum error between calculation and measurement velocities in flow field was less than 10%. The radioactive result was verified, and changes of radionuclide concentration in this research basically agreed with the monitoring data from the Fukushima Daiichi nuclear accident. So the model is considered to be reasonable and credible. [ABSTRACT FROM AUTHOR]

In this study, we have investigated the spray system's capabilities to reduce fission products in the containment atmosphere and control the containment pressure in a VVER-1000 nuclear power plant. Large Break Loss of Coolant Accident (LB-LOCA) without access to the active part of the Emergency Core Cooling System (ECCS), as a worst-case severe accident, has been investigated as a case study. Only containment is simulated via the MELCOR code, while leakage of various materials from the nuclear power plant's core and the primary circuit is considered as the boundary conditions. Following the Fukushima accident in 2011, many nuclear safety regulatory organizations focused on using portable equipment that is essential for decay heat removal in nuclear power plants during severe accidents. Based on the literature studies, the accident management scenarios have changed recently since severe accidents inside nuclear power plants have worried many scientists and the public. Evaluating the current spray system in a severe condition based on which this system was not designed is vital and novel. We considered that the spray system would be fully restored and perform its safety function using the existing portable equipment to depressurize the containment and deposition of the fission products. To calculate the maximum allowed delay time of the spray system recovery, the accident was modeled without the operation of the spray system while a safety criterion was considered to avoid reaching the ultimate pressure of containment. For a more detailed evaluation of the spray system, two different operating modes including the activation of one spray train and the activation of both spray trains have been selected. According to the obtained results, if the spray system is revived before 4,000s from the beginning of the accident, this system would be able to reasonably mitigate the accident consequences. Moreover, without the need for any external water supply source for the spray system, this system can be permanently fed from the containment sump for at least 800,000s after the initiation of the accident, requiring the relevant heat exchanger of the spray system to be recovered for this period. • Containment is modeled by the MELCOR code to investigate the ex-vessel phenomena. • Design-based LB-LOCA is modeled to validate the containment and spray system model. • Spray system capabilities during a combination of LB-LOCA and SBO are evaluated. • The spray system can preserve the containment integrity during the severe accident. [ABSTRACT FROM AUTHOR]

We conducted a review on the application of two-phase thermosyphons in passive residual heat removal systems (PRHRs). Thermal management to handle the heat of decay in nuclear power plants is crucial for maintaining the integrity of the reactor core. Nuclear power plants are equipped with decay heat removal systems under normal and accident conditions. These systems are generally known as residual heat removal systems. The technology of decay heat removal that relies on electric power, known as an active system, will become ineffective when there is a total loss of electrical power for an extended period (e.g., extended station blackout (SBO)), such as the accident at the Fukushima Daiichi Nuclear Power Plant. To deal with SBO accidents, the design of Generation III+ and small- and medium-size nuclear power plants already includes PRHR systems, which is designed to function for at least 72 h without operator intervention. On the other hand, the use of two-phase thermosyphons as part of thermal management in nuclear power plants is also increasing. This review aims to determine the current status and challenges of the PRHR system design and the application of two-phase thermosyphons in PRHR design. The tools and models used for analyzing PRHR designs are also included in this review. Therefore, a review of thermal-hydraulic performance PRHR that uses the two-phase thermosyphon has been prepared based on simulation and experimental approaches found in the literature. This review focuses on the design and analysis of PRHR systems for light water reactors (LWR) and the usage of two-phase thermosyphon as cooling devices for nuclear-spent pools and non-transient cooling for both LWR and non-LWR reactors. We highlight the PRHR system design and the application of two-phase thermosyphons in PRHR design. As major contributions, different configurations of long-term cooling of PRHRs are discussed. It has been found that to extend the working life of PRHR beyond 72 h, several researchers have proposed new designs with unlimited cooling capabilities. One of the proposed technologies to support PRHR systems of this type is the sub-atmospheric two-phase thermosyphon. It has also found that RELAP5 is the most frequently used PRHR system design analysis tool. This paper, as a whole, acts as a current reference for researchers and lays the groundwork for future research in the use of two-phase thermosyphon for long-term decay heat removal systems. • Passive residual hear removal systems (PRHRs) are to increase the safety of nuclear power plants (NPPs). • Currently, PRHRs are generally designed to function passively for 3 days. • Heat Pipes and two-phase thermosypons are studied as passive cooling in the NPPs. • Two-phase thermosyphons are capable to extend the PRHR's operation for indefinite. • Thermal hydraulic system codes are widely used for PRHR performance analysis. [ABSTRACT FROM AUTHOR]

Interest in evaluation of reactor response and off-site consequences following beyond design basis accidents has significantly increased after Fukushima. MELCOR is an advanced computational aid that has been widely used and adapted to various reactor designs for severe core damage analyses. This paper summarizes the core thermal hydraulic response for a hypothetical severe accident caused by station blackout with failure of the steam generator safety relief valve at a Chinese pressurized reactor 1000-MW power plant – CPR1000 using MELCOR. Analytical results for a) temperature distribution of the fuel pellets, b) the fuel cladding, c) flow rate of the coolant, and, d) hydrogen mass changing over time are presented. The analyses are focused on safety assessment of the reactor core for severe accidents and are part of the overall evaluation of safety features of the CPR1000 reactor for residual risk posed by severe accidents. [ABSTRACT FROM AUTHOR]

One month after the TEPCO accident, the number of the disaster-related deaths in Fukushima Prefecture exceeded that in Miyagi and Iwate which were not affected by the radiation, due to increase in disuse syndrome during the prolonged evacuation by the accident. About 90% of the deaths were the elderly. The risk and benefit of the evacuation of 11 municipalities in Fukushima are analyzed using data published by the Japanese government. The total loss of life expectancy (LLE) of the disaster-related deaths is 5,100,000 person-days. The LLE avoided by the evacuation is estimated from the cumulative doses of 172 locations. The average cumulative dose avoided is 205 mSv in 10 years. Total LLE avoided for 100,000 evacuees is 1,850,000 person-days. It is 2.8 times less than that accompanies evacuation. The delay in lifting the evacuation orders caused many disaster-related deaths among the elderly. The long-term evacuation of the TEPCO accident followed the dose rate recommended by ICRP, but is not justified. Discussions are provided for future improvements: (1) Lifetime fatal cancer probability from radiation exposure in the elderly is one-fifth of the average. Current guidelines do not take that into account. The difference needs to be incorporated into the nuclear emergency guidelines. (2) Radiation protection measures cannot describe other health risks, but the present guides are developed from the perspective of radiation protection. (3) Need to revise emergency guides to consider other health risks such as disuse syndrome of Fukushima, unnecessary abortions of Chernobyl and mental and psychological problems. (4) Risk communication with the general public does not work in a nuclear accident. (5) Exempt the ALARA principle from the public in low-dose area in the event of a nuclear accident. (6) Allow evacuees to participate in decontamination and restoration work. The lesson is important for countries where nuclear power plants are located in agricultural areas and are aging. • Cost benefit analysis shows higher evacuation risk than avoided at TEPCO accident. • Delay in lifting evacuation orders increased deaths of the elderly in Fukushima. • Consider low life time fatal cancer risk of old people in protective action guides. • Revise nuclear emergency guides to consider public health risks other than radiation. • Exempt public from as low as principle of radiation protection in a nuclear accident. [ABSTRACT FROM AUTHOR]

Abstract: A key challenge for engineers and scientists over the coming decades is to develop and deploy power plants with sufficient capacity and flexibility to meet the growing demand for energy (mainly electrical) whilst simultaneously reducing emissions (primarily greenhouse gases). With fusion-based power plants not currently being considered viable for large-scale deployment for at least 40 years, other technologies must to be considered. Renewable and high efficiency combined gas-fired plants, along with nuclear solutions, are regarded as the most suitable candidates, with Small Modular Reactors (SMRs) developing as a favoured choice. However, two main impediments to the current deployment of SMRs exist: (1) safety concerns, particularly following the Fukushima accident, and (2) their economic models, with high capital costs only being available through a limited number of investors. The goal of this paper is to provide a review and a holistic assessment of this class of nuclear reactor, with specific focus on the most common technology: the Light Water Reactor (LWR). In particular, the paper provides a state-of-the-art assessment of their life cycle, along with a comparison of their relative merits with other base-load technologies. It is shown that SMRs are a suitable choice when the power to be installed is in the range 1–3GWe and the social aspects of the investment, such as the creation of new employment positions, is a goal of policy makers. The paper thereby provides governments and stakeholders with key economic and social boundaries for the viable deployment of SMRs. [Copyright &y& Elsevier]

Chernobyl and Fukushima nuclear accidents showed that nuclear accident is a global concern. Dispersion and faraway transportation of radioactive materials under atmospheric conditions are the main goals of this paper. Radioactive materials have a lot of damages, undesirable and dangerous long-term effects on the environment and human health. Therefore, in this paper, the simulation of transport, dispersion and deposition of radioactive materials released from a postulated accident in Bushehr Nuclear Power Plant have been studied using HYSPLIT. The postulated accident for Bushehr Nuclear Power Plant is designed based on world's largest nuclear accident, the Chernobyl accident. In this study, it is assumed that the postulated accident in Bushehr nuclear plant occurred on March 21, 2015. The simulation results show that in the period of one month, nuclear pollution resulting from this accident would affects almost the whole of Iran. The amount of radioactive materials deposition is very high in some cities of Iran. The most deposition of radioactive materials would occur in Bandar Bushehr and Ahvaz. In these cities, the maximum deposition is related to the second day of the accident, which are: 1.55 × 10 8 B q / m 2 and 1.06 × 10 6 B q / m 2 , respectively. During this period, in addition to Iran, even most Asian countries, North America, Europe continent and Africa are affected by nuclear pollution while the most pollution is related to Asia, North America and Europe, respectively. Radioactive materials after passing from the continents come back to the East Asian again. In Asian countries after Iran, the maximum deposition of radioactive materials is related to Turkmenistan that the deposition value is 1.78 × 10 5 B q / m 2 for the second day of the accident. In continent of America, the most nuclear pollution is related to the state of Arizona. In this state, the maximum deposition is 1.00 × 10 5 B q / m 2 on the fifteenth day of the accident. In Europe, the maximum deposition is in Ukraine and it is 4.16 × 10 3 B q / m 2 on 12 days after the accident. In this accident, radioactive materials would contaminate the most of Northern hemisphere and also would influence the southern hemisphere in very negligible amount. [Display omitted] • Simulation of postulated accident helps managers to find the best solution. • Dispersion and faraway transportation of radioactive materials under atmospheric conditions are very important. • The simulation of radioactive materials released from a postulated accident have been studied using HYSPLIT. [ABSTRACT FROM AUTHOR]

In response to the Fukushima nuclear accident in Japan, the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) and other countries have conducted extensive monitoring and research. This article used an improved global atmospheric dispersion model WRF-Chem to simulate the global transport of radioactive materials released by the Fukushima nuclear. And mainly analyzed its impact on China, using the monitoring data of the environmental protection department to analyze the environmental level of atmospheric radioactivity and evaluate the model. The results showed that the radioactive plume spread from west to east, across the Pacific Ocean, the American continent and Europe, and finally, it spread to the entire northern hemisphere. China first detected I-131 in Heilongjiang Province on March 23, and until April 3, all provinces in China detected radionuclides from Fukushima nuclides. Combining the time taken by the radioactive plume to reach various parts of China, it can be inferred that the radioactive material can reach China through three routes, which may be the main reason why the leaked nuclides reach China and Europe in the same time. The global proliferation model tested by the FDNPP accident can provide a reference for the potential radiation impact of nuclear power plants, help understand the potential threat of future nuclear accidents, and provide support for nuclear accident monitoring and emergency measures. [ABSTRACT FROM AUTHOR]

After the two major nuclear accidents in history considered to be beyond-design-basis accidents (BDBA) – the Chernobyl Reactor Four explosion in Ukraine (1986) and the Fukushima Daiichi accident in Japan (2011), the United States Nuclear Regulatory Commission and other countries have included in the Final Safety Analysis Report (FSAR), a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), with significant damage to the reactor core. The Safety Analysis Report (SAR) is the most important document used by a regulatory body to assess the adequacy of plant safety at all stages of the life time of a Nuclear Power Plant (NPP), and presents the licensing basis of a plant. This study aims at comparing the progression of severe accidents during a Total Loss of Feed Water Accident (TLOFW) and a TLOFW combined with An Anticipated Transient Without SCRAM (ATWS) reactivity accident for a Pressurized Water Reactor (PWR) type NPP. This paper takes into account the results of the PSA Level 1 for the reference NPP. The calculations were performed with the deterministic computer code MELCOR version 2.2 used to analyze severe accident progression in NPP with a Light Water Reactor (LWR). The paper presents a hydrogen risk assessment – the assessment is deterministic, s a prerequisite before performing PSA Level 2. The hydrogen source is determined, too. Saphiro's diagram is applied to investigate the containment atmosphere condition (inert or combustible) for two scenarios. The main results of the study reveal that the containment integrity of the reference PWR NPP when both accidents occurred is maintained. However, the core is severely damaged. A future work will investigate mitigation strategies to limit the hydrogen risk during accident. • The representation of the reference Small Nuclear Power Plant was made to simulate different types of transients and severe accidents and subsequently to verify the performance of safety and mitigation components. • This paper results provide a better understanding of progression of in-vessel core degradation processes for both postulated accidents (TLOFW and TLOFW + ATWS) in a Small Modular Reactor (SMR). • The analysis of severe accident scenarios for a generic Small Modular Reactor (SMR) on 48MWt Pressurized Water Reactor (PWR) using the MELCOR v.2.2 code resulting in a better understanding of the phenomena involved. [ABSTRACT FROM AUTHOR]

This paper summarizes the existing framework of safety analysis and licensing of Canada Deuterium Uranium (CANDU) reactors in Canada, established by the Canadian Nuclear Safety Commission (CNSC) and discusses challenges with implementing the lessons learned from the Fukushima accident, in the nuclear safety analysis conducted for the existing older and new CANDU reactors. This paper compares the Canadian regulatory requirements with other international nuclear regulatory guidelines, in terms of dealing with Beyond Design Basis Accidents (BDBA) and Severe Accidents, and also reviews the past and present event classification principles in safety analysis, set by the CNSC. The CNSC has published regulatory documents and technical papers outlining the challenges to be addressed with respect to implementations of the Design Extension Conditions (DEC). This paper explores the current requirements and the principles of the regulatory framework in Canada, and discusses practical approaches to determine where in every accident sequence, crediting an Emergency Mitigating Equipment (EME)would be most beneficial, for example, using a portable EME for mitigating the consequences of a Severe Accident, or preventing them. A systematic examination of CANDU Probabilistic Safety Assessment (PSA) results shows that crediting mobile EME in Level 1 PSA accident sequences have challenges and may not be the most beneficial use of EME. While adding another layer of redundancy to backup support systems (electrical power, cooling water or instrument air) further reduces the frequency of a core damage accident scenario, portable EME can only have a short-term mission time, and the main benefit of using them in Level 1 PSA, is to delay the consequences and releases. This paper proposes a thorough review of both Level 1 and Level 2 PSA models and conducting sensitivity analysis results, before making decision on where to allocate EME (i.e., in Abnormal Incident Manuals (AIM), Emergency Operating Procedures (EOP) or Severe Accident Management Guidelines (SAMG). It is crucial to balance between prevention and mitigation of severe core damage and large release frequencies to maximize the safety of the plant and minimize the impact on public safety. Instead of focusing on reducing the probability of occurrence (or frequency) of accidents by allocating every available EME to Level 1 PSA, an alternative approach would be determining where to allocate them to maintain a balance between prevention and mitigation of accidents. • Illustrative process flow-diagrams of pre and post Fukushima CANDU safety and licensing framework in Canada. • Using Probabilistic Safety Assessment (PSA) results for identification of BDBA and Design Extension Condition (DEC). • Most likely BDBA (DEC) in CANDU design are Local and Widespread Fuel Damage, with no regulatory requirement limit in Canada. • Using PSA to find balance between core-damage prevention and mitigation for allocating Emergency Mitigating Equipment (EME). • Challenges for crediting EME in CANDU's safety analysis and focusing on core-damage prevention. [ABSTRACT FROM AUTHOR]

Human Reliability Analysis (HRA) studies human actions from the safety point of view, modeling the most important of them and assessing their probabilities. The Technique for Human Error Rate Prediction (THERP) is a classical HRA approach that models Human Error Probability (HEP) for a specific action regardless of the human cognitive process. Another HRA method is Human Cognitive Reliability (HCR) that expresses the non-response probability as a function of available time regarding the performance type and ignores other affecting factors. In this study, the strengths of the Deterministic Safety Analysis (DSA), THERP and HCR are integrated to form a systematic framework introduced as Cognitive Based THERP (CB-THERP) method to calculate HEP for the Nuclear Power Plant (NPP) post-accident analysis. The calculated values of HEP during the Total Loss of Feedwater (TLFW) accident reveal that a better Human-Machine Interface (HMI) condition reduces HEP. However, operators' higher training level compensates even the worst HMI. In addition, the HMI condition has the greatest impact on an averaged-knowledge Main Control Room (MCR) operator and then an expert and a novice. Dependency almost doesn't influence the worst HMI condition and it much affects highly trained operators when the HMI condition is excellent. Also, the calculated HEP values from the Standardized Plant Analysis Risk-Human Reliability (SPAR-H) and CB-THERP methods for different cases of the scenario have similar trend while THERP calculates a fixed value for them. • Human Error Probability Quantification using the Cognitive-Based THERP method. • Dependency almost doesn't influence the worst HMI condition. • HMI affects highly trained operators when the HMI condition is excellent. • SPAR-H and CB-THERP have a similar trend while THERP calculates a fixed value. [ABSTRACT FROM AUTHOR]