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Security monitoring subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: Security Monitoring acquires and processes sensor data for use by security personnel in the performance of their function. Security Monitoring is designed and implemented as a part of an overall security plan which is classified as Safeguards Information under 10CFR73.21.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

Test and verification of a reactor protection system application-specific integrated circuit

Description: Application-specific integrated circuits (ASICs) were utilized in the design of nuclear plant safety systems because they have certain advantages over software-based systems and analog-based systems. An advantage they have over software-based systems is that an ASIC design can be simple enough to not include branch statements and also can be thoroughly tested. A circuit card on which an ASIC is mounted can be configured to replace various versions of older analog equipment with fewer design types required. The approach to design and testing of ASICs for safety system applications is discussed in this paper. Included are discussions of the ASIC architecture, how it is structured to assist testing, and of the functional and enhanced circuit testing.
Date: March 1, 1997
Creator: Battle, R.E.; Turner, G.W.; Vandermolen, R.I.; Vitalbo, C. & Naser, J.
Partner: UNT Libraries Government Documents Department

Fire protection system operating experience review for fusion applications

Description: This report presents a review of fire protection system operating experiences from particle accelerator, fusion experiment, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of fire protection system component failure rates and fire accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with these systems are discussed, including spurious operation. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor.
Date: December 1, 1995
Creator: Cadwallader, L.C.
Partner: UNT Libraries Government Documents Department

Seismic active control by a heuristic-based algorithm

Description: A heuristic-based algorithm for seismic active control is generalized to permit consideration of the effects of control-structure interaction and actuator dynamics. Control force is computed at onetime step ahead before being applied to the structure. Therefore, the proposed control algorithm is free from the problem of time delay. A numerical example is presented to show the effectiveness of the proposed control algorithm. Also, two indices are introduced in the paper to assess the effectiveness and efficiency of control laws.
Date: May 1996
Creator: Tang, Yu
Partner: UNT Libraries Government Documents Department

Defense against common mode failures in protection system design

Description: The introduction of digital instrumentation and control into reactor safety systems creates a heightened concern about common-mode failure. This paper discusses the concern and methods to cope with the concern. Common-mode failures have been a ``fact-of-life`` in existing systems. The informal introduction of defense-in-depth and diversity (D-in-D&D)-coupled with the fact that hardware common-mode failures are often distributed in time-has allowed systems to deal with past common-mode failures. However, identical software operating in identical redundant systems presents the potential for simultaneous failure. Consequently, the use of digital systems raises the concern about common-mode failure to a new level. A more methodical approach to mitigating common-mode failure is needed to address these concerns. Purposeful introduction of D-in-D&D has been used as a defense against common-mode failure in reactor protection systems. At least two diverse systems are provided to mitigate any potential initiating event. Additionally, diverse displays and controls are provided to allow the operator to monitor plant status and manually initiate engineered safety features. A special form of conimon-mode failure analysis called ``defense-in-depth and diversity analysis`` has been developed to identify possible conimon-mode failure vulnerabilities in digital systems. An overview of this analysis technique is provided.
Date: August 27, 1997
Creator: Wyman, R. H. & Johnson, G. L.
Partner: UNT Libraries Government Documents Department

Aging assessment for active fire protection systems

Description: This study assessed the impact of aging on the performance and reliability of active fire protection systems including both fixed fire suppression and fixed fire detection systems. The experience base shows that most nuclear power plants have an aggressive maintenance and testing program and are finding degraded fire protection system components before a failure occurs. Also, from the data reviewed it is clear that the risk impact of fire protection system aging is low. However, it is assumed that a more aggressive maintenance and testing program involving preventive diagnostics may reduce the risk impact even further.
Date: June 1, 1995
Creator: Ross, S.B.; Nowlen, S.P. & Tanaka, T.
Partner: UNT Libraries Government Documents Department

Design and testing of integrated circuits for reactor protection channels

Description: Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built.
Date: June 1, 1995
Creator: Battle, R. E.; Vandermolen, R. I.; Jagadish, U.; Swail, B. K. & Naser, J.
Partner: UNT Libraries Government Documents Department

Effects of Gamma Radiation on Individual and Mixed Ion Exchange Resins

Description: The ion exchange resins that are used to deionize moderator in the reactor purification systems may accumulate sufficient radiation dose to damage the resins. This radiation damage would be manifested by: (1) loss of useful exchange capacity of the bed, which is costly since resins from the reactor deionizers are not reused; (2) shrinking or swelling of the resins, which may have some effect on the hydraulic behavior of the beds; (3) release of resin degradation products into the process stream, which pollutes moderator with impurities and precursors of the neutron-induced radioisotopes. This document details results of a laboratory study to determine the magnitude of these three effects by gamma irradiation of individual resins and their mixtures.
Date: January 6, 2003
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Statistical analysis of random duration times

Description: This report presents basic statistical methods for analyzing data obtained by observing random time durations. It gives nonparametric estimates of the cumulative distribution function, reliability function and cumulative hazard function. These results can be applied with either complete or censored data. Several models which are commonly used with time data are discussed, and methods for model checking and goodness-of-fit tests are discussed. Maximum likelihood estimates and confidence limits are given for the various models considered. Some results for situations where repeated durations such as repairable systems are also discussed.
Date: April 1, 1996
Creator: Engelhardt, M.E.
Partner: UNT Libraries Government Documents Department

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

Description: The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity.
Date: November 19, 1997
Creator: Heard, F.J.
Partner: UNT Libraries Government Documents Department

Scientific design of Purdue University Multi-Dimensional Integral Test Assembly (PUMA) for GE SBWR

Description: The scaled facility design was based on the three level scaling method; the first level is based on the well established approach obtained from the integral response function, namely integral scaling. This level insures that the stead-state as well as dynamic characteristics of the loops are scaled properly. The second level scaling is for the boundary flow of mass and energy between components; this insures that the flow and inventory are scaled correctly. The third level is focused on key local phenomena and constitutive relations. The facility has 1/4 height and 1/100 area ratio scaling; this corresponds to the volume scale of 1/400. Power scaling is 1/200 based on the integral scaling. The time will run twice faster in the model as predicted by the present scaling method. PUMA is scaled for full pressure and is intended to operate at and below 150 psia following scram. The facility models all the major components of SBWR (Simplified Boiling Water Reactor), safety and non-safety systems of importance to the transients. The model component designs and detailed instrumentations are presented in this report.
Date: April 1, 1996
Creator: Ishii, M.; Ravankar, S.T. & Dowlati, R.
Partner: UNT Libraries Government Documents Department

Recent results of an experimental study on the impact of smoke on digital I and C equipment

Description: A program to assess the impact of smoke on digital Instrumentation and Control (I and C) safety systems began in 1994, funded by the US Nuclear Regulatory Commission Office of Research. Digital I and C safety systems are likely replacements for today`s analog systems. The nuclear industry has little experience in qualifying digital electronics for critical systems, part of which is understanding system performance during plant fires. The results of tests evaluating the performance of digital circuits and chip technologies exposed to the various smoke and humidity conditions representative of cable fires are discussed. Tests results show that low to moderate smoke densities can cause intermittent failures of digital systems. Smoke increases leakage currents between biased contacts, leading to shorts. Chips with faster switching times, and thus higher output drive currents, are less sensitive to leakage currents and thus to smoke. Contact corrosion from acidic gases in smoke and inductance of stray capacitance are less important contributors to system upset. Transmission line coupling was increased because the smoke acted as a conductive layer between the lines. Permanent circuit damage was not obvious in the 24 hrs of circuit monitoring. Test results also show that polyurethane, parylene, and acrylic conformal coatings are more effective in protecting against smoke than epoxy or silicone. Common-sense mitigation measures are discussed. Unfortunately industry is a long way from standard tests for smoke exposure that capture the variations in smoke exposure possible in an actual fire.
Date: October 1, 1997
Creator: Tanaka, T.J. & Antonescu, C.
Partner: UNT Libraries Government Documents Department

Performance requirements of the advanced neutron source reactor protection system

Description: Research reactors often have protection-systems performance requirements very different from those of commercial reactors. This paper discusses the special characteristics of the Advanced Neutron Source (ANS) reactor that control these requirements, and it presents some calculations used to quantify this performance.
Date: April 1, 1995
Creator: March-Leuba, J. & Battle, R.E.
Partner: UNT Libraries Government Documents Department

Design and testing of integrated circuits for reactor protection channels

Description: Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing.
Date: June 1, 1995
Creator: Battle, R. E.; Vandermolen, R. I.; Jagadish, U.; Swail, B. K.; Naser, J. & Rana, I.
Partner: UNT Libraries Government Documents Department

Software safety hazard analysis

Description: Techniques for analyzing the safety and reliability of analog-based electronic protection systems that serve to mitigate hazards in process control systems have been developed over many years, and are reasonably well understood. An example is the protection system in a nuclear power plant. The extension of these techniques to systems which include digital computers is not well developed, and there is little consensus among software engineering experts and safety experts on how to analyze such systems. One possible technique is to extend hazard analysis to include digital computer-based systems. Software is frequently overlooked during system hazard analyses, but this is unacceptable when the software is in control of a potentially hazardous operation. In such cases, hazard analysis should be extended to fully cover the software. A method for performing software hazard analysis is proposed in this paper.
Date: February 1, 1996
Creator: Lawrence, J.D.
Partner: UNT Libraries Government Documents Department

Testing existing software for safety-related applications. Revision 7.1

Description: The increasing use of commercial off-the-shelf (COTS) software products in digital safety-critical applications is raising concerns about the safety, reliability, and quality of these products. One of the factors involved in addressing these concerns is product testing. A tester`s knowledge of the software product will vary, depending on the information available from the product vendor. In some cases, complete source listings, program structures, and other information from the software development may be available. In other cases, only the complete hardware/software package may exist, with the tester having no knowledge of the internal structure of the software. The type of testing that can be used will depend on the information available to the tester. This report describes six different types of testing, which differ in the information used to create the tests, the results that may be obtained, and the limitations of the test types. An Annex contains background information on types of faults encountered in testing, and a Glossary of pertinent terms is also included. This study is pertinent for safety-related software at reactors.
Date: December 1, 1995
Creator: Scott, J. A. & Lawrence, J. D.
Partner: UNT Libraries Government Documents Department

Reactor scram report for period of July 1, 1965--December 31, 1965

Description: The reactor scrams are summarized in a table which identifies the component.that opens the safety circuit and the reason or cause for the safety circuit trip. Caution should be exercised in the use of the summary for specific analysis of reactor scram causes. In a high percentage of cases, the component that opens the safety circuit is not at fault. A description of the reason or cause for a particular scram is given in the body of the report. The actual outage time charged to the cause of the reactor scram is given with the exception of scrams caused by tube leaks or fuel failures.
Date: February 23, 1966
Creator: Newell, L. J.
Partner: UNT Libraries Government Documents Department

Aging study of boiling water reactor high pressure injection systems

Description: The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.
Date: March 1, 1995
Creator: Conley, D.A.; Edson, J.L. & Fineman, C.F.
Partner: UNT Libraries Government Documents Department

Temperature and liquid level control monitor, port plug (fabrication only)

Description: Requirements for fabrication, testing, inspection, cleaning, packaging, delivery, and on-site handling of the mechanical system for a Temperature and Liquid Level Monitor (TLLM) Port Plug Assembly are presented. The TLLM port plug provides access, support, and protection for the plant protection system (PPS) thermocouples and liquid level probes required to measure reactor sodium outlet plenum temperatures and monitor sodium levels within the reactor vessel. Each TLLM assembly consists of three PPS removable thermocouple assembly units and capabilities to accept at least one removable PPS liquid level probe assembly. The TLLM assembly shall consist of but not be limited to closed thimbles to accommodate a dry thermowell temperature sensor and dry inductive coil-type liquid level sensor, respectively. The TLLM is designed to provide an adequate structural, thermal, and paramagnetic environment and primary system boundary seal for the sensors and to minimize radiation streaming into the head compartment. (auth)
Date: October 1, 1973
Partner: UNT Libraries Government Documents Department

Westinghouse Reactor Protection System Unavailability, 1984--1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. Westinghouse commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Rasmuson, D. & Marksberry, D.
Partner: UNT Libraries Government Documents Department

General Electric Reactor Protection System Unavailability, 1984--1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Hamzehee, H. & Rasmuson, D.
Partner: UNT Libraries Government Documents Department