An accident scenario-based statistical test method for RPS in nuclear power plant

UK version of the Hua-long Pressurised Reactor (UK HPR1000), designed by China General Nuclear Power Corporation (CGN), has been successfully completed the Generic Design Assessment (GDA), with the issue of a Design Acceptance Confirmation (DAC) from Office for Nuclear Regulation (ONR) and a Statement of Design Acceptability (SoDA) from the Environment Agency on 7 February 2022. In the UK GDA process, one of the main fundamental requirements is that the nuclear safety risk should be reduced to ALARP (As Low As Reasonably Practicable). During the UK HPR1000 design process, Probabilistic Safety Assessment (PSA) has been widely used as a key tool to inform the design and evaluate risk levels, and to demonstrate that the nuclear safety risk has been reduced to ALARP. The process of PSA ALARP demonstration is established in accordance with the methodology that reflects the UK context requirements and consistent with relevant good practice. The process is fully applied by using PSA to systematically review the risk insights and to support other topic areas. After the systematic review of PSA risk insights from internal events Level 1 PSA, internal fire Level 1 PSA, internal flooding Level 1 PSA, external hazards Level 1PSA, external flooding Level 1 PSA, seismic PSA, Spent fuel pool PSA and Level 2 PSA, the related conservatisms, potential design improvements have been identified and addressed. Current UK HPR1000 PSA results and risk insights show that the risk level of UK HPR1000 design is relatively low and the risk of UK HPR1000 design has been reduced to ALARP in GDA stage.

Software verification and validation (V&V) is internationally recognized as an important technology to improve software reliability. The promulgation of new regulations and standards related to nuclear safety software may put forward new and higher requirements, so that existing software V&V technical solutions cannot fully cover the new requirements. There is still a gap between the new requirements of the new regulations, the new standards and the software V&V experience feedback on issues of Generic Design Assessment (GDA) of Office for Nuclear Regulation (ONR). Therefore, relevant research of software V&V is needed so as to meet the new requirements for domestic and international safety reviews. Based on the comparative analysis of the new and old nuclear safety standards, such as Institute of Electrical and Electronic Engineers (IEEE) 1012 and International Atomic Energy Agency (IAEA) No.SSG-39 and ONR review principles, as well as the technical opinion report of European Union (EU) safety software certification, the main technical differences were sorted out to provide technical reference for the establishment of better applicability or the optimization of the nuclear safety I&C system software V&V solution.

Environmental contamination with radioactive materials caused by the Fukushima Dai-ichi Nuclear Power Plant (NPP) accident in 2011 raised a serious health concern among residents in Japan, and the demand for radiation experts who can handle the radiation-associated problems has increased. The Human Resources Development Center (HRDC) of the National Institute of for Quantum and Radiological Science and Technology in Japan has offered a variety of training programs covering a wide range of technologies associated with radiation since 1959. In this study, the time-course change in the number and age of the applicants for training programs regularly scheduled at HRDC were analyzed to characterize the demand after the NPP accident. The results suggested that the demand for the training of industrial radiation experts elevated sharply after the NPP accident followed by a prompt decrease, and that young people were likely stimulated to learn the basics of radiation. The demand for the training of medical radiation experts was kept high regardless of the NPP accident. The demand for the training of radiation emergency experts fluctuated apparently with three components: a terminating demand after the criticality accident that occurred in 1999, an urgent demand for handling of the NPP accident, and a sustained demand from local governments that undertook reinforcement of their nuclear disaster prevention program. The demand for the training of school students appeared to be increasing after the NPP accident. It could be foreseen that the demand for training programs targeting young people and medical radiation experts would be elevated in future.

A combination of Convolutional Neural Network (CNN), Long-Short Term Memory (LSTM), and Convolutional LSTM (ConvLSTM) is constructed in this work for the fault diagnosis and post-accident prediction for Loss of Coolant Accidents (LOCAs) in Nuclear Power Plants (NPPs). The advantages of ConvLSTM, such as effective feature determination and extraction, are applied to the classification of LOCA cases. The prediction accuracy is enhanced via the collaborative work of CNN and LSTM. Such a hybrid model is proved to be functional, accurate, and adaptive, offering quick accident judgment and a reliable decision basis for the emergency response purpose. It then allows NPPs to have an Artificial Intelligence (AI)-based solution for fault diagnosis and post-accident prediction.

Reliable evaluations of the radiological consequences of hypothetical severe accidents in nuclear power plants (NPPs) play a key role in supporting the emergency and preparedness plans during such abnormal events. Also triggered by the growing requests of the national regulators, one of the milestones of the Nuclear Safety Research program (NUSAFE) of the Karlsruhe Institute of Technology (KIT) is the improvement of the performance of the codes employed for such evaluations in the current and innovative NPPs, which are in operation and going to be built in EU and worldwide. Having this in mind, a calculation platform of state-of-art codes has been assessed at KIT to evaluate the accident progression, the radiological source term, and the behavior of the fission product dispersion to the environment by also taking into account the corresponding uncertainties. In particular, the CASMO5 lattice physics code, developed by Studsvik Scandpower, is used for a proper evaluation of the fuel inventory loaded in the reactor core. The European reference Accident Source Term Evaluation Code (ASTEC), developed by IRSN (France), is employed to evaluate the accident progression from the initiation to the fission produced release to the environment. The use of the KArlsruhe Tool for Uncertainty and Sensitivity Analysis (KATUSA), developed by KIT, aims at performing the uncertainty quantification of the source term. Finally, the Java Real-time On-line DecisiOn Support (JRODOS) is employed for analyzing the radiological consequences of the accident. The KIT platform has been employing since some years in the frame of the KIT participation to different European projects, i.e., MUSA, SASPAM-SA on SMRs, SANE. In the paper, an overall view of the KIT strategy on severe accident analyses will be provided, and examples of the application of the above-mentioned platform to generic NPPs will be provided, i.e., PWR Konvoi.

Accident management support tools in nuclear power plants: A post-Fukushima review

In the situation of nuclear accidents of nuclear power plants(NPP), the protective actions should be taken by the public are ascertained by Generic Intervention Level(GIL) and the controls should be placed in food are ascertained by Generic Action Level(GAL). GIL and GAL were not designed to be used during an emergency; they cannot be promptly measured in the field and do not address facility conditions. Operational Intervention Level(OIL) which can easily be measured during an emergency, and on which the need for protective action can be rapidly ascertained, was introduced and described in detail by International Atomic Energy Agency(IAEA). According to the approach of calculation and revision proposed by IAEA, the adaptive OILs are studied and given in each nuclear power plant in China. Further, software system for calculation and revision of OILs and determining protective actions have been also established and applied in these NPPs. After Fukushima nuclear accident, emergency response preparedness for multi-reactor accidents is taken seriously. As the generic assessment procedures for determining protective actions during a reactor accident based on OILs mentioned above are not fully applicable to the situation of multi-reactor accident, it is necessary to optimize the approach of calculation of OILs and determining protective actions for multi-reactor accident. The paper first introduces and summarizes the basic concept and approach of OIL during a reactor accident. In this approach, environmental data are assessed primarily through the use of OILs which are quantities directly measured by the field instruments. The default OILs which have been calculated in advance on the basis of the characteristics of severe reactor accidents are used to assess environmental data and take protective actions until sufficient environmental samples are taken and analyzed to provide a basis for their revision. This approach allows data to be quickly evaluated, and decisions on protective actions to be promptly made. The paper studies the essence of OILs during single reactor accident and presents the method for OILs calculation during multi-reactor accident. The core idea of the method is to evaluate the environmental radioactive consequence of each single reactor accident which can be overlaid together finally during a multi-reactor accident. Using the final environmental radioactive consequence, the suitable OILs can be calculated in the original calculation formula. As the default assumptions are not related to characteristics of reactor and site, the default OIL5 of IAEA is suitable for multi-reactor accident. The default of other OILs (OIL1$1 \sim$OIL4, OIL6$6 \sim$OIL9) can be calculated in this optimized method. The paper also indicates that the approach of revision of OILs during a single reactor accident is also suitable for multi-reactor accident without optimizing. In this method, the procedures and strategies of determining protective actions are also optimized. Different from sector areas based on distance and direction during a reactor accident, regions based on administrative boundaries and natural boundaries are considered as unified units for taking a determined protective action. On the basic of the optimized method, the corresponding software system is also designed and implemented. Suitable for both single-reactor and multi-reactor accident, the software provides complete process for use of OILs which include the function of data management, environmental data input, calculation, revision and decisions on protective actions. The optimized method and system in this paper have been applied to several NPPs in China, e.g. Qinshan, Ningde, Sanmen, Taishan and Yangjiang. In the application, the method has been verified and the system has also been constantly improved.

On 11 March 2011, Japan was struck by an earthquake measuring over 9·0 on the Richter magnitude scale, the strongest known to hit Japan and one of the top five largest earthquakes ever recorded in the world. The epicentre was 175 km (110 miles) east-north-east from the Fukushima Daiichi nuclear power plant, and within an hour a massive tsunami had inundated the site. This caused a serious nuclear accident, with an International Nuclear and Radiological Event Scale (INES) rating of Level 7 (the highest). Following the events at Fukushima, the nuclear industry in the UK responded quickly to review UK nuclear installations against seismic and flooding hazards. On 14 March 2011 the UK Secretary of State for Energy and Climate Change requested that Dr Mike Weightman, HM Chief Inspector of Nuclear Installations, examine the circumstances of the Fukushima accident to see what lessons could be learnt to enhance the safety of the UK nuclear industry. An Interim Report was published in the middle of May 2011, with a Final Report six months later. A further report was published in October 2012 that provided an update on progress in implementing the lessons for the UK’s nuclear industry. In addition to the Weightman Reports, the UK Office for Nuclear Regulation (ONR) produced two national reports on the European Council ‘Stress Tests’ focusing on licensed nuclear sites. The first covered all civil nuclear power plants with the second, on the instructions of the Chief Inspector, covering all of the remaining UK nuclear installations. In both these reports, areas for potential improvement (known as ‘considerations’) were identified by licensees and these were augmented by Stress Test Findings identified by ONR. The key findings of these reports are reviewed, with the objective of extracting and drawing out key learning which could translate to and impact nuclear transport safety. The potential impact of these findings on INS and its nuclear transport operations is assessed, and the progress made to date reported.

Emergency evacuation time model of nuclear power plant pedestrians

[This corrects the article DOI: 10.1371/journal.pone.0172442.].

Three severe nuclear power accidents (Fukushima accident, Chernobyl accident and Three Mile Island accident) have proved that human errors in the mitigation of severe accidents will exacerbate the deterioration of accidents and cause immeasurable consequences. Therefore, there is an urgent need for finding human reliability analysis methods to effectively analyze and predict possible human errors in severe accident mitigation.In order to compare the applicability of the Human Reliability Analysis (HRA) method under the severe accident in the nuclear power plant, this paper put forward four criteria, qualitative analysis, quantitative analysis, traceability and availability, based on the characteristics of the personnel in the process of severe accident mitigation. Several commonly used HRA methods were compared and analyzed. The results show that the common HRA methods which applied to the analysis have shortcomings, so it is necessary to establish a new HRA method, or to improve the original methods to solve the issues under severe accident conditions and the IDAC model and Phoenix HRA methodology are considered better choice. Combined with computer programs,the IDAC model can simulate the process of accident and the dynamic response of personnel, which helps to explain the dynamic interaction characteristics of human behavior and environment, but the model still has some limitations in the aspects of uncertainty analysis and simulation data availability. While Phoenix method provides some Performance influencing factors and crew failure modes that can reflect the response characteristics of emergency personnel. This applicability study can provide guidance for the establishment of HRA method in severe accidents.

The building of more nuclear power plants is one of the proposed solutions to the increasing production of anthropogenic CO2 and the mitigation of the global warming threat. Had the USA constructed an additional 56 nuclear power plants in the 1990s, the country would have been in compliance with the Kyoto protocol. Nuclear power plants are typically very large (1,000 MW) and very complex power producing units. The nuclear reactor itself contains large amounts of radioactive materials, which if released in the environment, may cause large-scale environmental accidents. For this reason, a nuclear reactor must have multiple levels of safety systems and its controls must be designed to shut down the reactor within a very short time. All nuclear reactors have six basic components, which are described in detail in this chapter. Several types of nuclear reactors and nuclear power plants have been developed by different countries. The special design characteristics and the operation of these plants are summarized. Because accidents in nuclear power plants are feared the most by the public, the three accidents that received the highest notoriety, at the Three-Mile Island in the U.S.A., at Chernobyl in the former Soviet Union and at the Fukushima Dai-ichi power plant in Japan are described in detail. The causes of the accidents are examined and early actions that could have been taken by the operators are presented. Finally, it is apparent that, if the world is to rely on nuclear energy in the long-term, the more abundant uranium-238 and other fertile nuclear materials must be utilized. This makes necessary the use of breeder reactors, which may become the next generation of nuclear reactors.KeywordsNuclear Power PlantFuel ElementFuel AssemblySteam GeneratorNatural UraniumThese keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

Reactor protection system (RPS) plays critical role in digital control system (DCS), and ensures the safety for nuclear power plant (NPP). System test is a necessary step during system development, verification and validation (V&V), which ensure the safety and reliability for RPS. The debugging of test environment and equipment is an important step that ensures the effective and efficiency of test. The system, such as RPS that contains complicated logic and large number of interfaces, cost a lot of time and human resource for debugging. A structured debugging method has been proposed in this paper, this method establishes debugging architecture with a hierarchical model in according to signal transmission path, and it designs the debugging process from down to top. The result from engineering practice show that this method has improved the effective and efficiency of debugging provides the support and reference for system test environment establishment.

Background. Endocrine comorbidity, a concomitant radiation injury of several endocrine glands, is a topical issue in radiation medicine. The purpose of the study was to establish a correlation of secondary hyperparathyroidism with thyroid and pancreatic disorders among the Chornobyl nuclear power plant (NPP) accident survivors. Materials and methods. A blind sample (n = 60) of subjects treated at the radiation endocrinology department in 2019 was selected for the study. Study sample included the Chornobyl NPP accident survivors (group 1, n = 40) and patients not exposed to radiation (group 2, n = 20). The average age of people was 60.2 ± 9.8 years. Decompensation of type 2 diabetes was the most common cause for hospital admission (48.3 %), decompensation of hypothyroidism was in second place (46.7 %). Research methods: clinical, instrumental, laboratory, statistical. Results. Incidence of the increased risk of vitamin D deficiency and excessive parathyroid hormone production in the Chornobyl NPP accident survivors was somewhat higher than in individuals not exposed to radiation (81.08 and 78.12 %; χ2 = 0.257, p = 0.612; 31.12 and 28.08 %; χ2 = 0.462, p = 0.319, respectively). Using a multivariate analysis, the reliable model was obtained confirming the relationship of increased parathyroid hormone level and early subclinical signs of the target organ damage (F = 4.294; p = 0.042), which proves the relevance of using clinical questionnaires. A reliable positive correlation was found between the vitamin D and free thyroxine levels (r = 0.729; p = 0.001), and negative one was found with the thyroid-stimulating hormone content (r = –0.803; p = 0.001). In the Chornobyl NPP accident survivors with type 2 diabetes, a decreased vitamin D concentration correlated with an increased duration of metformin therapy (r = –0.421; p = 0.001) and body mass index (r = –0.787; p = 0.001). It was found that the parathyroid hormone level increases along with the duration of type 2 diabetes and an increase in creatinine clearance (r = 0.724; p = 0.001, and r = 0.799; p = 0.001). Conclusions. Vitamin D deficiency has a negative effect on the course of comorbid disorders (type 2 diabetes and hypothyroidism) and excessive production of parathyroid hormone.

Changes in the carbon mitigation responsibility of Japan's capital city, Tokyo – analysis of power supply shocks due to nuclear power plant accidents