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Plant Closings, Mass Layoffs, and Worker Dislocations: Data Issues

Description: For at least 15 years Members of Congress have continued to ask: How many U.S. manufacturing plants have closed? For at least 15 years they have continued to ask: How many U.S. manufacturing plants have relocated abroad, and where have they gone? For at least 15 years the answer has been: For the most part, those questions can't be answered, based on Government data. How many plants are moving to Mexico? What industries and what States are the plants from? How many U.S. workers are losing their jobs as a result? It appears that still, after two legislative attempts to mandate collection of these data, the Government publishes no counts of U.S. plant closings, and almost no information on plant relocations. Options for strengthening the data systems include addressing three main weaknesses: inadequate data program design, a plant closing definition that misses its mark, and publication of partial instead of complete survey results.
Date: March 29, 1993
Creator: Bolle, Mary Jane
Partner: UNT Libraries Government Documents Department

Plant closing: advance notice and rapid response: special report

Description: The report also assesses the ability of public agencies to provide worker adjustment services rapidly and effectively when employers do give notice. Much of the benefit of advance notice depends on the prompt provision of effective services.
Date: September 1986
Creator: United States. Congress. Office of Technology Assessment.
Partner: UNT Libraries Government Documents Department

Resilience of Microgrid during Catastrophic Events

Description: Today, there is a growing number of buildings in a neighborhood and business parks that are utilizing renewable energy generation, to reduce their electric bill and carbon footprint. The most current way of implementing a renewable energy generation is to use solar panels or a windmill to generate power; then use a charge controller connected to a battery bank to store power. Once stored, the user can then access a clean source of power from these batteries instead of the main power grid. This type of power structure is utilizing a single module system in respect of one building. As the industry of renewable power generation continues to increase, we start to see a new way of implementing the infrastructure of the power system. Instead of having just individual buildings generating power, storing power, using power, and selling power there is a fifth step that can be added, sharing power. The idea of multiple buildings connected to each other to share power has been named a microgrid by the power community. With this ability to share power in a microgrid system, a catastrophic event which cause shutdowns of power production can be better managed. This paper then discusses the data from simulations and a built physical model of a resilient microgrid utilizing these principles.
Date: May 2018
Creator: Black, Travis Glenn
Partner: UNT Libraries

Criticality alarm system verification at the Los Alamos critical experiments facility: Past experience and present capabilities

Description: The Los Alamos Critical Experiments Facility (LACEF) has been involved in the testing and evaluation of criticality accident alarm systems since 1980. At that time we designed and built the solution critical assembly SHEBA for this purpose. The response of the alarms to neutron pulses was done using the GODIVA-IV Burst Assembly. Currently, SHEBA and a new fast assembly, SKUA, are being modified for burst operation. 8 refs., 7 figs., 2 tabs.
Date: January 1, 1988
Creator: Plassmann, E.A. & Spriggs, G.D.
Partner: UNT Libraries Government Documents Department

Verifying seismic design of nuclear reactors by testing. Volume 1: test plan

Description: This document sets forth recommendations for a verification program to test the ability of operational nuclear power plants to achieve safe shutdown immediately following a safe-shutdown earthquake. The purpose of the study is to develop a program plan to provide assurance by physical demonstration that nuclear power plants are earthquake resistant and to allow nuclear power plant operators to (1) decide whether tests should be conducted on their facilities, (2) specify the tests that should be performed, and (3) estimate the cost of the effort to complete the recommended test program.
Date: July 20, 1979
Partner: UNT Libraries Government Documents Department

The CEBAF (Continuous Electron Beam Accelerator Facility) fast shutdown system

Description: Because of the high power in the CEBAF beam, equipment must be protected in the event of beam loss. The policy that has been adopted is to require a positive permissive signal from each of several inputs in order to operate the gun that starts the beam. If the permissive is removed, the gun shuts off within 20 {mu}s. The inputs that are now monitored include radiation monitors that detect beam loss directly, vacuum monitors (which also observe the status of various in-line valves), and general input from the rf system, which combines detection of klystron failure, arcs, and rf window high temperature. The system is expandable, so other fault detectors can be added if experience shows their necessity.
Date: September 1, 1990
Creator: Perry, J. & Woodworth, E.
Partner: UNT Libraries Government Documents Department

The role of SASSYS-1 in LMR (Liquid Metal Reactor) safety analysis

Description: The SASSYS-1 liquid metal reactor systems analysis computer code is currently being used as the principal tool for analysis of reactor plant transients in LMR development projects. These include the IFR and EBR-II Projects at Argonne National Laboratory, the FFTF project at Westinghouse-Hanford, the PRISM project at General Electric, the SAFR project at Rockwell International, and the LSPB project at EPRI. The SASSYS-1 code features a multiple-channel thermal-hydraulics core representation coupled with a point kinetics neutronics model with reactivity feedback, all combined with detailed one-dimensional thermal-hydraulic models of the primary and intermediate heat transport systems, including pipes, pumps, plena, valves, heat exchangers and steam generators. In addition, SASSYS-1 contains detailed models for active and passive shutdown and emergency heat rejection systems and a generalized plant control system model. With these models, SASSYS-1 provides the capability to analyze a wide range of transients, including normal operational transients, shutdown heat removal transients, and anticipated transients without scram events. 26 refs., 16 figs.
Date: January 1, 1988
Creator: Dunn, F.E. & Wei, T.Y.C.
Partner: UNT Libraries Government Documents Department

Markovian analysis of limiting conditions of operation for the reactor protection system

Description: The main conclusions of the point value calculations are supported by the uncertainty analysis. An uncertainty analysis was performed by the Monte Carlo sampling technique with the Markov model to assess the effect of the uncertainty in the data base on the results of point value calculations. A modified version of the SAMPLE program in WASH-1400 that can receive multiple outputs from MARELA was used. Comparison between two alternatives characterized by uncertainties usually requires a preference assessment (under certainty). In some special cases, namely, in the case of stochastic dominance, the comparison is straightforward. Stochastic dominance means that the likelihood that the attribute of interest will be less than a specific value is always larger for one alternative than for the other. Policy 1 stochastically dominates (is better than) Policy 2 on the ATWS core damage probability while Policy 2 stochastically dominates Policy 1 on the spurious scram core damage probability. As far as the total core damage probability is concerned, Policy 1 stochastically dominates Policy 2. However, Policy 2 stochastically dominates Policy 1 on the average reactor downtime.
Date: January 1, 1985
Creator: Papazoglou, I.A. & Cho, N.Z.
Partner: UNT Libraries Government Documents Department

Thermionic switched self-actuating reactor shutdown system

Description: A self-actuating reactor shutdown system is described which has a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the control rod in its ready position (exterior of the reactor core).
Date: June 4, 1981
Creator: Barrus, D.M.; Shires, C.D. & Brummond, W.A.
Partner: UNT Libraries Government Documents Department

Procedure for determining the SSE response from the OBE response. [Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (OBE)]

Description: Regulatory Guide 1.61 specifies the damping that should be used for all modes that are considered in an elastic spectral or time history dynamic seismic analysis of Seismic Category I components. Table 1 of R.G 1.61 specifies damping values for dynamic analysis for two different earthquakes, the Safe Shutdown Earthquake and the Operating Basis Earthquake. The guide specifies that ''...if the maximum stresses due to static, seismic and other dynamic loading are significantly lower than the yield stresses and 1/2 yield stress for SSE and 1/2 SSE respectively, in any structure a component damping values lower than those specified in Table 1 ....should be used .... to avoid underestimating the amplitude of vibration of dynamic stress.'' The guide requires that the appropiate damping values be used which reflect the state of stress that will be experienced by the equipment. In applying these values to the response of equipment, to an OBE and to an SSE, the selected damping should result in a dynamic response for the SSE that is greater than the response due to the OBE, all other factors being equal. The purpose of the statement in the guide is to note that at higher stress levels, the higher damping values could be used, but at lower stress levels, the lower values of damping should be used. Current procedures that are used in implementing R.G. 1.61 frequently result in an OBE response that is greater than the SSE response. This is because the higher damping under the SSE is used at all stress levels, low as well as high. This is obviously not the intent of the Regulatory Guide. A procedure has been developed which derives an expression relating the SSE response to the OBE response. Two factors are involved in the equation. The first involves the damping ratios ...
Date: January 1, 1985
Creator: Curreri, J.
Partner: UNT Libraries Government Documents Department

[News Clip: FAA]

Description: Video footage from the KXAS-TV/NBC station in Fort Worth, Texas, to accompany a news story.
Date: December 11, 1979, 10:00 p.m.
Creator: KXAS-TV (Television station : Fort Worth, Tex.)
Partner: UNT Libraries Special Collections

Evaluations of the generalized area and the. cap alpha. /. nu. methods of interpreting pulsed neutron measurements for subcritical reactivity

Description: The Generalized Area and ..cap alpha../..nu.. Methods of interpretation of pulsed-neutron measurements for subcritical reactivity were found to minimize the error caused by source-induced harmonics and kinetic distortion. The use of multiple neutron detectors distributed over the core of the reactor in the pulsed source measurements is recommended for increased accuracy of interpretation. The measured data are reduced to a reported value of the subcritical reactivity by the use of numerical solutions to the reactor eigenvalue problems. In the Generalized Area Method, the numerical solution provides an estimate of the static adjoint function which is used to weight the prompt and delayed neutron flux integrals measured in the experiment. These weighted integrals are then used to form the subcritical reactivity. In the ..cap alpha../..nu.. Method, the static eigenequation solved by numerical methods is transformed to a time eigenequation to provide a bridge between the measured decay constant of the fundamental mode and the subcritical static reactivity. Also, the transformation provides a means of normalizing the reported static reactivity. The evaluations were performed by applying both methods to numerical data generated by one-dimensional, space--time diffusion theory for which k/sub eff/ was known precisely. The Generalized Area Method was found to deduce reasonably accurate reactivity values from simulations of data from pulsed-neutron experiments in both large and small reflected reactors. The errors from the true reactivity ranged from +3% to -7% in rho at k/sub eff/ = 0.9. However, the single-detector Sjoestrand analysis failed badly for all pulsing simulations except for that of the small reactor with the source at the center of the core. Errors from the true reactivity ranged from +90% to -260% in rho at k/sub eff/ = 0.9. The ..cap alpha../..nu.. Method was applicable only to the small-reactor simulation where the error was -1% in rho at k/sub ...
Date: August 1, 1978
Creator: Parks, P.B.
Partner: UNT Libraries Government Documents Department

Geothermal heating project at St. Mary's Hospital, Pierre, South Dakota. Final report

Description: St. Mary's Hospital, Pierre, South Dakota, with the assistance of the US Department of Energy, drilled a 2176 ft well into the Madison Aquifer ot secure 108/sup 0/F artesian flow water at 385 gpm (475 psig shut-in pressure). The objective was to provide heat for domestic hot water and to space heat 163,768 sq. ft. Cost savings for the first three years were significant and, with the exception of a shutdown to replace some corroded pipe, the system has operated reliably and continuously for the last four years.
Date: December 1, 1984
Partner: UNT Libraries Government Documents Department

Intermediate leak protection/automatic shutdown for B and W helical coil steam generator

Description: The report summarizes a follow-on study to the multi-tiered Intermediate Leak/Automatic Shutdown System report. It makes the automatic shutdown system specific to the Babcock and Wilcox (B and W) helical coil steam generator and to the Large Development LMFBR Plant. Threshold leak criteria specific to this steam generator design are developed, and performance predictions are presented for a multi-tier intermediate leak, automatic shutdown system applied to this unit. Preliminary performance predictions for application to the helical coil steam generator were given in the referenced report; for the most part, these predictions have been confirmed. The importance of including a cover gas hydrogen meter in this unit is demonstrated by calculation of a response time one-fifth that of an in-sodium meter at hot standby and refueling conditions.
Date: January 1, 1981
Partner: UNT Libraries Government Documents Department

An analysis of loss of offsite power with a PWR at shutdown

Description: In many Probabilistic Risk Assessments (PRAs), loss of offsite power (LOOP) when a nuclear power plant is operating was found to be a significant contributor to core damage. The purpose of this study is to provide an analysis of a loss of offsite power event that occurs while a PWR is shut down. The importance of such an analysis was recognized as part of a study to evaluate the core damage frequency due to a loss of decay heat removal capability during an outage. 5 refs., 1 fig.
Date: June 1, 1987
Creator: Chu, T.L.; Yoon, W.H. & Fitzpatrick, R.G.
Partner: UNT Libraries Government Documents Department

Design criteria for a self-actuated shutdown system to ensure limitation of core damage. [LMFBR]

Description: Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times.
Date: September 1, 1981
Creator: Deane, N.A. & Atcheson, D.B.
Partner: UNT Libraries Government Documents Department

Component failures at pressurized water reactors. Final report

Description: Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.
Date: October 1, 1980
Creator: Reisinger, M.F.
Partner: UNT Libraries Government Documents Department

Interpretation of fuel centerline thermocouple response to reactor scrams

Description: This report compares and contrasts fuel thermocouple scram responses from the low-burnup assemblies IFA-513 and IFA-505 and the high-burnup assembly IFA-432. Even on a qualitative basis it is observed that the IFA-432 thermocouple responses are sluggish relative to that of low-burnup counterparts in rods of equivalent thermal resistance and design/enrichment. By numerical analysis, it is concluded that this apparent sluggishness is due to thermocouple decalibration, and an estimate of the decalibration is made.
Date: January 1, 1980
Creator: Lanning, D.D. & Cunningham, M.E.
Partner: UNT Libraries Government Documents Department

Uncertainty reduction requirements in cores designed for passive reactivity shutdown

Description: The first purpose of this paper is to describe the changed focus of neutronics accuracy requirements existing in the current US advanced LMR development program where passive shutdown is a major design goal. The second purpose is to provide the background and rationale which supports the selection of a formal data fitting methodology as the means for the application of critical experiment measurements to meet these accuracy needs. 6 refs., 1 fig., 2 tabs.
Date: January 1, 1988
Creator: Wade, D.C.
Partner: UNT Libraries Government Documents Department

Report to the US Nuclear Regulatory Commission on Analysis and Evaluation of Operational Data, 1986

Description: This annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during calendar year 1986. Comments and observations are provided on operating experience at nuclear power plants and other NRC licensees, including results from selected AEOD studies; summaries of abnormal occurrences involving US nuclear plants; reviews of licensee event reports and their quality, reactor scram experience from 1984 to 1986, engineered safety features actuations, and the trends and patterns analysis program; and assessments of nonreactor and medical misadministration events. In addition, the report provides the year-end status of all recommendations included in AEOD studies, and listings of all AEOD reports issued from 1980 through 1986.
Date: May 1, 1987
Creator: none,
Partner: UNT Libraries Government Documents Department

Self-actuated shutdown system for a commercial size LMFBR. Final report

Description: A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility and reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.
Date: August 1, 1978
Creator: Dupen, C.F.G.
Partner: UNT Libraries Government Documents Department

Validation of ANS-5. 1 as the decay heat standard at the Savannah River Plant

Description: The Savannah River Laboratory (SRL) is upgrading the methods used to predict the post shutdown decay heat of the Savannah River reactors by implementing procedures based on the ANS Decay Heat Power in Light Water Reactors standard. This approach takes advantage of the large volume of research used in developing the standard and establishes compatibility with the nuclear industry. To qualify the decay heat standard for use, a series of comparisons were made between detailed decay heat calculations performed using the SHIELD code system and results obtained from the standard.
Date: January 1, 1982
Creator: Apperson, Jr, C E
Partner: UNT Libraries Government Documents Department