Simulation Modeling and Probabilistic Safety Analysis in Nuclear Power Plant Risk Management

Probabilistic safety analyses developments resurgence in Bulgaria—Kozloduy nuclear power plant example

At present, probabilistic safety analysis (PSA) is widely used in the many fields of nuclear power plant (NPP), such as design, manufacture and etc. And PSA has gradually developed into the important tool of safety evaluation and decision making. With the development and improvement of PSA technology, the risk-informed safety management has also made great progress. The overhaul of NPP is an important work to keep safe and stable operation of NPP. During the overhaul, fuel replacement and maintenance of a large number of equipment will be carried out. Under different plant operation status (POS) in the overhaul process, several important equipment will exit operation at the same time, which will lead to the increase of core loss frequency of different degrees. At present, only few NPPs within the scope of China National Nuclear Power Co., Ltd (CNNP) has adopt qualitative PSA method to carry out the risk assessment of the overhaul, but there is no precedent to use quantitative PSA method to carry out the assessment. The paper will introduce how to use the risk-informed PSA method to assess the risk of the overhaul in a quantitative way. The higher risk stages in the overhaul will be identified through the assessment to optimize the work schedule of the overhaul to ensure that the work schedule will not lead to excessive increase in core damage frequency (CDF). And taking the fifth overhaul of unit one of Fangjiashan NPP (FJS NPP) (105 overhaul) as an example, to assess the work schedule of 105 overhaul and give reasonable Suggestions.

According to existing research results, fire risk makes a significant contribution to the total risk of a nuclear power plant (NPP). So fire probabilistic safety analysis (PSA) for NPPs is becoming more and more important in recent years. How to perform human reliability analysis (HRA) which is an essential part of PSA is therefore being paid more and more attention in fire PSA. This paper describes the characteristics and special considerations of HRA in fire PSA, and demonstrates in fire PSA how to use SPAR-H method which is so-called an advanced second-generation HRA method and is being widely used in PSA for Chinese NPPs. The study results can be a reference for other HRA analysts to use SPAR-H method in fire PSA models or other PSA models in Chinese NPPs or the world-wide nuclear industry.

Fusion or thermonuclear power is a promising, literally endless source of energy. Development of fusion power is still in investigation and experimentation phases and a number of fusion facilities are under construction in Europe. Since fusion energy is innovative and fusion facilities contain unique and expensive equipment an issue of their reliability is very important from their efficiency perspective. A Reliability, Availability, Maintainability, Inspectability (RAMI) Analysis is being performed or is going to be performed in the nearest future for such fusion facili­ties as ITER and DEMO in order to ensure reliable and efficient operation for experi­ments (e. g. in ITER) or for energy production purposes (e. g. in DEMO). On the other hand, rich experience of the Reliability and Probabilistic Safety Analysis (PSA) exists in nuclear industry for fission power plants and other nuclear installations. In this article, the Wendelstein 7-X (W7-X) facility is mainly considered. It is a stel­larator type fusion facility under construction in the Max-Planck-Institut fur Plasmaphysik, Greifswald, Germany (IPP). In the frame of cooperation between the IPP and the Lithuanian Energy Institute (LEI) under the European Fusion Development Agreement a pilot project of a reliability analysis of the W7-X systems was performed with a purpose to adopt NPP PSA experience for fusion facility systems. During the project a reliability analysis of a divertor target cooling circuit, which is an important system for permanent and reliable operation of in-vessel components of the W7-X, was performed.

Fire PSA has to be performed as part of the Probabilistic Safety Analysis (PSA) for full power and low power and shutdown states within the mandatory comprehensive Safety Reviews for German nuclear power plants (NPP) at time intervals of ten years. The German guidance document on PSA methods requires a Fire PSA of several steps starting with alternatively a qualitative and quantitative or an advanced combined screening approach providing estimated values for damage frequencies. The existing combined approach has been enhanced by automating the screening process including a systematic calculation of fire propagation probabilities and standardized simplified fire simulations to roughly predict critical parameters. Further enhancements focus on fire induced cable failures and circuit faults and improvements in the uncertainty and sensitivity analysis. Thus the uncertainties in the results of fire PSA can be reduced as far as feasible and the predicted CDF values are more reliable. The results of a Fire PSA for a German NPP with boiling water reactor designed to earlier standards (BWR-69 type) with an overall fire induced core damage frequency of 1.9 E−06/a are outlined. This is in good agreement with results publicly available from Fire PSA for NPP in other countries. However, the CDF values are higher than those for some other German plants resulting from some pessimistic assumptions made.

In the last four decades, as the nuclear industry grew and got mature, the importance of adequate risk evaluating tools became decisive. Therefore, the Probabilistic Risk Assessment (also known as Probabilistic Safety Analysis) became a cornerstone of the decisions in such high energy and high-risk industry. PSA has an internationally recognised standard, and it is supported by a group of highly trained experts, (no more than a few hundred worldwide).This work can be used as a guide for the improving the required individual and teamwork skills needed in a Probabilistic Safety Analysis - PSA Team. The necessity of such a moment in a PSA Training was imagined by Dan Serbanescu, doctor in science, nuclear energy expert, risk and safety analyst, in May 2017. After few discussions and according to recognised international standards (Probabilistic Risk Assessment procedures guide, 1983), a first time delivery was possible in the PSA Training delivered in Centrala Nuclearoelectrica Cernavoda / Nuclear Power Plant Cernavoda (2017).This article presents a systematic approach for team improvement skills, consisting of the observation, presentation of the skills required, the skills practised in the proposed exercises, the techniques used during this module (coaching included), and results. The Purpose of the newly proposed combination of training and coaching methods with the specific traditional one oriented mainly to the technical and procedural skills is to raise participants’ awareness about how soft-skills can be used in the PSA Teamwork. As Nuclear Power Plant can be easily compared with a complex organisation, soft skills are vital to be developed within the teams. PSA is becoming more required not only in nuclear but also in the aerospace industry (it was adopted by NASA - National Aeronautics and Space Administration for all future space program and by some hazardous chemical industries, as also stated in international documents (of the European Commission for instance).

At present, probabilistic safety analysis (PSA) has been widely used in many fields such as the design and manufacture of nuclear power plants, and PSA has gradually developed into an important tool for safety evaluation and decision-making. With the increasing peak-to-valley difference between power grid loads, the peak-shaving situation of the power system is becoming more and more severe. Due to the increase in the proportion of nuclear power in the power grid, the power system has an increasing demand for nuclear power units to participate in power grid peak-shaving. The nuclear power plant technical specifications and the nuclear power plant final safety analysis report in the nuclear safety regulations have clear requirements for the operating conditions of the nuclear power plant. As a power generation device with special safety requirements, the safest and most economical operation mode of nuclear power plants is to operate at rated load. The Chinese government and nuclear safety regulatory authorities attach great importance to nuclear power safety, and nuclear safety must be foolproof. From the perspective of probabilistic safety analysis, this paper analyzes the impact of Fangjiashan Nuclear Power Plant’s (FSJ NPP) participation in peak-shaving by using the load-following operation mode, and gives reasonable suggestions.

Maintenance risk management in Daya Bay nuclear power plant: PSA model, tools and applications

Probabilistic safety analysis (PSA) is one of the tools that has been used for a long time to quantitatively assess the safety status of nuclear power plants (NPPs) around the world. However, despite all the positive aspects of the PSA of nuclear power plants, it, for example, does not take into account such a component as physical protection, which carries with it a corresponding negative effect, which is associated with the vulnerability of nuclear facilities to sabotage. It is not possible to directly involve the PSA of NPPs to evaluate the effectiveness of physical protection of nuclear facilities, since the existing methodology and procedure for applying this tool is not suitable for physical protection, which requires the development of a new methodology for the probabilistic analysis of physical protection. At the same time, it should be noted that in some cases it is possible to apply the methods of the traditional NPP PSA to perform some elements of the probabilistic analysis of physical protection without changes (for example, the analysis of equipment reliability), in other cases it is necessary to develop new methods for some of the elements of the probabilistic analysis of physical protection by analogy with the traditional NPP PSA (that is, to adapt, for example, the analysis of success criteria), and finally, in other cases, it is necessary to develop own new methods (for example, the analysis of the offender's actions) for the elements of the probabilistic analysis of physical protection, which are absent in the traditional NPP PSA. In general, the process of applying PSA of nuclear power plants to evaluate the effectiveness of physical protection systems of nuclear facilities requires not only the development of a new methodology, but also requires a rather large expenditure of time with the involvement of significant human and material resources. Therefore, for the first iterative step of using the traditional PSA of nuclear power plants to evaluate the effectiveness of physical protection of nuclear facilities, as an example, the possibility of adapting the methodology of the analysis of the success criteria of the traditional PSA of nuclear power plants for the probabilistic analysis of physical protection of nuclear facilities is considered.

Tianwan Nuclear Power Plant (TNPP) Units 3&4 (the second phase) are a landmark project for Chinese and Russian governments to jointly promote China-Russia nuclear energy cooperation.All the reactor units adopt the Russian VVER-1000 improved nuclear power units which are now operated by project owner, Jiangsu Nuclear Power Corporation (JNPC). The depth and quality of PSA development, at the stage of Units 3&4 commercial operation,should be sufficient to obtain the risk-informed insights, enough to support the Risk-Informed application of PSA, and also reflect the actual design and its operation experience of TNPP. This paper presents the brief introduction of PSA of Unit 3&4 of TNPP (power operation mode). The important parts of this presentation are as follows:probabilistic safety indices, study scope, technical elements and the analysis of Core Damage Frequency (CDF) quantitative results. Through quantitative calculation, the point estimation value of the total CDF of TNPP under the power condition is obtained. And the frequency of occurrence of each event sequence and the CDF contribution of each initiating event (IE) etc., are analyzed. Based on the developed PSA models, four PSA application tools have been developed and applied in the risk monitoring activities of nuclear power plant. This PSA model is used to take the lead in carrying out the Online maintenance in China.

Level-1, -2 and -3 PSA for AHWR

The German PSA Guideline and its technical documents on PSA methods and data require probabilistic safety analyses (PSA) to be carried out in the frame of safety reviews for nuclear power plants. Since 2005 this also includes a seismic PSA (SPSA) for sites with design earthquake intensities exceeding the value VII (MSK-64/EMS-98). It is shown how the plant model of PSA Level 1 for internal events can be extended on the level of fault tree basic events to get a quantifiable seismic plant model. A two-step screening procedure can be applied to derive the seismic equipment list (SEL) and a list of all possibly seismic-induced dependent equipment failures. The screening procedure is supported by a database. The database keeps at hand all the data and information to extend the plant model of PSA Level 1 in a proper manner.

The main steam system is an important system connecting the primary and secondary loop of the nuclear power plant. It transports the steam generated by the steam generators to the steam turbine to drive the turbine to turn and do work. The main steam isolation valve (MSIV) is important safety devices in nuclear power plant. Its safety function is to automatically shut down under the accident condition of large break in the main steam pipe or main water supply pipe in order to limit the steam leakage and mitigate the consequences of the accident. The main steam partial shutdown test is a test to periodically verify the operability of the MSIV during the power operation of the nuclear power plant. The test period is usually once a month. However, during the test, there may be shutdown events caused by the mistakenly close of the MSIV, so the test is a high-risk periodic test. At the same time, due to the limitations of field condition and space, there are so many industrial risks and the risk of accidental collision leading to shutdown during the test. At present, the probabilistic safety analysis (PSA) is widely used in operation guidance, maintenance rules and many other fields of nuclear power plants. PSA has gradually developed into an important tool for safety evaluation and decision making. With the development and improvement of PSA technology, risk-informed safety management has been widely used. The this paper, the use of risk-informed PSA method to demonstrate the extension of periodic test period will be introduced. And taking FJS Nuclear Power Plant as an example, the extension of the periodic test period of the partial closing test of the MSIV is demonstrated.

The Fukushima nuclear disaster has raised the importance on the reliability and risk research of the spent fuel pool (SFP), including the risk of internal events, fire, external hazards and so on. From a safety point of view, the low decay heat of the spent fuel assemblies and large water inventory in the SFP has made the accident progress goes very slow, but a large number of fuel assemblies are stored inside the spent fuel pool and without containment above the SFP building, it still has an unignored risk to the safety of the nuclear power plant. In this paper, a standardized approach for performing a holistic and comprehensive evaluation approach of the SFP risk based on the probabilistic safety analysis (PSA) method has been developed, including the Level 1 SFP PSA and Level 2 SFP PSA and external hazard PSA. The research scope of SFP PSA covers internal events, internal flooding, internal fires, external hazards and new risk source-fuel route risk is also included. The research will provide the risk insight of Spent Fuel Pool operation, and can help to make recommendation for the prevention and mitigation of SFP accidents which will be applicable for the SFP configuration risk management.

In nuclear power plant design and, after, when they are under work, in front of any change in the design or periodical safety review, it is necessary to perform safety studies in order to guarantee the safety operation along their useful life. These safety studies, traditionally has been divided between deterministic safety analysis and probabilistic safety analysis, although the last years trending is to integrate the characteristics of both classes of analysis in order to build more complete safety studies. Among the deterministic safety analysis, when the Best Estimate (BE) codes are employed and, in addition, the uncertainty effect are taken into account, we are inside of the methodology called Best Estimate Plus Uncertainty (BEPU).\nThis Thesis provides new tools and procedures in order to perform the deterministic safety analysis of nuclear power plants by means of Best Estimate methodology through of several applications employed. Statistical tools are provided for performing BEPU analysis. Particularly, a procedure is presented for built BEPU studies that can be applied in almost all the transients and facility in a methodical way. This procedure is comprehensive and include from the development of the transient scenario by means of BE thermalhydraulic code and the input parameters selection to the uncertainty propagation over the safety criteria and the verification of their compliance using different uncertainty analysis methods, both parametric and non parametric methods. \nWith the purpose of demonstrating the procedure versatility, this it is applied to the studio of transients and facilities with different phenomenology. Specifically, it have been applied to: PWR nuclear power plant for a Large Break Loss of Coolant Accident (LBLOCA), experimental facility for a Small Break Loss of Coolant Accident (SBLOCA) and in a spent fuel storage of a PWR nuclear power plant for a loss of coolant accident.\nLast, this Thesis contribute with a methodology accordingly for incorporating assumptions from probabilistic safety analysis about system configuration availability inside the deterministic safety analysis. Therefore, an approach enclosed into the known as Extended BEPU (EBEPU) methodologies is constructed. In order to demonstrate the viability and applicability of this methodology, an application case is provided, which consists in Loss of Feed Water system (LOFW) in a PWR nuclear power plant.\nThe work carried out in this PhD thesis are enclosed into the grant of "Formación de Personal Investigador (FPI)-Subprograma1 de la convocatoria de 2012" supported by the "Programa de Ayudas de Investigación y Desarrollo (PAID)" of the "Universitat Politècnica de València".