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In-Vessel Retention Technology Development and Use for Advanced PWR Designs in the USA and Korea

Description: In-Vessel Retention (IVR) of molten core debris by means of external reactor vessel flooding is a cornerstone of severe accident management for Westinghouse's AP600 (advanced passive light water reactor) design. The case for its effectiveness (made in previous work by the PI) has been thoroughly documented, reviewed as part of the licensing certification, and accepted by the US Nuclear Regulatory Commission. A successful IVR would terminate a severe accident, passively, with the core in a stable, coolable configuration (within the lower head), thus avoiding the largely uncertain accident evolution with the molten debris on the containment floor. This passive plant design has been upgraded by Westinghouse to the AP1000, a 1000 MWe plant very similar to the AP600. The severe accident management approach is very similar too, including In-Vessel Retention as the cornerstone feature, and initial evaluations indicated that this would be feasible at the higher power as well. A similar strategy is adopted in Korea for the APR1400 plant. The overall goal of this project is to provide experimental data and develop the necessary basic understanding so as to allow the robust extension of the AP600 In-Vessel Retention strategy for severe accident management to higher power reactors, and in particular, to the AP1000 advanced passive design.
Date: May 18, 2004
Creator: Theofanous, T. G.; Oh, S. J. & Scobel, J. H.
Partner: UNT Libraries Government Documents Department


Description: The APPR-I is described and the various hazards are reviewed. Because of the reactor's location near the nation's Capitol, containment is of the utmost importance. The maximum energy release in any possible accident is 7.4 million Btu's which is completely contained within a 7/8 inch thick steel cylindrical shell with hemispherical ends. The vapor container is 60 ft high and 32 ft in diameter and is lined on the inside with 2 ft of reinforced concrete which provides missile protection and is part of the secondary shield. All possible nuclear excursions are reviewed and the energy from any of these is insignificant compared to the stored energy in the water. The maximum credible accident is caused hy the reactor running constantly at its maximum power of 10 Mw and through an extremely unlikely sequence of failures, causing the temperature of the water in the primary and secondary systeras to rise to saturation; whereupon a rupture occurs releasing the stored energy of 7.4 million Btu's into the vapor container. If the reactor core melts during the incident, a maximum of 10/sup 8/ curies of activity is released. While it appears impossible for a rupture of the vapor container to oecur except by sabotage or bombing, the hazards to the surrounding area are discussed in the event of such a rupture occurring simultaneously with the maximum credible accident. (auth)
Date: July 27, 1955
Partner: UNT Libraries Government Documents Department

Facilitation of Decommissioning Light Water Reactors

Description: Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented in this report. A wide range of facilitation methods--from improved documentation to special decommissioning tools and techniques--is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given.
Date: December 1, 1979
Creator: Moore, E. B., Jr.
Partner: UNT Libraries Government Documents Department

PNL Technical Review of Pressurized Thermal Shock Issues Supplement 1: Technical Critique of the NRC Near-Term Screening Criteria

Description: Pacific Northwest Laboratory (PNL) provided a technical critique of the draft report, NRC Staff Evaluation of Pressurized Thermal Shock, dated September 13, 1982. This report provided the basis for the NRC near-term regulatory position on pressurized thermal shock {PTS) and recommended a generic screening criteria for welds in the vessel beltline region. The PNL staff concluded that the screening criteria were adequate to meet the intent of the NRC safety goal and to retain past predictions of vessel reliability. The conclusion was based on selecting the plant-specific nilductility transition reference temperature (RT{sub NDT}) in the conservative manner described within the staff report. Conservative and unconservative factors were mentioned throughout the NRC staff report. The PNL staff has listed these factors together with unknown (may be either conservative or unconservative) factors and estimated, where possible, the range in °F RT{sub NDT}. The unknown factors were so widespread that the PNL staff recommended that specific conservatisms not be reduced until the unknowns are further resolved.
Date: May 1, 1983
Creator: Pederson, L. T.; Apley, W. J.; Bian, S. H.; Pelto, P. J.; Simonen, E. P.; Simonen, F. A. et al.
Partner: UNT Libraries Government Documents Department

An Independent Assessment of Evacuation Time Estimates for A Peak Population Scenario in the Emergency Planning Zone of the Seabrook Nuclear Power Station

Description: This study comprises two major tasks. First, it includes an independent assessment of the methods and assumptions used in calculating evacuation time estimates (ETEs) applicable to the general population for a peak population scenario in the emergency planning zone {EPZ) of the Seabrook Nuclear Power Station. This consists of a review and analysis of previous work by Public Service of New Hampshire {PSNH) and the Federal Emergency Management Agency (FEMA), as well as an independent calculation of evacuation times using the CLEAR model for the demographic data reported by PSNH. Secondly, this study includes independent estimations of evacuation time for the peak population scenario developed using demographic data prepared by the U.S. Nuclear Regulatory Commission {NRC). These evacuation time estimates are approximately 60% and 84% greater, respectively, than the estimate provided by PSNH for a simulataneous evacuation of the entire EPZ under peak conditions. The CLEAR model, which was developed by Pacific Northwest Laboratory {PNL) under the sponsorship of the. NRC, was also used for these latter calculations. The results of this study reveal the importance of the assumptions used for calculating evacuation times. Because traffic routings and management plans have not been prepared for the area, the CLEAR calculations utilized indepdently prepared traffic routings and assumptions. A detailed analysis of the results suggests that the ETEs submitted by PSNH are consistent with the methods and assumptions which provide the bases for PSNH•s evacuation time estimates. Differences among evacuation time estimates stem largely from differences jn the assumed size of the evacuating population and the estimated effectiveness of traffic controls.
Date: November 1, 1982
Creator: Moeller, M. P.; Urbanik, II, T.; Mclean, M. A. & Desrosiers, A. E.
Partner: UNT Libraries Government Documents Department

An Analysis of Evacuation Time Estimates Around 52 Nuclear Power Plant Sites Analysis and Evaluation

Description: On November 29, 1979, the NRC sent a letter to 52 nuclear power plants requesting evacuation time estimates for 10 sectors within a 10-mile radius of each plant. The requirements for these evacuation times are contained in NUREG-0654, Rev. 1, and include such factors as population density, weather conditions, warning time, response time and confirmation time. Fifty responses were received. The analysis of these findings are presented for review.
Date: May 1, 1981
Creator: Urbanik, II, T. & Desrosiers, A. E.
Partner: UNT Libraries Government Documents Department


Description: In December 1983, 13 nuclear utilities that own TDI diesel generators formally established an Owners• Group to address concerns regarding the reliability and operability of these engines. The Owners' Group program for engine requalification consisted of four major elements: 1) resolution of known problems with potentially generic implications, 2) a design review and quality revalidation (DR/QR) effort aimed at identifying and correcting potential problems with the important engine components, 3) expanded engine testing and inspection, and 4) enhanced engine maintenance and surveillance (M/S) to maintain the qualification of the diesel engines for the lifetime of the nuclear plants that they service. In providing technical support to NRC, the PNL project staff, assisted by a number of diesel engine consultants, focused on the four major elements of the Owners' Group engine requalification program, addressing both generic and plant-specific areas.
Date: December 1, 1985
Creator: Berlinger, C. H.
Partner: UNT Libraries Government Documents Department

Power Histories for Fuel Codes

Description: Computations of power history effects on the pre-loss-of-coolant accident (LOCA) conditions of generic pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rods were performed at Pacific Northwest Laboratory using the U.S. Nuclear Regulatory Commission (NRC) code FRAPCON-2. Comparisons were made between cases where the fuel operated at a high ( 11 LOCA-limited") power throughout life (20,000 MWd/MTU) and those where the fuel was at a lower power for most of its burnup and ramped to the high power at 10,000 or 20,000 MWd/MTU burnup. The PWR rod was calculated to have more cladding creepdown during the lower power cases, which resulted in slightly lower centerline temperatures (as much as 100{degrees}C). This result was insensitive to the method used to increase the power during the ramps (i.e., by increasing the average rod power or by changing the peak-to-average (P/A} ratio of the axial power shape). The calculations also indicate that the highest fuel centerline temperatures were reached at startup. The BWR rod, however, demonstrated a substantial dependence on the power history. In this case, the constant high-power rod released considerably more fission gas than the lower power cases (21% versus 0.4%), which resulted in temperature differences of up to 350°C. The hiqhest temperature was reached at end-of-life (EOL) in the constant high-power case.
Date: January 1, 1982
Creator: Gilbert, E. R.; Rausch, W. N. & Panisko, F. E.
Partner: UNT Libraries Government Documents Department

Survey of Remote Area Monitoring Systems at U.S. Light-Water-Cooled Power Reactors

Description: A study was made of the capabilities and operating practices, including calibration, of remote area monitoring (RAM) systems at light-water-cooled power reactors in the United States. The information was obtained by mail questionaire. Specific design capabilities, including range, readout and alarm features are documented along with the numbers and location of detectors, calibration and operational procedures. Comments of respondents regarding RAM systems are also included.
Date: April 1, 1982
Creator: Kathren, R. L. & Mileham, A. P.
Partner: UNT Libraries Government Documents Department

Fuel Performance Annual Report for 1980

Description: This annual report, the third in a series, provides a brief description of fuel performance in conmercial nuclear power plants. Brief summaries of fuel surveillance programs and operating experience, fuel performance problems, and fuel design changes are provided. References to additional, more detailed, information and related NRC evaluation are included.
Date: December 1, 1981
Creator: Bailey, W. J.; Rising, K. H. & Tokar, M.
Partner: UNT Libraries Government Documents Department

Fuel Performance Annual Report for 1979

Description: This annual report, the second in a series, provides a brief description of fuel performance in commercial nuclear power plants. Brief summaries are given of fuel surveillance programs, fuel performance problems, and fuel design changes. References to additional, more detailed, information and related NRC evaluation are provided.
Date: January 1, 1981
Creator: Tokar, M.; Mailey, W. J. & Cunningham, M. E.
Partner: UNT Libraries Government Documents Department

Use of a Moving Heat Conductor Mesh to Perform Reflood Calculations with RELAP4/MOD6

Description: RELAP4 is a computer code which can be used for the transient therm~l hydraulic analysis of light water reactors and related systems .. Various versions of the RELAP4 code are widely used throughout the world for experimental system analysis, reactor design,and nuclear system safety studies. RELAP4/MOD6 includes many new analytical models which were developed primarily for the analysis of the reflood phase of a PWR loss-of-coolant accident (LOCA) transient. The key feature forming the basis for the MOD6 reflood calculation is a unique moving finite differenced heat conductor. This paper will describe the development and application of the moving heat conductor mesh for use in reflood analysis.
Date: May 6, 1979
Creator: Fischer, S R; Ellis, L V & Chen, Y S
Partner: UNT Libraries Government Documents Department

Pre-Phase 1 Aging Assessment of the BWR and PWR Accumulators

Description: Accumulators are important components used in many systems at commercial boiling water reactors (BWRs) and pressurized water reactors in the United States. The accumulators are vessels attached to fluid systems to provide 1) a limited backup source of stored fluid energy for hydraulic/pneumatic mechanical equipment, 2) a damping effect on pressure pulses in fluid systems, and 3) a volume of fluid to be injected passively into a fluid system. Accumulators contain a gas that is compressed or expanded as the fluid from the system enters or exits the accumulator. The gas and fluid in accumulators are usually separated from each other by a piston or bladder. In support of the U.S. Nuclear Regulatory Commission's Nuclear Aging Research Program (NPAR), the Pacific Northwest Laboratory conducted an analysis of available industry databases to determine if accumulator components already had been studied in other NPAR assessments and to evaluate each accumulator type for applicable aging issues. The results of this preliminary study indicate that two critical uses of accumulators have been previously evaluated by the NPAR program. NUREGICR-5699, Aging and Service Wear of Control Rod Drive Mechanisms for BUT Nuclear Plants (Greene 199 I), identified two hydraulic control unit components subject to aging failures: accumulator nitrogen-charging cartridge valves and the scram water accumulator. In addition, NUREGICR-6001, Aging Assessment of BWR Standby Liquid Control Systems (Buckley et al. 1992), identified two predominant aging-related accumulator failures that result in a loss of the nitrogen blanket pressure: (charging) valve wear and failure of the gas bladder. The present study has identified five prevalent aging-related accumulator failures: rupture of the accumulator bladder separation of the metal disc from the bottom of the bladder leakage of the gas from the charging valve leakage past the safety injection tank manway cover gasket leakage past O-rings. An additional study of ...
Date: August 1, 1995
Creator: Buckely, G. D.
Partner: UNT Libraries Government Documents Department

Pre-Phase 1 Aging Assessment of the BWR Isolation Condenser System

Description: The isolation condenser system (ICS) is part of the emergency core cooling system in five U.S. boiling-water reactors. In the event that the reactor pressure vessel becomes isolated from the main condenser, the ICS removes decay heat from the reactor. The ICS is important to reactor safety because it is relied on to help mitigate core damage during a loss-of-coolant accident. In support of the U.S. Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, staff from the Pacific Northwest Laboratory researched the aging of the ICS by reviewing available industry databases. Each component of the ICS was evaluated to 1) identify applicable aging issues and also to 2) determine if the component had already been studied as a part of other NPAR assessments. The results of this preliminary study indicate that most of the critical ICS components have been previously evaluated by the NPAR program. The one ICS component that has not been specifically studied is the isolation condenser itself. There is little evidence in the databases to suggest that there have been problems with the isolation condenser. Only one plant, Millstone Unit 1, has ever had an isolation condenser tube failure problem recorded. This instance resulted from events that occurred early in the life of the plant. The problem was remedied through tube replacement. The isolation condenser and the pressurized-water reactor (PWR) steam generator were compared to illustrate that even though the isolation condenser is a heat exchanger, it is not subjected to the same service dynamics as the PWR steam generator. The isolation condenser operates for most of its service life in a relatively benign, static environment, resulting in a comparatively good service record. PNL staff recommend that the results of this research be used to continue studying the ICS to determine if the aging isolation condenser ...
Date: August 1, 1995
Creator: Orton, R. D.
Partner: UNT Libraries Government Documents Department

Efficiency Studies with Gamma Ray Portion of Specialized Reactor-Shield Monte Carlo Program 18-0

Description: Application studies were made with Specialized Reactor-Shield Monte Carlo Program 18-0 to determine the efficiency and feasibility of calculating energy deposition due to primary core gamma rays throughout the XNJ140E-1 reactor-shield assembly. Monte Carlo results are presented in tabular form for all geometrical regions used to describe the shield. Described here is a means of obtaining adequate and valid heating rates in about 47 hours on the IBM-704 digital computer. Comparison of Monte Carlo and point kernel data are included.
Date: August 1, 1961
Creator: Capo, M. A.
Partner: UNT Libraries Government Documents Department

Mechanistic Considerations Used in the Development of the PROFIT PCI Failure Model

Description: A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) Interactions (PC!) failure model for estimating the probability of failure in !ransient increases in power (PROFIT) was developed. PROFIT is based on 1) standard statistical methods applied to available PC! fuel failure data and 2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmental and strain-rate dependent strain energy absorption to failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-disloction interaction effects in the Zircaloy cladding. Assuming that the power ramping rate is the operating corollary of strain-rate in the Zircaloy cladding, then the variables of first order importance in the PCI fuel failure phenomenon are postulated to be: 1. pre-transient fuel rod power, P{sub I}, 2. transient increase in fuel rod power, {Delta}P, 3. fuel burnup, Bu, and 4. the constitutive material property of the Zircaloy cladding, SEAF.
Date: May 1, 1980
Creator: Pankaskie, P. J.
Partner: UNT Libraries Government Documents Department

Study of Air Ingress Across the Duct During the Accident Conditions

Description: The goal of this project is to study the fundamental physical phenoena associated with air ingress in very high temperature reactors (VHTRs). Air ingress may occur due to a nupture of primary piping and a subsequent breach in the primary pressure boundary in helium-cooled and graphite-moderated VHTRs. Significant air ingress is a concern because it introduces potential to expose the fuel, graphite support rods, and core to a risk of severe graphite oxidation. Two of the most probable air ingress scenarios involve rupture of a control rod or fuel access standpipe, and rupture in the main coolant pipe on the lower part of the reactor pressure vessel. Therefor, establishing a fundamental understanding of air ingress phenomena is critical in order to rationally evaluate safety of existing VHTRs and develop new designs that mimimize these risks. But despite this importance, progress toward development these predictive capabilities has been slowed by the complex nature of the underlaying phenomena. The combination of interdiffusion among multiple species, molecular diffusion, natural convection, and complex geometries, as well as the multiple chemical reactions involved, impose significant roadblocks to both modeling and experiment design. The project team will employ a coordinated experimental and computational effort that will help gain a deeper understanding of multiphased air ingress phenomena. THis project will enhance advanced modeling and simulation methods, enabling calculation of nuclear power plant transients and accident scenarios with a high degree of confidence. The following are the project tasks: Perform particle image velocimetry measurement of multiphase air ingresses Perform computational fluid dynamics analysis of air ingress phenomena
Date: May 6, 2013
Creator: Hassan, Yassin
Partner: UNT Libraries Government Documents Department

Radiological Assessment of Steam Generator Removal and Replacement: Update and Revision

Description: A previous analysis of the radiological impact of removing and replacing corroded steam generators has been updated based on experience gained during steam generator repairs at Surry Unit 2. Some estimates of occupational doses involved in the operation have been revised but are not significantly different from the earlier estimates. Estimates of occupational doses and radioactive effluents for new tasks have been added. Health physics concerns that arose at Surry included the number of persons involved in the operation, tne training of workers, the handling of quantitites.of low-level waste, and the application of the ALARA principle. A review of these problem areas may help in the planning of other similar operations. A variety of processes could be used to decontaminate steam generators. Research is needed to assess these techniques and their associated occupational doses and waste volumes. Contaminated steam generators can be stored or disposed of after removal without significant radiological problems. Onsite storage and intact shipment have the least impact. In-placing retubing, an alternative to steam generator removal, results in occupational doses and effluents similar to those from removal, but prior decontamination of the channel head is needed. The retubing option should be assessed further.
Date: December 1, 1980
Creator: Hoenes, G. R.; Mueller, M. A. & McCormack, W. D.
Partner: UNT Libraries Government Documents Department

Technical Letter Report: Evaluation and Analysis of a Few International Periodic Safety Review Summary Reports

Description: At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commission (NRC) incorporate lessons learned from periodic safety reviews (PSRs) performed in other countries as an input to the NRC’s assessment processes. In the U.S., commercial nuclear power plants (NPPs) are granted an initial 40-year operating license, which may be renewed for additional 20-year periods, subject to complying with regulatory requirements. The NRC has established a framework through its inspection, and operational experience processes to ensure the safe operation of licensed nuclear facilities on an ongoing basis. In contrast, most other countries do not impose a specific time limit on the operating licenses for NPPs, they instead require that the utility operating the plant perform PSRs, typically at approximately 10-year intervals, to assure continued safe operation until the next assessment. The staff contracted with Argonne National Laboratory (Argonne) to perform a pilot review of selected translated PSR assessment reports and related documentation from foreign nuclear regulatory authorities to identify any potential new regulatory insights regarding license renewal-related topics and NPP operating experience (OpE). A total of 14 PSR assessment documents from 9 countries were reviewed. For all of the countries except France, individual reports were provided for each of the plants reviewed. In the case of France, three reports were provided that reviewed the performance assessment of thirty-four 900-MWe reactors of similar design commissioned between 1978 and 1988. All of the reports reviewed were ...
Date: December 17, 2013
Creator: Chopra, Omesh K.; Diercks, Dwight R.; Ma, David Chia-Chiun & Garud, Yogendra S.
Partner: UNT Libraries Government Documents Department


Description: A preliminary analysis has been made to compare the performance.of a ceramic ramjet reactor powerplant, such as described in General Electric Report No. XDC-56-5-81, with that of such a powerplant supercharged by a metallic-vapor-cycle compressor jet. Performance at sea level, Mach 2.5 is ·reported for the vapor-cycle compressor jet alone, for the ramjet alone, and for the compressor jet - ramjet combination. Results indicate that adding the compressor-jet as a supercharger for the ramjet provides an increase in specific thrust of about 20 percent over that of the ramjet alone, with an attendant increase in thermal efficiency of about 20 percent over that of the ramjet alone.
Date: May 17, 1957
Creator: Boppart, J. A.
Partner: UNT Libraries Government Documents Department


Description: Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.
Date: April 1, 1981
Creator: Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N. et al.
Partner: UNT Libraries Government Documents Department

Application of Linear Propagation of Errors to Fuel Rod Temperature and Stored Energy Calculations

Description: Linear propagatlon of errors evaluates modeling uncertainty by approximating a function of interest by first-order Taylor's series expansions and then approximating the variance of the function by the variance of the linear approximation. This report discusses uncertainty analysis for different nuclear fuel rod designs, the process of model validation, and the effect of cracked pellet fuel models upon temperabre uncertainty. Using a postulated power history, the uncertainty for the predicted thermal response of boiling water reactor (BWR) and pressurized water reactor (PWR} fuel rods was evaluated. Beginning-of-life (BOL) relative uncertainty for BWR and PWR fuel rods is approximately the same. while different end-of-fife {EOL} thermal response results in different EOL uncertainty. Determining the validity of modeling relative to reality is discussed in qualitative terms. Validity is dependent upon verifying that the code correctly implements the model and that satisfactory agreement is found between the model and measurements. Fuel modeling codes are now using cracked pellet fuel models, which result in decreased fuel surface temperature. Estimated stored energy is lowered; but its relative uncertainty is increased. In general, however, the absolute upper uncertainty bound for stored energy is lower for a cracked pellet model than for a solid pellet model.
Date: October 1, 1980
Creator: Cunningham, M. E.; Olsen, A. R.; Lanning, D. D. & Willford, R. E.
Partner: UNT Libraries Government Documents Department


Description: This report presents information to participants in the Team Interaction Skills study conducted at Diablo Canyon Power Plant from September to November 1989. A study was conducted to develop and assess measures of team interaction skills of nuclear power plant control room crews in simulated emergency conditions. Data were collected at a boiling water reactor (BWR) and pressurized water reactor (PWA) using three sets of rating scales; Behaviorally Anchored Rating Scales (BARS), Behavioral Frequency rating scales, and Technical Performance rating scales. Diablo Canyon Power Plant agreed to serve as the PWR plant in the study. Obse!Vers consisting of contract license examiners, Diablo Canyon Power Plant training instructors, and project staff used the rating scales to provide assessments of team interaction skills and technical skills of control room crews during emerg-3ncy scenarios as part of license requalification training. Crew members were also asked to providH self-ratings of their performance to gather information regarding crew responses to the Team Interactions Skills rating scales.
Date: December 1, 1990
Creator: Hauth, J. T.; Toquam, J. L.; Bramwell, A. T. & Fleming, T. E.
Partner: UNT Libraries Government Documents Department