Aging management of instrumentation & control sensors in nuclear power plants

Pressure to improve plant efficiency and maximize safety and the increasing age of existing NPPs are forcing the global nuclear power industry to confront the challenges of aging - caused by stressors such as temperature, humidity, radiation, electricity, and vibration - in key instrument & control (I&C) components like pressure transmitters, temperature sensors, neutron detectors, and cables. Traditional aging management methods, such as equipment replacement, required the process to be shut down. Recent aging management technologies, collectively known as online monitoring (OLM), enable plants to monitor the condition and aging of their installed I&C while the plant is operating. Developed through R&D initiatives worldwide, such OLM techniques include low- and high-frequency methods that use existing sensors, such as noise analysis; methods based on test or diagnostic sensors, such as for vibration-measuring accelerometers; and methods, such as the power interrupt (PI) test, based on active measurements made by injecting a test signal into the component under test. A review of these aging management methods, their effectiveness, and their interrelation provides a foundation for understanding the next stage in the evolution of OLM: truly integrated hybrid OLM systems capable of robust condition monitoring in both novel and familiar operating conditions.     

Experiences with diagnostic instrumentation in nuclear power plants

Over the past several years, Westinghouse has developed a coordinated system of on-line diagnostic instrumentation for the acquisition and analysis of data for diagnostics and incipient failure detection of critical plant equipment and systems. Primary motivation for this work is to improve NSSS availability and maintainability through the detection of malfunctions at their inception. These systems encompass the following areas: (1) Vibration Monitoring System for detection of changes in vibrational characteristics of the major components of Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP); (2) Acoustic Monitoring System for detection and location of leaks in the primary system pressure boundary and other piping systems in PWRs; (3) Metal Impact Monitoring for detection of loose debris in the reactor vessel and steam generators; (4) Nuclear Noise Monitoring System for monitoring core barrel vibration; (5) Sensor Response Time Measurement System for detecting any degradation of process sensors; and (6) Transit Time Flow Meter for determining primary coolant flow rate. Summarized in this paper are some of the features of the systems and in-plant experience. These experiences demonstrate that diagnostic systems in combination with analytical and laboratory work for data interpretation do improve plant availability.     

On-line monitoring applications in nuclear power plants

The nuclear power industry is working to reduce generation costs by adopting condition-based maintenance strategies and automating testing activities. These developments have stimulated great interest in on-line monitoring (OLM) technologies and new diagnostic and prognostic methods to anticipate, identify, and resolve equipment and process problems and ensure plant safety, efficiency, and immunity to accidents. This paper provides examples of these technologies with particular emphasis on eight key OLM applications: detecting sensing-line blockages, testing the response time of pressure transmitters, monitoring the calibration of pressure transmitters on-line, cross-calibrating temperature sensors in situ, assessing equipment condition, performing predictive maintenance of reactor internals, monitoring fluid flow, and extending the life of neutron detectors. These applications are discussed in the following sections. Emphasis is placed on the principles of a core OLM method - noise analysis - and the technical requirements for an integrated OLM system are summarized.     

Probabilistic descriptions of resistance of safety-related structures in nuclear plants

Calculations of reliability of safety-related nuclear plant structures require a knowledge of the probability distributions that describe their resistance. This study considers the applicability of existing statistical data for describing the resistance of steel and reinforced concrete nuclear plant structures. Probability distributions are recommended which can be used in assessing the reliability of containments and other Category I structures, developing fragilities, and selecting appropriate resistance criteria for probability-based structural design.     

Replacement of safety-related valves in Philippsburg 1 nuclear power plant

The Philippsburg nuclear power plant Unit 1, initially commissioned in 1979 is presently still partly equipped with safety-related isolation valves which correspond to the technical standard of the 1970s. Inspection of these valves according to today's technical rules revealed that a number of these isolation valves can be upgraded in such a way that compliance with the present state of the art is achieved. However, the advanced state of wear of frequently operated valves suggested that a program should be set up for the gradual replacement of these valves, with a view to this upgrading. At present, the replacement program relates to a period from 1993–1997 with the selection of valves to be replaced depending on the signs of wear and the operational possibilities. Within this period, it is planned to replace 50% of the safety related valves with those that have been in compliance with the latest technical rules, and that have been improved with regard to that resistance to wear and ease of maintenance.     

压水堆核电厂仪表控制与计算机化的发展概况

阐述了当今世界压水堆核电站（ＰＷＲ）仪表与控制及自动化设备的发展概况;描述了模拟仪表与控制所存在的缺陷和问题;重点论述了当代先进核电站数字化仪表控制、保护系统与先进主控制室的性能和对计算机化仪表控制提出和要求。     

核反应堆棒控电源电路仿真与优化

棒控电源是核电站反应堆燃料棒的驱动机构的电源,核电站内的棒控电源柜通常与驱动机构相距较远,通过电缆相连接。为了提高设备的集成度,棒控电源柜中多个直流斩波模块共用一个整流模块,这种设计大大增加了直流母线的长度。直流母线和电缆的分布电感和分布电容,会产生电压过冲、增加开关损耗、产生高频振荡,并对电源控制产生巨大干扰。通过对电源电路的分析,计算直流母线的分布电感和电缆的分布电容,建立棒控电源系统PSIM仿真模型,得到了与样机测试波形近似的仿真结果。然后提出了减小直流母线分布电感的优化方法,经理论推导和仿真验证,可有效改善高频振荡和电压过冲等问题,显著提高电路性能指标。     

Single event upset mitigation techniques for FPGAs utilized in nuclear power plant digital instrumentation and control

Field programmable gate arrays (FPGAs) are integrated circuits being increasingly used for digital instrumentation and control (I&C) in nuclear power plants (NPPs) because of low cost, re-configurability and low design turn-around time. However, to ensure reliability, proper design techniques must be employed since the memory and logic in FPGAs are susceptible to single event upsets (SEUs). Triple modular redundancy (TMR) has become a common SEU mitigation design technique because of its straightforward implementation and reliable results. Partitioned TMR approaches are introduced in this paper, and formulae derived indicate that the maximum probability of two simultaneous errors [P_E]_max is inversely proportional to the number of logic partitions in a TMR design, when each redundant logic block in every logic partition has the same number of sensitive nodes. However, the maximum logic partitioning design cannot completely eliminate the possibility of two simultaneous upsets. For the example test circuit it is found that [P_E]_max is reduced dramatically from 66.67% for minimum logic partitioning to 4.44% for maximum logic partitioning. Because TMR introduces significant overhead due to its full hardware redundancy, a dual modular redundancy approach is also examined for application to less demanding situations. By comparative analysis this study reaches the conclusion that the maximum logic partitioning TMR implementation is the best solution for digital I&C applications in NPPs where obtaining robustness is of the highest importance, despite its higher area overhead.     

核电仪控技术应用中的基本问题

就核电仪控技术应用中的基本问题作了简要概述,希望能引起对基础性技术工作研究的足够重视。基本问题分为理论基础问题和技术应用问题,内容涉及核反应堆物理、热工和控制仪表技术。这些问题一方面导致研发工作方向性失误;另一方面反映出在工程应用中基础性研究工作的缺失。     

Seismic instrumentation for nuclear power plants: An interpretative review of current practice and the related standard in Germany

The German nuclear safety standard KTA 2201: “Design of nuclear power plants against seismic events”, consists of the following parts: 1. basic principles; 2. characteristics of seismic excitation; 3. design of structural components; 4. design of mechanical and electrical parts; 5. seismic instrumentation; and 6. measures subsequent to earthquakes.While Part 1 was published in June 1975, Part 5 was approved by the Nuclear Safety Standards Commission — Kerntechnischer Ausschuss (KTA) — in June 1977. The other parts are still under development. The requirements of the safety standard KTA 2201.5 deal with

1. (a) number of location (number and location of acceleration recording systems for different sites, single-block plants and multi-block plants);

2. (b) characteristics of instruments (readiness and operation of instruments, margin or errors, dynamic and operation characteristics, duration of records, seismic switch);

3. (c) triggering and information (loss of electric power, start of the acceleration recording systems, threshold of acceleration for triggers and seismic switches, optical and acoustic information); and

4. (d) documentation (results of recordings, inspection and tests).

The purpose of this paper is to present some detailed requirements of the safety standard KTA 2201.5, with its philosophy, and compare these with corresponding requirements in the US. It will be shown that with relatively few instruments, which are very reliable in operation and in triggering, an optimum of data may be available after an earthquake.     

Welding consumables for nuclear power plants

In nuclear power plants, submerged arc welding and covered arc welding have long been employed especially for main weld seams, including the core region of RPV.This paper investigates the mechanical properties of several welding consumables we have developed for industrial plants — that is, welding consumables which lower the phosphorus and copper content of the welded metal, those for plates possessing particularly high tensile strength and those for the narrow gap welding method.Recent data derived from irradiation embrittlement tests show that these welded metals using a non-copper coating are highly effective in minimizing shifts in the transition curve.Welding consumables for A533B C1.2, A543 C1.1 or A508 C1.4 steels have a higher tensile strength than those for A533B C1.1 or A508 C1.3.We have developed submerged arc and covered arc welding consumables to be used with these kinds of steels, and it was confirmed that these consumables possess excellent tensile strength and notch toughness.Our tests also confirmed that the narrow gap SAW and MIG welds are more efficient than the conventional ones. Moreover, the mechanical properties of the welded metals are also excellent.     

Flood hazards for nuclear power plants

Flooding hazards for nuclear power plants may be caused by various external geophysical events. In this paper the hydrologic hazards from flash floods, river floods and heavy rain at the plant site are considered. Depending on the mode of analysis, two types of hazard evaluation are identified: (a) design hazard which is the probability of flooding over an expected service period, and (b) operational hazard which deals with real-time forecasting of the probability of flooding of an incoming event. Hazard evaluation techniques using flood frequency analysis can only be used for type (a) design hazard. Evaluation techniques using rainfall-runoff simulation or multi-station correlation can be used for both types of hazard prediction.     

The real-time functional test facility for advanced instrumentationand control in nuclear power plants

Testing and validation of the functions and performance of the digital instrumentation and control (I&C) system should be done prior to installation in nuclear power plants. The objective of the I&C Functional Test Facility (FTF) is to test and validate the functions of developed digital control and various monitoring systems. The FTF provides the simulated testing environment as an experimental test bed. The FTF software consists of a mathematical modeling program which simulates a three-loop 993 MWe pressurized water reactor and a supervisory program that comprises all the instructions necessary to run the FTF. The hardware equipment provides an interface between a host computer and a simple test panel or the developed target systems to be tested. The graphical user interface supports an easy and friendly interface between the FTF and users. It is implemented through a Picasso-3 graphic tool developed by the Halden Reactor Project. The FTF is applied to an advanced I&C system prototype, such as an automatic start-up intelligent control system, dynamic alarm system, accident identification system, and intelligent logic tracking system, to test its algorithm or performance. The results of the test show good operational performance of the FTF in normal and transient conditions     

Regulatory practices for nuclear power plants in India

Atomic Energy Regulatory Board (AERB) is the national authority to ensure that the use of ionizing radiation and nuclear energy in India does not cause undue risk to the health of workers and members of the public, and the environment. AERB fulfills its mission by stipulating and enforcing rules and regulations concerned with nuclear and radiological safety. This paper describes the regulatory authorization process of AERB as applicable to nuclear power plants (NPPs) during their construction as well as operating phases. The safety review process during construction is presented as case studies. Some current issues related to operating plants are also described. Two typical examples of safety upgradations made in old generation nuclear plants are given.The regulatory process in India is continuously evolving to cater to the new developments. Some of the recent initiatives taken by AERB in that direction are briefly described. Today AERB faces new challenges like simultaneous review of a large number of new projects of diverse designs, a fast growing nuclear power program and functioning of operating plants in a competitive environment. This paper delineates how AERB is gearing up itself to meet these challenges in an effective manner.     

Alexisismon isolation engineering for nuclear power plants

This paper reviews the special requirements regarding efficiency, licensibility (reliability) and cost which should be met to achieve an optimum base isolated nuclear power plant design. It then describes the Alexisismon-2, patented isolation system developed by the author, underlines its original properties (linearity and separation of functions) and presents a conceptual design of its application to a nuclear power plant. The great reliability of the system components is demonstrated. The efficiency of the A-2 is found to be very high: a reduction factor of the base shear induced in the plant higher than 25 is achieved for all examined real accelerograms scaled to 1 g GPA. So the isolation components, the structural system of the plant, its equipment and systems can be easily designed to remain in the elastic range of stresses and strain even for seismic input with GPA higher than 2 g.     

Inherently safe containments for nuclear power plants

Nuclear energy cannot be avoided in the near future. To regain public acceptance the safety of nuclear power plants has to be increased. Consequently, feasibility studies have been carried out for a containment proposal for future pressurized water reactors which will keep people unharmed even in the case of severe nuclear accidents under the assumption “all that can go wrong, will go wrong”. The main features of the design concept are a core melt cooling and retention device, a passively acting cooling system to remove the decay heat and a double-wall containment which is able to withstand high static and dynamic internal pressures due to hydrogen detonation. Internal structures are designed to resist extreme loadings resulting from various accident scenarios including in-vessel steam explosion and vessel failure under high system pressure.     

Unitized concept for earthquake-resistant nuclear power plants

This paper discusses a unitized concept for an earthquake-resistant nuclear power plant which can withstand major earthquake shaking and fault slips without releasing radioactive material into the atmosphere. A 1000 MWe pressurized water reactor power plant of recent design is adapted to a unitized concept, and cost studies are made for the incremental cost.