Reactor Safety Systems Research Articles

Modal Decomposition Method for BWR Stability Analysis.

In-phase and out-of-phase instabilities have been observed in BWR reactors. To improve the safety systems of these reactors, it is necessary to be able to detect in a reliable way these oscillations from the neutronic signals. In this paper a methodology to decompose the neutronic signals in its modal amplitudes is proposed. This decomposition is based on the normal and the adjoint dominant Lambda modes of a static configuration of the reactor core. The calculation of these eigenmodes for a realistic problem is reviewed and the oscillation parameters for the modal decompositions of the neutronic signals from Ringhals reactor have been calculated using the proposed methodology.

A Comparative Risk Assessment for the Russian V213 Power Reactor

Probabilistic risk assessment methodology is applied to generate an evaluation of the relative likelihood of safe recovery following selected pressurized water reactor (PWR) design basis accidents for a Russian V213 nuclear power reactor. US-designed PWRs similar to the V213 are used for reference and comparison. This V213 risk assessment is based on comparison analyses of the following aspects: accident progression event tree success paths for typical PWR accident initiating events, safety aspects in reactor design, and perceived performance of reactor safety systems. The four initiating events considered here are: loss of offsite power with station blackout, large-break loss-of-coolant accident (LOCA), medium-break LOCA, and small-break LOCA. The success probabilities for the V213 reaching a non-core-damage state after the onset of the selected initiating events are calculated for two scenarios: (a) using actual component reliability data from US PWRs and (b) assuming common component reliability data. US PWR component reliability data are used based of the unavailability of such data for the V213 at the time of the analyses. While the use of US PWR data in this risk assessment of the V213 does strongly infer V213 comparability to US plants, the risk assessment using common component reliability does not have suchmore » a stringent limitation and is thus a separate scoping assessment of the V213 engineered safety systems. The results of the analyses suggest that the V213 has certain design features that significantly improve the reactor`s safety margin for the selected initiating events and that the V213 design has a relative risk of core damage for selected initiating events that is at least comparable to US PWRs. It is important to realize that these analyses are of a scoping nature and may be significantly influenced by important risk factors such as V213 operator training, quality control, and maintenance procedures.« less

Passive modular gas safety system for a reactor

Reactor safety systems have developed gradually. Today in particular, auxiliary systems are being developed which are based on nontraditional operational concepts, by using gaseous neutron absorbers. The Scientific-Research and Design Institute of Power Technology (NIKIET) and the Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center (RNTs), have done preliminary development and experimental verification of separate elements of this system, in which helium is used as the absorber. This article presents a rapid passive safety system based on gaseous absorber, which is made as autonomous modules as the final stage of reactor safety. Its effectiveness is discussed by using an RBMK reactor as an example. As opposed to traditional active, systems, it does not require a functioning power supply and information signals from outside the reactors system, which makes it stable against unsanctioned actions by personnel, the influence of other systems, and also outside actions (sabotage and natural calamities which could destroy the the nuclear power plant structure). Because the gas safety system can operate instantaneously (0.1-0.3 sec), in principle, it can shut down the reactor even with fast-neutron runaway, where traditional safety systems are ineffective.

Power cutbacks as an alternative to scrams: an application of the MIT-SNL period generated minimum time control laws

A cutback is a rapid reduction in a reactor's neutronic power caused by the high-speed insertion of the control devices. The advantage of using cutbacks in lieu of scrams is that the cutback may make the reactor only slightly subcritical while still significantly decreasing the power. Thus, the thermal stress associated with a complete shutdown is avoided, and a rapid return to normal operation is possible. The physical basis for cutbacks is reviewed. and possible areas of application are enumerated. At present, cutbacks are implemented by inserting the control devices at a prespecified speed. The power will decrease, but neither the profile nor the magnitude of the decrease is specified. As a result, cutbacks are not used extensively in reactor safety systems. Use of the Massachusetts Institute of Technology-Sandia National Laboratories (MIT-SNL) period-generated minimum time control laws to rectify these deficiencies is proposed.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Formal Specification of the Software for a Reactor Safety System

For achievement of highly dependable software the use of formal specification techniques is a must, and an associated tool support is necessary.As an example for a technique and tool environment, the specification system CIP is introduced. CIP (Computer-aided, Intuition-guided Programming) was originally developed at the Technical University of München. It supports the formal specification of software and assists the program development by stepwise transformation. A high level specification is transformed by the application of rules into lower level representation, the lowest level being the programming language level. The transformations preserve the correctness of the specifications.The use of this environment is demonstrated by an example from a safety related reactor control system. Advantages of this approach as well as drawbacks and limitations are explained.

Risk and reactor safety systems adoption

How do firms react to the risk of accidents that disable their industrial plant? We examine this question by modeling the electric utility's decision to install reactor safety systems. The optimal adoption time is a function of the probability of a serious accident, the cost of the safety system, potential damages, and insurance premiums. We apply an econometric version of our model to the empirical evidence regarding implementation of a computer graphics warning system for nuclear reactors. We find that (1) firms that operate multiple reactors at one site, have higher equipment failures, or have higher discount rates, were quicker to adopt the safety system; (2) owners of larger plants were slower to adopt; and (3) the size of the population surrounding the plant, the state regulatory climate in which the reactor was operating, and the value of the initial plant investment had no influence on the speed of adoption.

Safety at nuclear power plants: Economic incentives under the Price-Anderson Act and state regulatory commissions

The Price-Anderson Act has indemnified the commercial nuclear power industry since its inception. This article proposes a present value model of firm investment in reactor safety systems to investigate the safety incentives of the 1988 amendments to the Act. While the amendments increase safety incentives for electric utilities through insurance regulation, safety behavior depends crucially on the anticipated reaction of the state regulatory commission after a reactor accident.

Recent Results at the Tritium Systems Test Assembly

The Tritium Systems Test Assembly (TSTA) project at Los Alamos is charged with developing and demonstrating the tritium technology required for the magnetic fusion energy program. Since June 1987 the project has been operated by Los Alamos National Laboratory under the joint funding of the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). At TSTA the technology under development includes reactor exhaust gas processing technology and safety systems. This project is not currently investigating the processing of tritium extracted from a breeding blanket. The processing of tritium extracted from a breeding blanket is a possible upgrade to TSTA. This possible upgrade is the subject of an ongoing study among Los Alamos, JAERI and the Argonne National Laboratory (ANL). This installing such on upgraded process systems at TSTA in the 1990--1992 time frame. The exhaust gas processing system at TSTA is designed to demonstrate full-scale exhaust gas processing for a fusion machine of a size currently being designed for ITER. Of course, deuterium is also a major component of the exhaust gas. Current gas processing uses a total gas stream comprised of /approximately/65% deuterium, 35% tritium, 1-2% reactive impurities (currently nitrogen and methane) with the remainingmore » gas hydrogen and helium. The tritium inventory required to process through this loop at this rate is about 110 g. The main process loop at TSTA has been operating at this flow rate since July 1987. In addition to the operation of the main process loop, a number of nonloop experiments are underway at TSTA. This paper will update the status of both loop and nonloop studies at Los Alamos.« less

Testing of compton-emission type detectors of neutrons and of triaxial fission chambers in a pulsed reactor

Compton-emission type neutron detectors with an emitter containing hafnium or gadolinium and small triaxial fission chambers are promising detectors for monitoring the energy liberation in nuclear reactors. By contrast to the B-emission type neutron detectors which are widely employed in power reactors and have an emitter of rhodium, silver, or vanadium [1-3], the Compton-emission type neutron detectors with an emitter containing hafnium or gadolinium are almost without the inertia 14, 5] which is characteristic of fission detectors. Therefore the Compton-emission type neutron detectors and the fission chambers are preferably employed in automatic monitoring and controlling systems and in safety systems of reactors where the detectors must meet increased requirements in regard to fast response. A technology which is used in analogy in the manufacture of coaxial and triaxial cables with magnesia insulation is used in the manufacture of Compton emission type neutron detectors and fission chambers. We tested three Compton-emiss~on type neutron detectors in cable form with a hafniumcontaining emitter, three Compton-emission type neutron detectors of the same form with a gadolinium-containing emitter, and four triaxial small-scale KtV fission chambers [6]. In regard to their design, the Compton-emission type neutron detectors resembled the previously tested G-emission type neutron detectors with a silver emitter [7]. Measurements which were made during the manufacture of the Compton-emission type neutron detectors and during their operation in a research reactor have shown that the mean-square spread of the emitter diameters of these detectors and of their initial sensitivity amount to about 1.5%. The sensie tive element of the Compton-emission type neutron detector with the hafnium-containing emitter has an outer diameter of 3.0 m~n and a length of 0.83 m; the detector with the gadolinium-containing emitter has an outer diameter of 3.4 mm and a length of 0.85 m. Thedesign of the fission chambers hardly differs from that described in [8]. The characteristics of the small axial fission chambers with an outer diameter of 6.0 mm and one to two 50-mm-long sensitive elements were studied. In the case of two sensitive elements, the collecting electrode of one of them was without a fissile cover, i.e., it was used to monitor the dose rate of the y radiation. The gap between the collecting electrode and the housing of the fission chamber was about 0.7 mmwide. The operative volume of the chamber was filled with argon ofanalysis-type purity under a pressure of 0.101 MPa (760 mm Hg) at a temperature of 20~ The operating voltage applied to the electrodes of the triaxial chamber was changed from 200 to 300 V; this is much higher than in the operation and testing of the chambers in power reactors and research reactors [6]. The increased operating voltage results from the irradiation intensity which in the pulsed graphite reactor is higher than in the reactors indicated above. Magnesium oxide powder of analysis-type purity corresponding to GOST 452675 was the insulation of the Compton-emission type neutron detectors; magnesium oxide powder

Health test of components in nuclear reactor instrument systems using process identification

A process identification technique using an auto regressive model has been applied to measurements at Ringhals 4, a PWR in Sweden. The purpose of the present work is to develop a method for testing the dynamics of sensors and signal conditioning circuits in safety systems as well as the regulator parameters during normal operation of the reactor. The noise analysis was performed for a coolant temperature safety system and a pressure regulator in a loop for control of boron concentration in the reactor. The results show convincingly that filter characteristics as well as temperature sensor response time can be estimated with good accuracy. Therefore the present method enables us to evaluate integral performance of instrumentation chains in a safety system during reactor operation. The same conclusion holds for the analysis of the controller. On-line application of the present method to test components in the reactor safety system would increase the system reliability. It also results in reduction of manpower by replacing the component test, performed manually, by automated on-line tests during the reactor operation.

Issues in the Closed-Loop Digital Control of Reactor Power: The MIT Experience

The 5 MWt MIT Research Reactor's license was recently amended to permit the general use of a closedloop digital control system for the adjustment of reactor power. This regulatory approval is significant because the avoidance of challenges to the reactor's safety system now depends on the design of the controller rather than, as was previously the case, on the characteristics of the actuator. Considerations involved in the granting of this license amendment are discussed including information on the technical basis for the controller, the prevention of challenges to the safety system, and the relation between the control and safety systems. The results of the initial use of the controller are presented together with a discussion of the possible applications of this controller.

Modern Control Technology for Improved Nuclear Reactor Performance

One of the main complaints leveled at reactor control systems by utility spokesmen is complexity. One only has to look inside a power reactor control room to appreciate this viewpoint. The high reliability and versatility of modern microprocessors makes possible distributed control systems with only performance data and abnormal conditions being relayed to the control room. In a sense, this emulates the human-body control system where routine repetitive actions are handled in an involuntary manner. The significance of expert systems to the nuclear reactor control and safety systems is their ability to capture human and other expertise and make it available, upon demand, and under almost all circumstances. Thus, human problem-solving skills acquired by the learning process over a long period of time can be captured and employed with the reliability inherent in computers. This is especially important in nuclear plants when human operators are burdened by stress and emotional factors that have a dramatic effect on performance level.

Maintainability analysis using Q-GERT network simulation

In reliability analyses of nuclear reactor safety systems, time- dependent phenomena (such as component failure during op eration and the ensuing repair activity) are often disregarded or treated quasi-statically because the time dependencies involved are not amenable to closed-form analytic formulation. In addi tion, conditional phenomena (such as a failure mode that depends on the failure mode in a previous event) and phenom ena that must be characterized by arbitrary probability distribu tions are often ignored. This paper reports the initial applica tion of the computer program Q-GERT, an existing code that uses dynamic Monte Carlo modeling, to reactor safety-system simulation and analysis. Using Q-GERT with its Monte Carlo techniques, the above shortcomings can, in principle, be dealt with, and the analysis can be elevated from the level of a reliabil ity analysis to that of a maintainability analysis. This new applica tion of Q-GERT, called the Maintainability Analysis Procedure (MAP), is illustrated by simulating and evaluating the behavior of a simple safety system as this system is subjected to each of four possible failure processes. These processes are distinguished by the number and complexity of the included phenomena, and the effects on estimates of the probability of system failure are shown to be significant as the processes become more complex.

Design Strategy for Control of Inherently Safe Reactors

Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. "Inherently safe" reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended.

Uncertainty analysis of reactor safety systems with statistically correlated failure data

The probability of occurrence of the top event of a fault tree is estimated from failure probability of components that constitute the fault tree. Component failure probabilities are subject to statistical uncertainties. In addition, there are cases where the failure data are statistically correlated. Most fault tree evaluations have so far been based on uncorrelated component failure data. The subject of this paper is the description of a method of assessing the probability intervals for the top event failure probability of fault trees when component failure data are statistically correlated. To estimate the mean and variance of the top event, a second-order system moment method is presented through Taylor series expansion, which provides an alternative to the normally used Monte-Carlo method. For cases where component failure probabilities are statistically correlated, the Taylor expansion terms are treated properly. A moment matching technique is used to obtain the probability distribution function (pdf) of the top event through fitting a Johnson S B distribution. The computer program (CORRELATE) was developed to perform the calculations necessary for the implementation of the method developed. The CORRELATE code is very efficient and consumes minimal computer time. This is primarily because it does not employ the time-consuming Monte-Carlo method.

A MOCVD reactor safety system for a production environment

MOCVD growth of GaAs frequently is accomplished using trimethylgallium and arsine. Although these materials are presently used safely worldwide in a research environment, their hazardous nature is an important consideration in the application of MOCVD to production. This paper describes a MOCVD reactor safety-interlock and automatic shutdown system now being evaluated on a research reactor for its suitability in a production environment. Some specific features employed are fire detection, reactor overpressure prevention, and toxic gas exposure control. This paper will also discuss the technique used to remove As and AsH 3 from the reactor exhaust.

A comparison of methods for calculating system unavailability due to common cause failures: The beta factor and multiple dependent failure fraction methods

The beta factor method for calculating common cause failures has been used in many probabilistic risk analyses to evaluate redundant systems reliability. The multiple dependent failure fraction (MDFF) method is a generalization of the beta factor (BF) method, and is extended here to derive reliability expressions for several multiple redundant systems. To demonstrate the difference in results that can arise by applying these methods, two example systems were analyzed consisting of reactor safety systems made up of two and three redundant trains (the high pressure injection and auxiliary feedwater systems). It is shown that the MDFF method is less conservative and produces more realistic calculations of system unavailability by factors of between two and 20 for the two versus three train systems, respectively. Additionally, both approaches produce higher values than does an analysis that excludes common cause effects, indicating that system unavailability may be grossly underestimated by several orders of magnitude if common cause effects are neglected. For these reasons, it is recommended that the MDFF method be utilized in redundant systems analysis as it produces more realistic results.

Implementation of an automated status analysis system in an operating nuclear power plant

A description is presented of a computer system designed to perform real-time, automatic analysis of the implications of arbitrary degraded states of a reactor's safety systems. In such states, one or more elements of one or more systems is inoperable, and the analysis problem is to determine the implications of this status with respect to reactor operations and the technical specifications stipulated by the regulatory authorities. The computer system is programmed to accomplish this analysis and to communicate the results to the reactor operator. The system has been designed for use by the reactor operating staff both for making real-time decisions when any of the reactor's systems are in an unanticipated state and also for planning maintenance and test activities. To verify the efficacy of the design concept an experimental version of the system (RGB-F) has been implemented at the Swedish Forsmark 1 nuclear power plant.

Transient heat transfer and fluid mechanics of a recirculating pressurized water loop during blowdown and cold water injection

AbstractMathematical models, in the form of computer codes, are used to evaluate the performance of process and safety systems of CANDU nuclear reactors during normal operation and postulated accident conditions. To validate the models, predictions are made of the transient heat transfer and fluid mechanics processes in pressurized water loops following simulated pipe ruptures. These loops have the essential geometric and physical characteristics of a CANDU reactor primary heat transport system. Comparisons with data from loops of different scales over the conditions of interest (blowdown and cold water injection) are used to illustrate that the models predict the trends in the experiments adequately when the assumptions inherent in the model (e. g. equal steam and water velocities) are valid. To improve accuracy in cases where these assumptions do not apply, a more sophisticated representation of two‐phase flow is required.

Modular fault tree synthesis—A new method for computer-aided fault tree construction

A new methodology for the construction of fault tree is developed. The methodology is based on fault tree modules constructed for different types of component circuits and stored in a module library. The modules are developed for nuclear reactor safety system applications. The methodology has been implemented into the CAFTS (computer-aided fault tree synthesis) computer code written in PL/1 language. This code enables fault trees of large systems to be constructed interactively with the computer, so that a more accurate and faster analysis of the system reliability can be performed.