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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

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

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

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

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

[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

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

Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

Description: The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events.
Date: May 29, 1981
Creator: Louison, R. & Boardman, C.E.
Partner: UNT Libraries Government Documents Department

SEP operating history of the Dresden Nuclear Power Station Unit 2

Description: 206 forced shutdowns and power reductions were reviewed, along with 631 reportable events and other miscellaneous documentation concerning the operation of Dresden-2, in order to indicate those areas of plant operation that compromised plant safety. The most serious plant challenge to plant safety occurred on June 5, 1970; while undergoing power testing at 75% power, a spurious signal in the reactor pressure control system caused a turbine trip followed by a reactor scram. Subsequent erratic water level and pressure control in the reactor vessel, compounded by a stuck indicator pen on a water level monitor-recorder and inability of the isolation condenser to function, led to discharge of steam and water through safety valves into the reactor drywell. No significant contamination was discharged. There was no pressure damage or the reactor vessel of the drywell containment walls. Six areas of operation that should be of continued concern are diesel generator failures, control rod and rod drive malfunctions, radioactive waste management/health physics program problems, operator errors, turbine control valve and EHC problems, and HPCI failures. All six event types have continued to recur.
Date: January 1, 1983
Creator: Mays, G.T. & Harrington, K.H.
Partner: UNT Libraries Government Documents Department

Experimental and analytical studies of passive shutdown heat removal systems

Description: Using a naturally circulating air stream to remove shutdown decay heat from a nuclear reactor vessel is a key feature of advanced liquid metal reactor (LMR) concepts developed by potential vendors selected by the Department of Energy. General Electric and Rockwell International continue to develop innovative design concepts aimed at improving safety, lowering plant costs, simplifying plant operation, reducing construction times, and most of all, enhancing plant licensability. The reactor program at Argonne National Laboratory (ANL) provides technical support to both organizations. The method of shutdown heat removal proposed employs a totally passive cooling system that rejects heat from the reactor by radiation and natural convection to air. The system is inherently reliable since it is not subject failure modes associated with active decay cooling systems. The system is designed to assure adequate cooling of the reactor under abnormal operating conditions associated with loss of heat removal through other heat transport paths.
Date: January 1, 1987
Creator: Pedersen, D.; Tessier, J.; Heineman, J.; Stewart, R.; Anderson, T.; August, C. et al.
Partner: UNT Libraries Government Documents Department

Self-actuating reactor-shutdown system. [LMFBR]

Description: A control system for the automatic or self-actuated shutdown or scram of a nuclear reactor is described. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.
Date: June 4, 1981
Creator: Barrus, D.M.; Brummond, W.A. & Peterson, L.F.
Partner: UNT Libraries Government Documents Department

Gap and impact of LMR (Liquid Metal Reactor) piping systems and reactor components

Description: Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content.
Date: January 1, 1987
Creator: Ma, D.C.; Gvildys, J. & Chang, Y.W.
Partner: UNT Libraries Government Documents Department

Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies

Description: Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system.
Date: April 21, 1985
Creator: Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H. & Wantland, J.L.
Partner: UNT Libraries Government Documents Department