National Nuclear Safety Administration Research Articles

Status and Assessment Method of Nuclear Safety Analysis Software in China

In recent years, China’s nuclear power industry has enjoyed a good momentum of development, and related companies have also developed many nuclear analysis software applications. However, as the National Nuclear Safety Administration (NNSA, Chinese nuclear regulatory institution) did not approve any software before 2018, all these software applications were not evaluated formally, so they have not yet been used in reactor safety analysis. In order to solve this problem, in 2018, the National Nuclear Safety Administration started to carry out an engineering applicability evaluation for software developed by Chinese companies. After several years of review, as the first approved Chinese domestic software, core physics analysis software PCM developed by the China General Nuclear Power Group officially passed the software safety evaluation of the China Nuclear Safety Administration. This study will present the basic situation of the development of China’s nuclear power engineering software and introduce the framework, methods, procedures, requirements, and other aspects of China’s software safety evaluation work. The evaluation process and evaluation key issues of PCM software will also be illustrated.

Discussion on Environmental Qualification Management System of Electrical Equipment in Nuclear Power Plants

In order to ensure the safety of the reactor system, the nuclear power plant (NPP) is designed in accordance with the principle of defense in depth. The single failure criterion, whose direct goal is to strengthen the safety-related system function through redundant settings, is an important cornerstone of the principle of defense in depth. It is the primary role of environmental qualification(EQ) to ensure that Class 1E equipment can perform its safety function(s) with no failure mechanism that could lead to common cause failures under postulated service conditions.This paper firstly describes the equipment qualification management system of NPPs in France as well as America. Secondly, equipment qualification regulatory requirements and management practices of NPPs in China were shared. Finally, combined with the regulatory requirements of National Nuclear Safety Administration of China and good experience from NPPs at home and abroad, suggestions are given for electrical equipment EQ management system in other NPPs in China.

Design study of hydrogen monitoring equipment for severe accident in containment of nuclear power plant

After the Fukushima accident, the risk of hydrogen shall be considered by most countries in nuclear power plants. China national nuclear safety Administration has required that nuclear power plants shall install the hydrogen monitoring measurement equipment for severe accident in containment. There were some equipment used for hydrogen monitoring at severe accident, but some equipment cannot solve the leakage problem, some equipment cannot measure the hypoxic environment, and some have slower response time or long startup time. In this study, we developed a based on Pd–Ag​ alloy hydrogen monitoring equipment, Pd–Ag alloy is selected as hydrogen sensitive material, alumina ceramic is used as substrate, and Pd–Ag alloy film hydrogen core measurement component is prepared by magnetron sputtering technology. Heating​ treatment in sensor is adopted to maintain the environment in which the sensor is located at a constant 198°C environment, and protective and filtering measures are taken to ensure that hydrogen detection is not affected by aerosol and water vapor. Then, study the characterization of hydrogen monitoring equipment. Testing results indicated that the hydrogen measuring component (Pd–Ag) working at a constant temperature does not affect the measurement results and this equipment has the constant line relationship between hydrogen concentration and resistant. The response time of hydrogen decrease is longer than that of hydrogen increase, but it is within 50 s. It also has the high accuracy keeping within 1 vol% at normal anoxic environment and harsh environment. If it is applied to severe accidents in NPPs, it is necessary to study the performance under irradiation conditions in the future.

Research on level 1 PSA and risk-informed design for ACP100

Advanced China Power 100 (ACP100) is a kind of multi-purpose small modular reactor (SMR) based on the present pressurized water reactor (PWR) technology. It adopts the integrated reactor technology routes, satisfies the current nuclear safety codes and guides issued by National Nuclear Safety Administration (NNSA), and also refers to the safety standards of International Atomic Energy Agency (IAEA). In the design phase, Probabilistic Safety Assessment (PSA) technology plays an important role to demonstrate the total safety level, identify the design vulnerabilities, and provide suggestions for the improvement of ACP100. After the brief summary of the level 1 PSA and its risk-informed design method, the level 1 PSA analysis was carried out for ACP100. The identification of initiating events (IEs), the event tree analysis, fault tree analysis and accident sequence quantification were performed. The main PSA results are shown in the paper, such as core damage frequency (CDF), dominating accident sequences, significant common cause failures (CCF). Finally, different design options and their impacts on risk which can be reflected by risk return were evaluated to avoid using the risk-increased design based on the quantitative risk calculation for the sake of increasing the ACP100’s safety. Consequently, the results show that the point estimated value of the total CDF of level 1 IEs PSA for operations at power of ACP100 is 1.42E-07/ry. After the design improvement proposed in this paper, the CDF (point estimate) of ACP100 can be reduced to 6.78E−08/ry. The risk were obviously reduced and the safety of the ACP100 is significant improved. This research provides a technical reference for future new nuclear reactor design.

The governance of nuclear power in China

Abstract China has the third largest fleet of nuclear power plants in the world, totalling more than 45 GWe at the end of 2019. With the current high rate of growth, its capacity will soon overtake that of France. The country’s nuclear power industry has suffered no serious accidents to date. Nevertheless, the poor safety record of some other heavy industries in China, combined with the rapid growth of civil nuclear power capacity, has raised concerns over the industry’s ability to prevent a serious accident. The organization, development and governance of China’s nuclear power industry reflects the high strategic importance that the government has placed on the industry over several decades. At the same time, it has taken steps to address domestic and international concerns over its ability to effectively govern nuclear safety and security. The country has become party to most major treaties and conventions relating to nuclear matters and has frequent interaction with the International Atomic Energy Agency. Despite many significant steps taken by the government, a number of questions remain concerning: the capacity and independence of the National Nuclear Safety Administration; the relatively incoherent nature of the body of laws, regulations and rules that govern nuclear safety and security; the absence of a clear legal basis for managing civil nuclear liability, especially in the context of an accident with transboundary consequences; and the quality of public participation, especially in the case of planned nuclear power plants.

An integrated framework for analysing operational events in China nuclear power plants

Operational events of nuclear power plants (NPPs) are safety-relevant incidents, which may impair the plant functionality and lead to undesired consequences. Thus, their analysis is important for the safety assessment of NPPs. Although several studies have been conducted on NPPs operational events, there is a lack of an integrated framework that effectively combines qualitative and quantitative analyses of operational events. In this study, a framework for the identification and quantification of root causes of operational events is proposed. The framework includes the root causes identification procedures and an online quantification tool based on Bayesian Network modeling. Thirty-seven unplanned reactor trip (URT) events occurred in Chinese NPPs have been used for the case study and the outcome indicates that this framework can be quite useful: on the one hand, it can support inspection and control of the measures of prevention of operational events and reduction of their occurrences from the perspective of regulatory authorities; on the other hand, it can be embedded within the national nuclear safety administration (NNSA) experience feedback (EF) system as an online support tool for living safety assessment.

Preliminary environmental radiation considerations for CFETR

Since Chinese fusion engineering test reactor (CFETR) stepped into the engineering design phase, it was significant to discuss the environmental radiation issues considering both Chinese radiation safety regulations and lessons learned from international thermonuclear experimental reactor (ITER) being constructed in France. In this study, firstly, the legal and regulatory framework of nuclear safety in China and ITER licensing process lessons were reviewed. Secondly, public dose limit was proposed for CFETR. Thirdly, safety limit and radioactive source terms were discussed and proposed for CFETR. Fourthly, public radiation dose was estimated, tritium emission limit was proposed under normal operation from the perspectives of tritium emission limit for nuclear power plants (NPPs) in China and reverse value based on public dose limit. Finally, potential consequences of environmental radiation were discussed under accidental conditions. Nuclear emergency countermeasures including no evacuation criteria needed or not were proposed. The preliminary environmental radiation considerations were proposed for CFETR: 1) It had better invite the experts from Chinese national nuclear safety administration (NNSA) to participate in the engineering design, which can make the related persons understand the fusion safety characteristics as early as possible, so as to further speed the licensing process; 2) 0.1 mSv/a was proposed as the public dose limit of CFETR under normal operation; 3) The same ITER safety limit was proposed to be set for CFETR from the consideration of both comparable size of vacuum vessel (VV) and hydrogen/dust explosion; 4) A tritium emission limit of 5–33 g/a was proposed for CFETR; 5) The “no off-site emergency response” was suggested as one of the safety goals during the design of CFETR, however, “off-site emergency measures” were still proposed to be deployed during the operation of CFETR. It was expected the environmental radiation considerations could provide a reference to speed the construction of CFETR on the premise of environment friendly and public safety.

Research and verification on CFD natural convection model of QM400

The spent fuel dry storage facility, QM-400, for the Third Qinshan Nuclear Power Co. Ltd. (TQNPC) is the first commercial dry storage facility in operation in China. The heat transfer in QM-400 consists mainly of natural convection, conduction, conjugate heat transfer, and radiation. In our research, we calculated the decay heat of each fuel basket at a typical surrounding temperature. Mesh sensitivity analysis was done, and a set of commercial computational fluid dynamics (CFD) models were established for QM-400 using ANSYS Fluent software. Several thermocouples were set up at typical locations in the QM-400 module in the nuclear power plant to validate the proposed CFD model. Good agreement was achieved between the simulated temperatures and the measured values. The trends varying with the surrounding temperature agreed well too, which showed that the CFD model is suitable for simulating the complex heat transfer phenomena in QM-400. Using the proposed model, we also simulated the temperature field after eliminating insulation panels. The calculation results show that the local concrete temperature was greater than 66 °C and less than 93 °C. The thermal stress and cross-sectional internal force introduced by the high local temperature met the requirements, indicating that it would be feasible to eliminate the insulation panels. The program of eliminating dry cask insulation panels has been approved by China’s National Nuclear Safety Administration.

Solutions for Incubation and Building of Nuclear Safety Based on Regulatory Perspective

It is indicated about the requirements of China’s current nuclear and radiation safety regulations on nuclear safety culture and the status of regulation of nuclear safety culture by the National Nuclear Safety Administration (hereafter as the “NNSA”) in this paper. With this information summarized, the characteristics of regulation of nuclear safety culture are analysed, and corresponding solutions are proposed.

Research on Nuclear Safety Culture Assessment by China’s Regulatory Body

Regulatory body on nuclear safety in China always pays much attention to nuclear safety culture, and keeps implementing values and principles of nuclear safety culture into every section of management. In 2014, with the issuance of “Nuclear Safety Culture Policy Statement”, the Ministry of Environmental Protection (MEP) (National Nuclear Safety Administration (NNSA)) set the research on evaluating nuclear safety culture as one of the key projects. Currently, Nuclear and Radiation Center of MEP has proposed the “eight traits” based on the “Nuclear Safety Culture Policy Statement”, and completed development of the evaluation model and analysis of nuclear safety culture indicators.

A data mining framework within the Chinese NPPs operating experience feedback system for identifying intrinsic correlations among human factors

With the continuous increase in the number of operating nuclear power plants (NPPs) in China, the amount of operating experience feedback (OEF) increases significantly. On the other hand, the safe operation of NPPs has become an urgent problem that the National Nuclear Safety Administration (NNSA) must solve. To this end, NNSA established a nationalwide OEF system to improve the safety level of NPPs and strengthen the exchange of operating experience. Analyzing the human factors events (HFEs) is an important part of OEF and it is significant to improve human performance and prevent human error. Data mining has been recognized as an effective way to analyze data. With the continuous increase in operating event reports, data mining related to nuclear safety becomes a new domain of study. In this paper, we propose a data mining framework in support of the OEF system. The framework combines three statistical approaches (i.e., correlation analysis, cluster analysis and association rule mining) for identifying intrinsic correlations among human factors: correlation analysis measures the strength of linear relationship between human factors; cluster analysis classifies human factors into relevant groups; association rule mining identifies associations and causalities among human factors. For illustration, we apply the proposed framework to 162 human factors events (screened out from 313 events collected from the OEF system), and the results reflect the feasibility and effectiveness of the framework in identifying the intrinsic correlations among human factors. Besides, further suggestions for improving human performance and preventing human errors in NPPs are also discussed.

Analysis of loss of residual heat removal system for CPR1000 under cold shutdown operation

After Fukushima accident, the safety assessments within the full range of operation conditions for the current fleet of nuclear power plants in China are requested to be performed by the national nuclear safety administration. Accordingly, twelve typical cases were simulated to study the thermal-hydraulic behavior of loss of residual heat removal system for CPR1000 under cold shutdown operation with RELAP5/MOD3.3 code. The transient behaviors of system pressure, coolant temperature, water inventory, coolant boiling and core uncovering time, and other important parameters under 12 different plant configurations are predicted. The results indicate that the system pressure of cases with intact system increases continuously to the set value of 3.8MPa for opening the relief valve in the RHR system without any mitigation action. The starting time of coolant boiling highly depends on the initial water inventory in the RPV as well as its behavior during the initial stage after losing the RHR system, while the starting time of core uncovering highly depends on the plant configurations. The results of the study can provide references for the plant safety improvement and guideline for an operator to strategize a mitigation action.

Component Ageing and Life Management Database Design for Multi-Nuclear-Power-Plant

According to the requirement of the nuclear power plant ageing management, combined with the nuclear safety guide HAD103/12 “nuclear power plant ageing management" released by NNSA (national nuclear safety administration), this paper introduces the applicable to multiple power plant equipment ageing and life management of the database design requirements, the development process, general structure, composition and function, application scope, maintenance and safety related content.

In-vessel retention coolability evaluation for Chinese improved 1000 MWe PWR

After the Fukushima Daiichi accident, the Chinese National Nuclear Safety Administration requires enhancements of safety and mitigation capability under severe accidents for nuclear power plants (NPPs). In-vessel retention (IVR) strategy is proposed to mitigate the consequences of severe accidents for pressurized water reactors (PWR) to satisfy the requirements. Therefore, a proposal with passive cavity injection system (PCIS) is designed for Chinese improved 1000MWe PWR for implementation of IVR strategy. Besides, to condense the steam generated in the cavity head, a passive heat exchange system (PHES) is also designed for passive containment cooling. To evaluate the coolability of the PCIS, a computer code is developed to assess the effectiveness of IVR strategy based on the integrated risk oriented accident analysis methodology (ROAAM) approach. In order to obtain input parameters for assessment, seven severe accident sequences are screened based on Level 1 PRA results and analyzed with integral accident analysis computer code. With the characteristics of the molten pool and key parameters gained from accident analysis, the IVR assessment program is developed with MATLAB code considering parameters of decay power, zirconium oxide fraction, the mass of stainless steel (SS). The influence of thermophysical parameters, critical heat flux (CHF) correlation and mass of metal layer on IVR effectiveness are discussed as sensitivity study for the assessment. Thermophysical parameters have the major effect. Combining the sensitivity study, assessment results show that the IVR strategy for Chinese improved 1000MWe PWR has an acceptable high success probability. However, for engineering practice, detail system design and analysis should be evaluated in the near future.

Safety evaluation of the modified auxiliary feedwater system for the Chinese improved PWR

Chinese National Nuclear Safety Administration (NNSA) has issued the general technique requirements of the modification actions for the Nuclear Power Plants (NPPs). The enlargement of the tank volume of the auxiliary feedwater system (ASG) is one of the modification actions to enhance the safety margin for the Chinese improved PWR NPPs. Therefore, it is necessary to evaluate the tank volume of ASG can supply enough cooling water to the secondary side of the steam generators under the transient accidents, to maintain the reactor under safety situation. Typical accident sequences, such as the category II accidents of the loss of feedwater (LOFW) with no loss of offsite power (no-LOOP) and the category IV accidents of the feedwater line break (FWLB) with no-LOOP and the FWLB combined with LOOP, are selected to analyze the minimum flow rate and volume for ASG, based on the RCC-P requirements, based on the model of the typical improved generation II 1000MWe nuclear power plant, built by using the mechanical safety analytical code. The auxiliary feedwater flow rates of 5m3/h, 10m3/h, 15m3/h and 25m3/h to each steam generator (SG) for the LOFW with no-LOOP are investigated by analyzing the transient progress, and the results show that the minimum flow rate is 25m3/h. FWLB transients with different flow rates of 30m3/h, 35m3/h and 45m3/h are also analyzed to show that 45m3/h is the minimum flow rate considering a safety margin. The minimum volume of the auxiliary feedwater tank is 713m3 for the LOFW, 723m3 for the FWLB with no-LOOP and 799m3 for the FWLB combined with LOOP. Therefore, considering the safety margin, the minimum auxiliary feedwater flow rate necessary to prevent bulk boiling in the reactor coolant system is 45m3/h for each unbroken SG, and a 900-m3 volume of auxiliary feedwater tank can satisfy the RCC-P requirements for the improved PWR NPPs, to give more time to take mitigation measurements.

Discussion on IAEA and China Safety Regulation for NPP Coastal Defense Infrastructures against Typhoon/Hurricane Attacks

The World Meteorological Organization estimates that about 90 percent of all natural disasters is extreme meteorological hazards like typhoon/hurricane and tropical cyclone triggered disasters. With the increasing tendency of natural hazards, the typhoon induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of the planned, designed and constructed NPP in China the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane/ Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations recommended for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compared with IAEA 2003-2011 recommended safety regulations for some NPP along China coast to make safety assessment based on the “As Low As Reasonable Practice” (ALARP) principle.