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Description: For the licensing of the current fleet of commercial nuclear power plants (NPPs), the Nuclear Regulatory Commission (NRC) used two key documents, NUREG-0800 and Regulatory Guide (RG) 1.70. RG 1.70 provided guidance to applicants on the contents needed in their Safety Analysis Reports (SARs) submitted as part of their application to construct or operate an NPP. NUREG-0800, the NRC Standard Review Plan (SRP), provides guidance to the NRR staff reviewers on performing their safety reviews of these applications. As part of the preparation for a new wave of improved NPP designs the NRC is in the process of updating the SRP and is also developing a new RG designated as draft RG or DG-1145, ''Combined License Applications for Nuclear Power Plants (LWR Edition).'' This will eventually become RG 1.206 and will take the place of RG 1.70. This will provide guidance for combined license (COL) applicants, as well as for other 10CFR Part 52 variations that are permitted.
Date: March 26, 2007
Partner: UNT Libraries Government Documents Department


Description: This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.
Date: September 26, 2008
Creator: Vierow, Karen
Partner: UNT Libraries Government Documents Department

AFCI-2.0 Library of Neutron Cross Section Covariances

Description: Neutron cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The primary purpose of the library is to provide covariances for the Advanced Fuel Cycle Initiative (AFCI) data adjustment project, which is focusing on the needs of fast advanced burner reactors. The covariances refer to central values given in the 2006 release of the U.S. neutron evaluated library ENDF/B-VII. The preliminary version (AFCI-2.0beta) has been completed in October 2010 and made available to the users for comments. In the final 2.0 release, covariances for a few materials were updated, in particular new LANL evaluations for {sup 238,240}Pu and {sup 241}Am were adopted. BNL was responsible for covariances for structural materials and fission products, management of the library and coordination of the work, while LANL was in charge of covariances for light nuclei and for actinides.
Date: June 26, 2011
Creator: Herman, M.; Herman,M.; Oblozinsky,P.; Mattoon,C.; Pigni,M.; Hoblit,S. et al.
Partner: UNT Libraries Government Documents Department

Review of the IEEE Standard for Computerized Operating Procedure Systems

Description: Increasingly nuclear power plant procedures, such as emergency operating procedures, are being presented in computer form with functionality to support operator use and management of the procedures. The U.S. Nuclear Regulatory Commission (NRC) currently has guidance for the review of computer-based procedures (CBPs); however, there remain CBP functions and human performance issues for which up-to-date guidance is lacking. The Institute of Electrical and Electronics Engineers (IEEE) has initiated a standard development effort to address the human factors engineering (HFE) aspects of CBP systems. When completed, it may provide guidance to supplement the NRC staff's review criteria. The purpose of our study was to evaluate the suitability of the IEEE Standard for use in the NRC's HFE safety reviews of CBP systems and to ensure that the guidance meets the NRC's standard for scientific and engineering rigor used in its own guidance development efforts. We established the following criteria with which to evaluate the Standard: (1) it should meet an existing need of NRC reviewers, (2) it should be based in sound HFE principles, (3) it should be thoroughly peer-reviewed, and (4) it should address CBP-related human performance issues identified in the literature. This report describes the methodology we used to evaluate each criterion. Our evaluation concluded that the Standard generally does meet these criteria, however several areas were identified for which additional clarifications are needed. Thus consideration of the Standard's use by the NRC is supported. The standard evaluation methodology developed in this study can be generally applied to the review of other HFE standards being considered for possible use or endorsement by the NRC.
Date: February 26, 2010
Creator: O'Hara, J. & Higgins, J.
Partner: UNT Libraries Government Documents Department

Technical Letter Report - Analysis of Ultrasonic Data on Piping Cracks at Ignalina Nuclear Power Plant Before and After Applying a Mechanical Stress Improvement Process, JCN-N6319, Task 2

Description: Pacific Northwest National Laboratory (PNNL) is assisting the United States Nuclear Regulatory Commission (NRC) in developing a position on the management of primary water stress corrosion cracking (PWSCC) in piping systems previously analyzed for leak-before-break (LBB). Part of this work involves determining whether inspections alone are sufficient or if inspections plus mitigation techniques are needed. The work described in this report addresses the reliability of ultrasonic phased-array (PA) examinations for inspection of cracks that have been subjected to the mitigation method of mechanical stress improvement process (MSIP). It is believed that stresses imparted during MSIP may make ultrasonic crack responses in piping welds more difficult to detect and accurately characterize. To explore this issue, data were acquired, both before and after applying MSIP, and analyzed from cracked areas in piping at the Ignalina Nuclear Power Plant (INPP) in Lithuania. This work was performed under NRC Project JCN-N6319, PWSCC in Leak-Before-Break Systems.
Date: February 26, 2008
Creator: Anderson, Michael T.; Cumblidge, Stephen E. & Crawford, Susan L.
Partner: UNT Libraries Government Documents Department


Date: June 26, 2006
Partner: UNT Libraries Government Documents Department

Technetium Waste Form Development Progress Report

Description: The approach being followed to evaluate the use of an iron-based alloy waste form to immobilize the Tc-bearing waste streams generated during the aqueous and electrochemical processing of used fuel that is being studied in the DOE Advanced Fuel Cycle Initiative (AFCI) is presented in this report. The objective is to develop an alloy waste form that provides high waste loading within waste form processing limitations, meets waste form performance requirements for durability and the long-term retention of radionuclides, and can be produced with consistent physical, chemical, and radiological properties that meet regulatory acceptance requirements for disposal. Microanalysis using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) was used to analyze non-radioactive Fe-Mo-Re samples. A sample was prepared for SEM; however, significant unforeseen instrument problems led to delays in conducting the detailed work. The TEM was not available for this particular sample and therefore only preliminary SEM work can be reported. The results are in agreement with previous studies [Ebert 2009]; however, a rhenium-rich region within the Re-Mo phase is clearly visible.
Date: February 26, 2010
Creator: Buck, Edgar C.
Partner: UNT Libraries Government Documents Department

Control of Urania Crystallite Size by HMTA-Urea Reactions in the Internal Gelation Process for Preparing (U, Pu)O<sub>2</sub>Fuel Kernels

Description: In the development of (U,Pu)O{sub 2} kernels by the internal gelation process for the Direct Press Spheroidized process at Oak Ridge National Laboratory, a novel crystal growth step was discovered that made it possible to prepare calcined porous kernels that could be used as direct-press feed for Fast Breeder Reactor pellet fabrication. High-quality pellets were prepared that were near theoretical density and that (upon examination) revealed no evidence of sphere remnants. The controlled crystal growth step involved using hexamethylenetetramine (HMTA)-urea stock solutions that were boiled for 60 min or less. Before this discovery, all the other crystal growth steps (when utilized) could reduce the tap density to only {approx}1.3 g/cm{sup 3}, which was not sufficiently low for use in ideal pellet pressing. The use of the boiled HMTA-urea solution allowed the tap density to be lowered to 0.93 g/cm{sup 3}, with the ideal density being about 1.0 g/cm{sup 3}. This report describes the development of this technology and its scaleup.
Date: April 26, 2005
Creator: Collins, J.L.
Partner: UNT Libraries Government Documents Department

Serpentine: Finite Difference Methods for Wave Propagation in Second Order Formulation

Description: Wave propagation phenomena are important in many DOE applications such as nuclear explosion monitoring, geophysical exploration, estimating ground motion hazards and damage due to earthquakes, non-destructive testing, underground facilities detection, and acoustic noise propagation. There are also future applications that would benefit from simulating wave propagation, such as geothermal energy applications and monitoring sites for carbon storage via seismic reflection techniques. In acoustics and seismology, it is of great interest to increase the frequency bandwidth in simulations. In seismic exploration, greater frequency resolution enables shorter wave lengths to be included in the simulations, allowing for better resolution in the seismic imaging. In nuclear explosion monitoring, higher frequency seismic waves are essential for accurate discrimination between explosions and earthquakes. When simulating earthquake induced motion of large structures, such as nuclear power plants or dams, increased frequency resolution is essential for realistic damage predictions. Another example is simulations of micro-seismic activity near geothermal energy plants. Here, hydro-fracturing induces many small earthquakes and the time scale of each event is proportional to the square root of the moment magnitude. As a result, the motion is dominated by higher frequencies for smaller seismic events. The above wave propagation problems are all governed by systems of hyperbolic partial differential equations in second order differential form, i.e., they contain second order partial derivatives of the dependent variables. Our general research theme in this project has been to develop numerical methods that directly discretize the wave equations in second order differential form. The obvious advantage of working with hyperbolic systems in second order differential form, as opposed to rewriting them as first order hyperbolic systems, is that the number of differential equations in the second order system is significantly smaller. Another issue with re-writing a second order system into first order form is that compatibility conditions often ...
Date: March 26, 2012
Creator: Petersson, N A & Sjogreen, B
Partner: UNT Libraries Government Documents Department

Failure Forewarning in NPP Equipment NERI2000-109 Final Project Report

Description: The objective of this project is forewarning of machine failures in critical equipment at next-generation nuclear power plants (NPP). Test data were provided by two collaborating institutions: Duke Engineering and Services (first project year), and the Pennsylvania State University (Applied Research Laboratory) during the second and third project years. New nonlinear methods were developed and applied successfully to extract forewarning trends from process-indicative, time-serial data for timely, condition-based maintenance. Anticipation of failures in critical equipment at next-generation NPP will improve the scheduling of maintenance activities to minimize safety concerns, unscheduled non-productive downtime, and collateral damage due to unexpected failures. This approach provides significant economic benefit, and is expected to improve public acceptance of nuclear power. The approach is a multi-tiered, model-independent, and data-driven analysis that uses ORNL's novel nonlinear method to extract forewarning of machine failures from appropriate data. The first tier of the analysis provides a robust choice for the process-indicative data. The second tier rejects data of inadequate quality. The third tier removes signal artifacts that would otherwise confound the analysis, while retaining the relevant nonlinear dynamics. The fourth tier converts the artifact-filtered time-serial data into a geometric representation, that is then transformed to a discrete distribution function (DF). This method allows for noisy, finite-length datasets. The fifth tier obtains dissimilarity measures (DM) between the nominal-state DF and subsequent test-state DFs. Forewarning of a machine failure is indicated by several successive occurrences of the DM above a threshold, or by a statistically significant trend in the DM. This paradigm yields robust nonlinear signatures of degradation and its progression, allowing earlier and more accurate detection of the machine failure.
Date: March 26, 2004
Creator: Hively, LM
Partner: UNT Libraries Government Documents Department

Interim progress report on safety and licensing strategy support for the ABR prototype.

Description: Argonne National Laboratory is providing support to the U.S. Department of Energy in the Global Nuclear Energy Partnership (GNEP) in certification of an advanced, sodium-cooled fast reactor. The reactor is to be constructed as a prototype for future commercial power reactors that will produce electricity while consuming actinides recovered from light water reactor spent fuel. This prototype reactor has been called the ABR, or Advanced Burner Reactor.
Date: June 26, 2007
Creator: Cahalan, J .E.
Partner: UNT Libraries Government Documents Department

Hydrodynamic design report intermediate size inducer pump for sodium operation

Description: An inducer pump has been designed to fit within the existing housing of the FFTF Facility. The design has been shown to meet the specified requirements in every area where an analytical prediction of the performance can be made. The head-rise and efficiency of the pump have design margins so that the calculated values actually exceed requirements over the full range of operation. The suction performance of the design provides a very large margin at the operating NPSH value which is the major advantage of using the inducer pump. The suction performance margin at off-design is estimated to be adequate for suction performance, but may not be as large as desired for life considerations. The design has incorporated features that provide for long life at the design point. These features have been established based on proven performance of commercial waterjet pumps designed by Rocketdyne.
Date: April 26, 1978
Creator: King, J.A.; O'Hern, F.C. & Jackson, E.D.
Partner: UNT Libraries Government Documents Department

A Study on the Tritium Behavior in the Rice Plant after a Short-Term Exposure of HTO

Description: In many Asian countries including Korea, rice is a very important food crop. Its grain is consumed by humans and its straw is used to feed animals. In Korea, there are four CANDU type reactors that release relatively large amounts of tritium into the environment. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out the experimental studies to obtain domestic data on various parameters concerning the direct contamination of plant. In this study, the behavior of tritium in the rice plant is predicted and compared with the measurement performed at KAERI. Using the conceptual model of the soil-plant-atmosphere tritiated water transport system which was suggested by Charles E. Murphy, tritium concentrations in the soil and in leaves to time were derived. If the effect of tritium concentration in the soil is considered, the tritium concentration in leaves is described as a double exponential model. On the other hand if the tritium concentration in the soil is disregarded, the tritium concentration in leaves is described by a single exponential term as other models (e.g. Belot's or STAR-H3 model). Also concentration of organically bound tritium in the seed is predicted and compared with measurements. The results can be used to predict the tritium concentration in the rice plant at a field around the site and the ingestion dose following the release of tritium to the environment.
Date: February 26, 2002
Creator: Yook, D.-S.; Lee, K. J. & Choi, Y-H.
Partner: UNT Libraries Government Documents Department

Estimating the Uncertainty in Reactivity Accident Neutronic Calculations

Description: A study of the uncertainty in calculations of the rod ejection accident in a pressurized water reactor is being carried out for the US Nuclear Regulatory Commission. This paper is a progress report on that study. Results are presented for the sensitivity of core energy deposition to the key parameters: ejected rod worth, delayed neutron fraction, Doppler reactivity coefficient, and fuel specific heat. These results can be used in the future to estimate the uncertainty in local fuel enthalpy given some assumptions about the uncertainty in the key parameters. This study is also concerned with the effect of the intra-assembly representation in calculations. The issue is the error that might be present if assembly-average power is calculated, and pin peaking factors from a static calculation are then used to determine local fuel enthalpy. This is being studied with the help of a collaborative effort with Russian and French analysts who are using codes with different intra-assembly representations. The US code being used is PARCS which calculates power on an assembly-average basis. The Russian code being used is BARS which calculates power for individual fuel pins using a heterogeneous representation based on a Green's Function method.
Date: October 26, 1998
Creator: Diamond, D. J.; Yang, C. Y. & Aronson, A. L.
Partner: UNT Libraries Government Documents Department

NRC Support for the Kalinin (Vver) Probabilistic Risk Assessment

Description: The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA.
Date: October 26, 1998
Creator: Bley, D.; Diamond, D. J.; Chu, T. L.; Azarm, A.; Pratt, W. T.; Johnson, D. et al.
Partner: UNT Libraries Government Documents Department

Considerations for advanced reactor design based on EBR-II experience.

Description: The long-term success of the Experimental Breeder Reactor-II (EBR-II) provides several insights into fundamental characteristics and design features of a nuclear generating station that enhance safety, operability, and maintainability. Some of these same characteristics, together with other features, offer the potential for operational lifetimes well beyond the current licensing time frame, and improved reliability that could potentially reduce amortized capital costs as well as overall operation and maintenance costs if incorporated into advanced plant designs. These features and characteristics are described and the associated benefits are discussed.
Date: May 26, 1999
Creator: King, R. W.
Partner: UNT Libraries Government Documents Department

A theoretical and numerical investigation of turbulent steam jets in BWR steam blowdown.

Description: The preliminary results of PHOENICS and RELAP5 show that the current numerical models are adequate in predicting steam flow and stratification patterns in the upper Drywell of a BWR containment subsequent to a blow-down event. However, additional modeling is required in order to study detailed local phenomena such as condensation with non-condensables, natural convection, and stratification effects. Analytically, the intermittence modified similarity solutions show great promise. Once {gamma} is accounted for, the jet's turbulent shear stress can be determined with excellent accuracy.
Date: June 26, 1998
Creator: NguyenLe, Q.
Partner: UNT Libraries Government Documents Department

Anomalies in Proposed Regulations for the Release of Redundant Material from Nuclear and Non-nuclear Industries

Description: Now that increasing numbers of nuclear power stations are reaching the end of their commercially useful lives, the management of the large quantities of very low level radioactive material that arises during their decommissioning has become a major subject of discussion, with very significant economic implications. Much of this material can, in an environmentally advantageous manner, be recycled for reuse without radiological restrictions. Much larger quantities--2-3 orders of magnitude larger--of material, radiologically similar to the candidate material for recycling from the nuclear industry, arise in non-nuclear industries like coal, fertilizer, oil and gas, mining, etc. In such industries, naturally occurring radioactivity is artificially concentrated in products, by-products or waste to form TENORM (Technologically Enhanced Naturally Occurring Radioactive Material). It is only in the last decade that the international community has become aware of the prevalence of T ENORM, specially the activity levels and quantities arising in so many nonnuclear industries. The first reaction of international organizations seems to have been to propose ''double'' standards for the nuclear and non-nuclear industries, with very stringent release criteria for radioactive material from the regulated nuclear industry and up to a hundred times more liberal criteria for the release/exemption of TENORM from the as yet unregulated non-nuclear industries. There are, however, many significant strategic issues that need to be discussed and resolved. An interesting development, for both the nuclear and non-nuclear industries, is the increased scientific scrutiny that the populations of naturally high background dose level areas of the world are being subject to. Preliminary biological studies have indicated that the inhabitants of such areas, exposed to many times the permitted occupational doses for nuclear workers, have not shown any differences in cancer mortality, life expectancy, chromosome aberrations or immune function, in comparison with those living in normal background areas. The paper discusses these ...
Date: February 26, 2002
Creator: Menon, S.
Partner: UNT Libraries Government Documents Department

Immobilization of Fast Reactor First Cycle Raffinate

Description: This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains &gt; 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.
Date: February 26, 2003
Creator: Langley, K. F.; Partridge, B. A. & Wise, M.
Partner: UNT Libraries Government Documents Department

Decommissioning Project of Bohunice A1 NPP

Description: The first (pilot) nuclear power plant A1 in the Slovak Republic, situated on Jaslovske Bohunice site (60 km from Bratislava) with the capacity of 143 MWel, was commissioned in 1972 and was running with interruptions till 1977. A KS 150 reactor (HWGCR) with natural uranium as fuel, D2O as moderator and gaseous CO2 as coolant was installed in the A1 plant. Outlet steam from primary reactor coolant system with the temperature of 410 C was led to 6 modules of steam generators and from there to turbine generators. Refueling was carried out on-line at plant full power. The first serious incident associated with refueling occurred in 1976 when a locking mechanism at a fuel assembly failed. The core was not damaged during that incident and following a reconstruction of the damaged technology channel, the plant continued in operation. However, serious problems were occurring with the integrity of steam generators (CO2 gas on primary side, water and steam on secondary side) when the plant had to be shut down frequently due to failures and subsequent repairs. The second serious accident occurred in 1977 when a fuel assembly was overheated with a subsequent release of D2O into gas cooling circuit due to a human failure in the course of replacement of a fuel assembly. Subsequent rapid increase in humidity of the primary system resulted in damages of fuel elements in the core and the primary system was contaminated by fission products. In-reactor structures had been damaged, too. Activity had penetrated also into certain parts of the secondary system via leaking steam generators. Radiation situation in the course of both events on the plant site and around it had been below the level of limits specified. Based on a technical and economical justification of the demanding character of equipment repairs for the restoration ...
Date: February 26, 2002
Creator: Stubna, M.; Pekar, A.; Moravek, J. & Spirko, M.
Partner: UNT Libraries Government Documents Department

In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

Description: The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and ...
Date: February 26, 2002
Creator: Haghighi, M. H.; Kring, C. T.; McGehee, J. T.; Jugan, M. R.; Chapman, J. & Meyer, K. E.
Partner: UNT Libraries Government Documents Department