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The international scenario and the requirements formulated for a German digital reactor protection system are described in the following. The hardware structure of the distributed multicomputer system and particular aspects resulting from the use of computers are discussed. In view of the multitude of factors influencing the design of such a system, this presentation can give no more than a general outline.

India was one of the first countries in the world to have realised a computer-based reactor protection system. It was in the late 1980s when fully fledged microprocessor-based reactor protection units were commissioned for research and then for power reactors. Since then the proprietary philosophy, applied to have proprietary electronics in reactor protection and control, has expanded to include a dozen nuclear power reactors. What we get from proprietary electronics, as much as reasonably achievable, are online fault coverage and fail-safe features, which are not available in commercial electronics. From design to fabrication and then maintainability, the process is traceable, controllable and sustainable. Therefore the entire modular electronics is proprietary and not commercial. The record of plant availability for the last two decades for the dozen reactors speaks for it. Conservation of signal properties and T-states, along with proprietary fault diagnostic features make the system truly robust.

The author establishes the essential information that is needed so that an adequate engineering and probabilistic assessment of the reactor protection system (RPS) can be carried out. This required information takes the form of questions that the assessor needs to ask of the system to aid his analysis. The scram (reactor-shutdown) function of the RPS is used as an example. >

Two unusual aspects of safety component operation--abnormal environmental conditions and aging--are combined analytically. The result is implemented in a computer code to quantify a safety system figure-of-merit. In conjunction with a probabilistic approach employing a specified risk criterion, the analytical method has been applied to a reactor protection system to determine a desired component-qualified age. The effect of a preventive renewal program as a substitute for or complement to an age qualification program is also discussed. Substantial economic benefits should be realized from the implementation of the analytical method.

Reactor Protection Systems for Nuclear Power Plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock & Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. This paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System.

The reactor protection system (both analog- and digital-based) setpoints of Combustion Engineering nuclear steam supply systems were examined to determine the feasibility of scram reduction by relaxing these setpoints. Representative safety analyses, using RETRAN-02, were performed to demonstrate that acceptable results were obtained with the relaxes setpoints. The steam generator low level, reactor trip on a turbine trip, and thermal margin/low pressure trip setpoints were successfully relaxed.

In a nuclear power generating station, the system used to protect the reactor and its containment is the plant protection sytem. The pressurized-water-reactor (PWR) protection systems encompass all electrical and mechanical devices and circuitry (from sensors to actuated device input terminals) involved in generating those signals associated with the protective function. These signals include actuation signals for the Reactor Protection System (RPS) and Engineered Safety Features System (ESFS) such as Safety Injection Actuation System, Containment Spray Actuation System, Main Steam Isolation System, Auxiliary Feedwater Actuation System, and Recirculation Actuation System. This paper deals with the present status and future development of the Plant Protection Systems utilized by Combustion Engineering, Inc. (C-E) and Kraftwerk Union (KWU) for their Pressurized Water Reactors. The latest techniques of signal processing, logic design, reactor trip and engineered safety features actuation circuitry, automatic testing, isolation and interfacing, are discussed. Current system design criteria, which as a result of the dynamic national and international licensing atmospheres influence system design philosophies, are highlighted.

An analysis of a boiling water reactor/4 plant response to a simultaneous closing of all main steam isolation valves (MSIVs) followed by a failure of the reactor protection system scram function wa...

The paper describes the full life cycle process of design, development and validation, which has been followed for the production of software for an experimental computerized reactor protection system at Casaccia Center. The aim of the paper is to put in emphasis phase- by-phase the criteria which have been followed, the verification procedures adopted and the errors discovered. Additionally the system testing activity is described.

The CANDU shut-down system comprises electro-mechanical shutdown rods and liquid poison injection, each of which includes sensors, instrument channels and mechanical and fluidic subsystems. Published work so far has reported ∼10 −5 for the unavailability of the reactor protection system. The basic component failure rate is assumed to be constant and to have a lognormal distribution. Mathematical models are developed to analyse the CANDU shutdown system, with aging of basic components. The Weibull distribution is used because it has a lower standard deviation. An unavailability of ∼10 −4 is obtained. Time constraint on system operation and aging of components over a year do not significantly affect system unavailability.

This paper presents a design procedure for selecting the optimal parameters of a filter used in extracting true signals from noisy reactor measurements. The equations for the optimal filter are presented and a procedure is given for approximating the optimal filter with a simpler and less expensive structure. A design example demonstrates how the method may be applied to a reactor protection system.

A dual-channel reactor protection system similar to the one that was developed for the Dresden Nuclear Power Station is described. This protection system utilizes commercially available components in coincidence circuits to achieve the system reliability that is required for power plant service.

An analysis is presented concerning the performance characteristics of sensors considered for use in protection systems of liquid metal fast breeder reactors. Alternate solutions to protection systems having one or more reactor-trip channels per fuel subassembly are proposed. Primary emphasis is placed on pressure pulse sensors, installed in the upper (and possibly also lower) coolant plenum, capable of sensing the collapse of sodium vapor bubbles. It is shown that, even for the limit-case of a complete coolant blockage of one or more fuel subassemblies, a reactor protection system based on pressure pulse sensors should be capable of effecting reactor-shutdown prior to generation of any substantial amount of molten fuel. Some preliminary experimental data regarding vapor bubble collapse are presented, showing reasonable agreement with analytical predictions. Recommendations are made regarding some potentially promising areas for further research and development with respect to protection systems for liquid metal fast breeder reactors.

A Common Mode Failure (CMF) study is described which supports the design basis for the Reactor Protection System on a General Electric Boiling Water Reactor. Ground rules are formulated which restrict the study to functional CMFs applied one at a time to a reactor plant subjected to initiating transients and accidents. Each outcome is traced to a stable condition and the results examined to determine that they are acceptable. CMF studies seem to parallel the development of Reliability studies and the results appear to be complimentary.

This review program is utilized during the design of a nuclear power plant to insure separation between interdiscipline design for the Reactor Protection System (RPS) and Engineered Safeguard Systems (ESS). Color coded transparent drawings of the RPS and ESS are produced by each discipline. The separation is then reviewed by overlaying drawings of different disciplines on a light table. When this inspection shows that RPS or ESS elements have less than the established minimum separation, an analysis is performed to determine what, if any, design revision is necessary to insure proper separation. "Hazard" drawings are also made for determination of each type of potential hazard in each area of the plant. The review is a continuing process as the design progresses and is revised by any discipline. A complete writeup of apparent separation violations and resolution of the apparent violations is made for each stage of the separation review.

The paper describes the full life cycle process of design, development and validation, which has been followed for the production of software for an experimental computerized reactor protection system at Casaccia Center. The aim of the paper is to put in emphasis phase- by-phase the criteria which have been followed, the verification procedures adopted and the errors discovered. Additionally the system testing activity is described.

This report documents the technical evaluation of the request for an amendment to Operating License NPF-4 for changes in the Technical Specifications for the North Anna Power Station, Unit 1. These changes were proposed by the licensee, Virginia Electric and Power Company, in its submittal of October 13, 1978. The basis for review included a report entitled Westinghouse Reactor Protection System/Engineered Safety Features Actuation System Setpoint Methodology. The requested changes to the Technical Specifications are found to be acceptable based on the information submitted by the licensee.

This report documents the technical evaluation of the application to amend the Technical Specifications for the Ft. Calhoun Unit No. 1 Nuclear Generating Plant. The review criteria are based on the Technical Specifications of St. Lucie and Calvert Cliffs, IEEE Standards, Combustion Engineering Standard Technical Specifications, and the Code of Federal Regulations. The evaluation compares the submittal made by the licensee with the NRC staff position and the review criteria and presents the reviewer's conclusion on the acceptability of the application to amend the Technical Specifications.

Using operational experience for plant parameters, the probability that the peak reactor coolant system pressure will exceed allowable pressure has been evaluated for the Loss of Feedwater (LOFW) Anticipated Transient Without Scram (ATWS) for the Standard 3800 MWt, 205 fuel assembly, Babcock and Wilcox reactor. The probabilistic approach to ATWS is reviewed including current estimates of anticipated transients' frequency, probability of scram system failure, and probability of excessive peak pressure. The result of the present study is a best estimate probability of 0.18 that an allowable pressure of 3750 psia would be exceeded if an ATWS were to occur and 0.6 that 3200 psia would be exceeded. Considering all uncertainties and using estimates of transient frequency and scram system failure, it is found that the frequency of significant ATWS events is smaller than 1.1 x 10/sup -6/ per reactor per year. Thus it is concluded that the ATWS safety goal of approximately 10/sup -6/ per reactor per year has been met with the current design. A proposed redundant Backup Scram System (BUSS) to the present generation reactor protection system would reduce the estimate to approximately 7.8 x 10/sup -7/ per reactor per year. Therefore, compliance with th ATWS safety goalmore » is demonstrated.« less

This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Brunswick Steam Electric Plant, Units 1 and 2. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications with time delays verified by GE, will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources.

This report documents the technical evaluation of the noise and isolation testing of the reactor protection system (RPS) for the Davis Besse Nuclear Power Station, Unit 1. The testing was performed in accordance to Section 4.6.11, Susceptibility, of MIL-N-19900B, and NRC approved plant test methods. Analysis of the test results shows that the reactor protection system did not degrade below acceptable levels when subjected to electromagnetic, electrostatic, isolation and noise level tests, nor was the system's ability to perform its Class 1E protective functions affected.

This is the fourth and final volume of Part II of this series of studies. The three earlier volumes of Part II considered the fraction of total potential risk attributable to ATWS (measured in REM, 30 day whole body dose). This analysis was based upon the methods of WASH 1400. The results obtained there showed that ATWS is responsible for less than 5 percent of the total LWR risk. This fourth volume considers the probability of an ATWS releasing sufficient radiation so as to violate the radiation limits set by 10CFR100. Calculations show that the probability is less than 3 x 10/sup -7/ per year (3 chances in 10 million). Since WASH 1270 stated that a desirable probability would be on the order of 10/sup -7/ per year, it would appear that such a limit has already been met.

This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Pilgrim Nuclear Power Station. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources.

After the Fukushima nuclear accident, safety of nuclear power plants (NPPs) is widespread concerned. The reliability of reactor protection system (RPS) is directly related to the safety of NPPs, however, it is difficult to accurately evaluate the reliability of digital RPS. The method is based on estimating probability has some uncertainties, which can not reflect the reliability status of RPS dynamically and support the maintenance and troubleshooting. In this paper, the reliability quantitative analysis method based on extenics is proposed for the digital RPS (safety-critical), by which the relationship between the reliability and response time of RPS is constructed. The reliability of the RPS for CPR1000 NPP is modeled and analyzed by the proposed method as an example. The results show that the proposed method is capable to estimate the RPS reliability effectively and provide support to maintenance and troubleshooting of digital RPS system.