Uncertainty and sensitivity analysis of MCPs’ trip events at Ignalina NPP

Eight main circulation pumps (MCPs) are employed for the cooling water forced circulation through the RBMK-1500 reactor at the Ignalina nuclear power plant (NPP). There have been a few events when one or more MCPs were inadvertently tripped.This paper presents investigation of a one MCP trip event and all MCPs’ trip events at Ignalina NPP. Thermal-hydraulic analysis was conducted using the best estimate system code RELAP5/MOD3.3. Uncertainty and sensitivity analysis of flow energy loss in different parts of the main circulation circuit (MCC), initial conditions and code-selected models was performed. Such analysis allows to estimate the influence of separate parameters on the calculation results and find those modelling parameters that have the largest impact on the investigated events. Uncertainty analysis indicates that natural circulation provides adequate cooling in the case of all MCPs tripped, and that the reactor is reliably cooled by forced circulation in the case of a single tripped MCP. On the basis of this analysis, recommendations for the further improvement of model are developed.     

A quantitative methodology for reactor vessel pressurized thermal shock decision making

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Consideration of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS.A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation.Once the specific event sequences of concern are identified, detailed thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. This paper addresses key aspects of the thermal-hydraulic and fracture mechanics analyses of the reactor vessel. The effects of incomplete mixing of safety injection flow in the primary cold leg and vessel downcomer and the application of warm prestressing are emphasized. The results of these analyses are being used to define further modifications in vessel and plant system design and to operating procedures.Previous design considerations that have evolved as a result of reactor vessel integrity evaluations are mentioned. These include the development of realistic design analysis tools and selection of plant system modifications. Modifications that are being developed or are under consideration are also mentioned. These include vessel fluence reductions, additional modifications to operating procedures, increased use of probabilistic event sequence and fracture mechanics analysis methods, enhanced material fracture toughness, and reductions in the severity or frequency of occurrence of dominant reactor vessel PTS transients.     

The evolution of the structure and application of U.S. NRC regulations and standards

NRC regulations and standards and their implementation have evolved from early adaptations of conventional engineering practices to a mature, cohesive set of regulations that govern NRC regulation of nuclear power plant safety in the United States.From a simple set of rules and design criteria and from the standards of the professional engineering societies, a hierarchy of practices, standards, guides, rules and goals has developed. Resting on a foundation of industrial practices, this hierarchy rises through levels of national standards, regulatory guides and standard review plans, policy statements and NRC regulations.The licensing process is evolving today toward one that permits both site approval and standard design certification before the plant is constructed. At the present time, NRC is reviewing five standard designs for certification for a period of 15 years. NRC focuses its regulation of operating nuclear plants on inspections conducted from five regional offices. Resident inspectors, specialist inspectors, and multi-disciplinary inspection teams examine specific plant situations. The results of all these inspections are used to develop a complete understanding of a plant's physical condition, its operation, maintenance and management.To improve safe operation of nuclear plants in the U.S., a most important program, the Systematic Assessment of Licensee Performance, measures operational performance, using a broad spectrum of functional areas.     

Performance evaluation of a new signal processing system design to improve CANDU SDS1 trip response during large break LOCA events

Performance of a recently developed signal processing system for CANDU (Canada Deuterium Uraniu) reactor shutdown system 1 (SDS1) is evaluated in this paper. The evaluation is carried out in MATLAB/Simulink software environment as well as with an existing power measurement and signal processing system. The new signal processing algorithm is obtained based on the synthesis of several first order low pass filters with different delayed time constants. Throughout this paper, a special attention has been paid to compare the new signal processing system with the existing one. The dynamic behavior of the new signal processing system in the practical large loss of coolant accidents (LLOCA) events has also been examined. Simulation results show that during the LLOCA event, the reactor trip time, as well as the peak power, is decreased remarkably. Through the simulation studies, it has convincingly demonstrated that the new signal processing system has significant advantages over the existing system in terms of the improved trip response and accommodation of the spurious trip immunity. This advantage will significantly enhance the safety margin, or will bring economical benefits to nuclear power plants.     

Westinghouse AP1000 advanced passive plant

The Westinghouse AP1000 Program is aimed at making available a nuclear power plant that is economical in the US deregulated electrical power industry in the near-term. The AP1000 is a two-loop 1000 MWe pressurizer water reactor (PWR). It is an uprated version of the AP600. Passive safety systems are used to provide significant and measurable improvements in plant simplification, safety, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 35 years of operating PWR experience. The AP1000 received Final Design Approval from the United States Nuclear Regulatory Commission in September 2004; the AP1000 has also received Design Certification by the USNRC in December 2005. The AP1000 and its predecessor AP600 are the only nuclear reactor designs using passive safety technology licensed anywhere in the world. The safety performance of AP1000 has been verified by extensive testing, safety analysis and probabilistic safety assessment. AP1000 safety margins are large and the potential for accident scenarios that could jeopardize public safety is extremely low.Simplicity is a key technical concept behind the AP1000. It makes the AP1000 easier and less expensive to build, operate, and maintain. Simplification also provides a hedge against regulatory driven operations and maintenance costs by eliminating equipment subject to regulation. The AP1000's greatly simplified design complies with NRC regulatory and safety requirements and the EPRI advanced light water reactor (ALWR) utility requirements document.Plans are being developed for implementation of the AP1000 plant. Key factors in this planning are the economics of AP1000 in the de-regulated US electricity market, and the associated business model for licensing, constructing and operating these new plants.     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.     

Using the Hurst’s exponent as a monitor and predictor of BWR reactor instabilities

Since the decade of the 1950s, when the development of boiling water reactor technology began, unstable situations have existed, which involve a high amplitude self-oscillatory process in the reactor’s thermal power. As the development progressed and the reactors increased power density, the possibility of instability under certain circumstances increased. Thus, in 1985, Caorso nuclear plant (Italy) reported the first event of this type, and in 1988 (NRC, 1988), such an event was reported at La Salle as well. Since then, multiple instability events have been reported. The danger of these unstable power situations resides in the possibility of exceeding a thermal limit, as expressed in Appendix A of 10FR50. Thus, the need arises to monitor and correct these situations in the industry.     

TINTE analysis of PBMR reactivity insertion transients

Several important PBMR reactivity insertion transients have been simulated by means of the dynamic reactor code TINTE (Time-dependent Neutronics and Temperatures). These transients include total control rod removal during normal operating conditions, cold zero power (CZP) conditions and hot zero power (HZP) conditions, as well as Reserve Shut-down System (RSS) removal during the cold zero power conditions. According to the TINTE results, the worst control rod removal scenario is the total control rod removal during hot zero power conditions in a postulated event with the power conversion unit (PCU) starting up and with continual operation without a trip. The amount of reactivity insertion by the RSS removal is much greater than the reactivity insertion by control rod removal. The event was analyzed but the RSS during the cold zero power conditions is a Beyond Design Basis Accident (BDBA). This paper presents the methodology followed to model these transient events with the TINTE code, as well as total power, maximum and average fuel temperatures and case comparison results. It is also shown that the maximum fuel temperatures for all of the Design Basis Accidents (DBA) are within acceptable safety limits.     

Activities for 4S USNRC licensing

4S (Super-Safe, Small and Simple) is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas, high-temperature steam supply more than 400 °C, seawater desalination, and hydrogen production. The system design of power output of 10 MWe (30 MWt) has been completed. The main feature is that it does not have to be refueled for a long period (i.e. 30 years for 10 MWe version), and enable the reactor closure sealed during plant operation. Furthermore, the small size of the reactor makes the reactor building suitable for below grade installing. These two features can provide resolutions for the issues relevant to safety, security, and safeguard, which become much more serious matter internationally these days.4S is a pool-type reactor which contains the whole primary cooling system in a vessel. For the purpose of reducing the maintenance requirements with the reactor, (1) reflectors to compensate for fuel burn-up instead of control rods, (2) electromagnetic pump (EMP) which has no rotating parts, and (3) residual heat removal system by natural circulation and natural air draft are adopted. Therefore, exchange of the reactor components is not required during plant operation, in addition to no needs for refueling.Toshiba has initiated the U.S. Nuclear Regulatory Commission (NRC) pre-application review of 10 MWe version for the purpose of applying for design approval (DA). A series of public meetings with NRC has been held four times, and five technical reports have been submitted to NRC in preparation for DA application. Topics discussed in these meetings included, plant design, metallic fuel, safety design philosophy, safety analysis, measures against severe accident, phenomena identification and ranking table (PIRT), etc. Some useful comments and questions on the issues regarding the specific feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings. Among them, those items which are applicable to general sodium-cooled fast reactors, e.g. principal design criteria, guideline for safety analysis, validation and verification for safety analysis code, quality requirements, severe accident, and emergency planning are presented in this paper.     

对美国罗宾逊核电厂由火灾导致的反应堆紧急停堆和主泵丧失轴封冷却的运行事件的经验反馈

描述和分析了美国罗宾逊(H.B.Robinson)核电厂发生的一次由电缆故障引起的火灾并导致安注启动、主泵丧失轴封冷却的运行事件。介绍了操纵员处理事件的过程和失误。从设备、管理、人员培训等方面探究了事件的直接原因和根本原因,并针对这些原因进行总结,对我国核电厂的运行管理工作提出了具体建议。     

三门核电站停机不停堆的运行分析

三门核电AP1000反应堆在满功率情况下发生汽轮机故障停机事件时,通过快速降功率系统、旁排系统和棒控系统等的快速响应,一回路的参数不会突破安全限值,避免了反应堆停堆,降低了该瞬态对反应堆冷却剂系统的冲击。文章对停机不停堆的实现方式和运行特点进行了详细的分析和阐述,以帮助电站人员对停机不停堆的理解,并提高他们面临瞬态的响应能力。     

Trip setpoint analysis for the reactor protection system of an advanced integral reactor

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria.     

冷却剂流量降低停堆保护系统整定值分析

在确保反应堆安全的基础上 ,尽量扩大电厂的运行区域是反应堆停堆保护系统设计以及整定值确定的原则。本文通过对电网运行要求的分析 ,得到了恰希玛核电厂主泵低转速和一回路低流量停堆整定值 ,随后的安全验证表明了其对冷却剂流量降低事故保护的有效性     

Managing aging in nuclear power plants

Aging degradation in nuclear power plants must be controlled to prevent safety margins from declining below limits provided in plant design bases. The NPAR Program and other aging-related programs conducted under the auspices of the NRC Office of Research are developing needed technical guidance for control of aging. Results from these programs, together with relevant information developed by industry and elsewhere, are implemented through various ongoing NRC and industry programs and initiatives as well as by means of conventional regulatory instruments. The aging control process central to these efforts consists of three key elements: (1) selection of components, systems, and structures (CSS) in which aging must be controlled, (2) understanding of the mechanisms and rates of degradation in these CSS, and (3) managing degradation through effective surveillance and maintenance. These elements are addressed in Recommended Practices Guidance that integrates information developed under NPAR and other studies of aging into a systems-oriented format that tracks directly with the Safety Analysis Reports and with the NRC Standard Review Plan (NUREG-0800).     

多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

Experimental study on the natural circulation of a passive residual heat removal system for an integral reactor following a safety related event

Natural circulation characteristics of an integral type reactor during the operation of a passive residual heat removal system (PRHRS) following a safety related event has been experimentally investigated by using the VISTA facility. A PRHRS actuation trip signal is generated by a high power trip signal following a steam flow increasing event. The experimental results show that the single-phase coolant flows steadily in the primary loop by a natural convection process and that it effectively removes the decay heat from the core through a steam generator during the PRHRS operation. The heat transfers through the PRHRS heat exchanger and the emergency cooldown tank (ECT) are sufficient enough to enable a two-phase natural circulation of the coolant in the PRHRS loop.     

Assessment of FBR MONJU Accident Management Reliability in Causing Reactor Trips

This paper describes a method and application of quantitatively evaluating Accident Management (AM) reliability upon a reactor trip failure for the MONJU fast breeder reactor using a PSA technique. The present method comprises an allowable time estimation that is based on plant transient response analysis using the Super-COPD code that was developed for use in best estimates of the plant dynamics of MONJU and in estimating failure probability of operator's actions in AMs within the allowable time based on time records obtained from simulator training. Application of this method to MONJU resulted in the estimation that the allowable time for an unprotected loss-of-heat sink event would be more than the longest observed time of 326 s. The corresponding operation failure probability would be less than 0.1 even after taking the uncertainty into consideration. Combining this with a level 1 PSA revealed that the total frequency of core damage accompanying a reactor trip failure at MONJU could be decreased by at least 50 percent due to the reactor trip AM.     

核仪表系统系数设置影响因素分析

核电厂在投入商业运行前,需进行一系列的调试试验确保系统和运行的可靠性。2010年8月,某百万千瓦级核电厂在50％FP甩负荷到厂用电的调试中意外停堆,导致停运两天,拖延了试验的进度。经现场分析发现意外停堆与RPN系数设置不当有关。中国核动力研究设计院快速响应并对事件原因进行了分析,提出了新的RPN系数,顺利完成了50％FP和100％FP甩负荷到厂用电试验。文章首先分析了RPN系数对中子注量率变化率计算的影响,然后分析了影响RPN系数设置的主要因素,给出了RPN系数设置的基本原则,为后续电站的调试试验与运行提供了参考,以避免同类事件再次发生。     

多个核电项目堆内构件原材料PT漏检事件回顾与分析

堆内构件是核电厂反应堆冷却剂系统的主要设备。某制造厂在堆內构件制造过程中出现批量原材料PT漏检事件,造成大量人力和物力的浪费,影响到现场的工程进度。漏检事件反映出制造厂质量保证体系的缺陷,事件相关方应该加强质量管理和过程控制、做好经验反馈工作,提高我国的设备国产化水平,推动我国核电的平稳发展。     

Real time reactor noise diagnostics for the Borssele (PWR) nuclear power plant

After the upgrade of Borssele NPP in 1997, core cycle 24, the power plant operated three years more with 91% availability. The authority of the power plant decided to enhance and upgrade the reactor trend monitoring and plant information recording system with higher frequencies than the plant data processing system (PPS) as well as installing a flexible and multiple-purpose reactor noise analysis system which may support the reactor maintenance group with on-line and off-line capabilities for several different signal processing applications. Two measuring and monitoring systems were built in 2001 and fully taken in implementation during the start-up of the new core 28. In this sense, the new system was used in power operation during the 29^th of September 2001. This paper will introduce the measuring system, the operational tasks, and the results obtained so far on the real-time core-barrel motions (CBM) and the two-primary coolant pump vibrations measured through the reactor noise analysis.