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Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

Description: The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.
Date: August 1, 1995
Partner: UNT Libraries Government Documents Department

Investigation of Burnup Credit Issues in BWR Fuel

Description: Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel.
Date: September 20, 1999
Creator: Broadhead, B.L. & DeHart, M.D.
Partner: UNT Libraries Government Documents Department

Shield for Water Boiler

Description: Siimplified shielding calculations indicating the proposed design for the water boiler assembly will reduce the radiation at normal operaton to values well below those which are considered tolerable.
Date: August 8, 1951
Creator: Balent, R.
Partner: UNT Libraries Government Documents Department

Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments

Description: The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments [Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a basis for ...
Date: January 1, 1994
Creator: Ott, L. J.
Partner: UNT Libraries Government Documents Department

Interpretation of the XR2-1 experiment and characteristics of the BWR lower plenum debris bed

Description: The Ex-Reactor (XR) experiments have been conducted to advance the understanding of BWR severe accident melt progression events. The XR2-1 experiment addresses the fate of the initial large (code-predicted) movements of molten metals from the upper core to the lower core and core plate region. For this question, which has ramifications for blockage formation in the core region, the XR2-1 test results provide significant and perhaps definitive insights. Nevertheless, some events that occurred during this test are creatures of the special features of the test apparatus, and there is a potential for misconceptions with respect to the direct applicability of some of the results. This paper describes the conclusions that can be drawn from the XR2-1 experiment results and identifies those areas (such as fuel pellet stack collapse and core plate integrity) where care must be taken not to misconstrue the test events. Another important area where much recent work has been performed is the effort to analyze the potential for maintaining core debris within the reactor vessel lower plenum by cooling of the outer vessel wall. One of the first steps in such an analytical endeavor is to attempt to establish the pattern of energy transfers into the wall inner surface. As a prerequisite to determination of this pattern, it is necessary to first consider the nature of the debris within the lower plenum. Too often is an easily represented homogeneous circulating liquid pool incorporated without adequate consideration of the true material conditions. Basic considerations of the relative quantities of materials present, the potentials for eutectics formations, and the associated melting points dictate otherwise. This paper offers some insights as to the true nature of the lower plenum debris and discusses the need for some relatively simple experiments that would contribute much toward the basic understanding necessary for accurate ...
Date: November 1997
Creator: Hodge, S. A. & Ott, L. J.
Partner: UNT Libraries Government Documents Department

Pump system characterization and reliability enhancement

Description: Pump characterization studies were performed at the Oak Ridge National Laboratory (ORNL) to review and analyze six years (1990 to 1995) of data from pump systems at domestic nuclear plants. The studies considered not only pumps and pump motors but also pump related circuit breakers and turbine drives (i.e., the pump system). One significant finding was that the number of significant failures of the pump circuit breaker exceeds the number of significant failures of the pump itself. The study also shows how regulatory code testing was designed for the pump only and therefore did not lead to the discovery of other significant pump system failures. Potential diagnostic technologies both experimental and mature, suitable for on-line and off-line pump testing were identified. The study does not select or recommend technologies but proposes diagnostic technologies and monitoring techniques that should be further evaluated/developed for making meaningful and critically needed improvements in the reliability of the pump system.
Date: September 1, 1997
Creator: Staunton, R.H.
Partner: UNT Libraries Government Documents Department

Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

Description: For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.
Date: April 1998
Creator: Kroeger, P. G.; Rohatgi, U. S.; Jo, J. H. & Slovik, G. C.
Partner: UNT Libraries Government Documents Department

Analysis of panthers full-scale heat transfer tests with RELAP5

Description: The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric`s (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit.
Date: January 1, 1996
Creator: Parlatan, Y.; Boyer, B.D.; Jo, J. & Rohatgi, S.
Partner: UNT Libraries Government Documents Department

Summary of technical information and agreements from Nuclear Management and Resources Council industry reports addressing license renewal

Description: In about 1990, the Nuclear Management and Resources Council (NUMARC) submitted for NRC review ten industry reports (IRs) addressing aging issues associated with specific structures and components of nuclear power plants ad one IR addressing the screening methodology for integrated plant assessment. The NRC staff had been reviewing the ten NUMARC IRs; their comments on each IR and NUMARC responses to the comments have been compiled as public documents. This report provides a brief summary of the technical information and NUMARC/NRC agreements from the ten IRs, except for the Cable License Renewal IR. The technical information and agreements documented herein represent the status of the NRC staffs review when the NRC staff and industry resources were redirected to address rule implementation issues. The NRC staff plans to incorporate appropriate technical information and agreements into the draft standard review plan for license renewal.
Date: October 1, 1996
Creator: Regan, C.; Lee, S.; Chopra, O.K.; Ma, D.C. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

An evaluation of the effects of valve body erosion on motor-operated valve operability

Description: INEL engineers evaluated effects of erosion-induced valve wall thinning on motor-operated valve operability. The authors reviewed reports that identified the extent and location of erosion damage in nuclear plant valves and chose a globe valve with severe erosion damage to assess the potential for loss of operability. They developed a finite element model of the selected valve and performed structural analyses with valve closing forces, seismic effects, and increased erosion areas to analyze effects of erosion on structural integrity. Results indicate that while some local stresses at the points of maximum erosion exceeded yield, the general stresses were well below yield. Therefore, displacements will be small and bending will not occur. It is concluded that erosion-related wall thinning is not likely to create an operability problem for motor-operated valves.
Date: December 1, 1995
Creator: Hunt, T. H.; Nitzel, M. E. & Weidenhamer, G. H.
Partner: UNT Libraries Government Documents Department

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

Description: This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.
Date: March 1998
Creator: Rohatgi, U. S.; Cheng, H. S.; Khan, H. J.; Mallen, A. N. & Neymotin, L. Y.
Partner: UNT Libraries Government Documents Department

Surveillance strategy for an extended operating cycle in commercial nuclear reactors

Description: The impetus for improved economic performance of commercial nuclear power plants can be partially satisfied by increasing plant capacity factors through operating cycle extension. One aspect of an operating cycle extension effort is the modification of plant surveillance programs to complete required regulatory and investment protection surveillance activities within the extended planned outage schedule. The goal of this paper is to introduce a general strategy for existing power plants to transition their surveillance programs to an extended operating cycle up to 48 months in length, and to test the feasibility of this strategy through the complete analysis of the surveillance programs at operating BWR and PWR case study plants. The reconciliation of surveillances at these plants demonstrates that surveillance performance will not preclude 48 month operating cycles. Those surveillance activities that could not be resolved to an extended cycle are identified for further study. Finally, a number of general issues are presented that should be considered before implementing a cycle extension effort.
Date: May 1, 1997
Creator: McHenry, R.S.; Moore, T.J.; Maurer, J.H. & Todreas, N.E.
Partner: UNT Libraries Government Documents Department

Workshop on gate valve pressure locking and thermal binding

Description: The purpose of the Workshop on Gate Valve Pressure Locking and Thermal Binding was to discuss pressure locking and thermal binding issues that could lead to inoperable gate valves in both boiling water and pressurized water reactors. The goal was to foster exchange of information to develop the technical bases to understand the phenomena, identify the components that are susceptible, discuss actual events, discuss the safety significance, and illustrate known corrective actions that can prevent or limit the occurrence of pressure locking or thermal binding. The presentations were structured to cover U.S. Nuclear Regulatory Commission staff evaluation of operating experience and planned regulatory activity; industry discussions of specific events, including foreign experience, and efforts to determine causes and alleviate the affects; and valve vendor experience and recommended corrective action. The discussions indicated that identifying valves susceptible to pressure locking and thermal binding was a complex process involving knowledge of components, systems, and plant operations. The corrective action options are varied and straightforward.
Date: July 1, 1995
Creator: Brown, E.J.
Partner: UNT Libraries Government Documents Department

Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

Description: The Final Report for the Decontamination and Decommissioning (D&D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D&D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D&D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D&D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a {open_quotes}Radiologically Controlled Area,{close_quotes} noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion).
Date: March 1, 1997
Creator: Fellhauer, C.R.; Boing, L.E. & Aldana, J.
Partner: UNT Libraries Government Documents Department

Three dimensional analysis of turbulent steam jets in enclosed structures : a CFD approach.

Description: This paper compares the three-dimensional numerical simulation with the experimental data of a steam blowdown event in a light water reactor containment building. The temperature and pressure data of a steam blowdown event was measured at the Purdue University Multi-Dimensional Integrated Test Assembly (PUMA), a scaled model of the General Electric simplified Boiling Water Reactor. A three step approach was used to analyze the steam jet behavior. First, a 1-Dimensional, system level RELAP5/Mod3.2 model of the steam blowdown event was created and the results used to set the initial conditions for the PUMA blowdown experiments. Second, 2-Dimensional CFD models of the discharged steam jets were computed using PHOENICS, a commercially available CFD package. Finally, 3-Dimensional model of the PUMA drywell was created with the boundary conditions based on experimental measurements. The results of the 1-D and 2-D models were reported in the previous meeting. This paper discusses in detail the formulation and the results of the 3-Dimensional PHOENICS model of the PUMA drywell. It is found that the 3-D CFD solutions compared extremely well with the measured data.
Date: April 20, 1999
Creator: Ishii, M. & NguyenLe, Q.
Partner: UNT Libraries Government Documents Department

Experimental and Theoretical Analysis of Flashing Instability for Next Generation Natural Circulation Reactors

Description: The project had four parts: (1) Modeling and simulation of nonlinear dynamics in forced BWRs using reduced order models; (2) Modeling and simulation of nonlinear dynamics in natural circulation BWRs using reduced order models; (3) Comparison of results with those obtained using large scale system codes; and (4) Experiments to investigate natural circulation flashing phenomenon.
Date: May 13, 2005
Creator: Zboray, Robert; Kruijf, Wilhelmus J. M. de; Hagen, Tim H.J.J. van der & Rizwan-uddin
Partner: UNT Libraries Government Documents Department

Nuclear-Coupled Flow Instabilities and Their Effects on Dryout

Description: Nuclear-coupled flow/power oscillations in boiling water reactors (BWRs) are investigated experimentally and analytically. A detailed literature survey is performed to identify and classify instabilities in two-phase flow systems. The classification and the identification of the leading physical mechanisms of the two-phase flow instabilities are important to propose appropriate analytical models and scaling criteria for simulation. For the purpose of scaling and the analysis of the nonlinear aspects of the coupled flow/power oscillations, an extensive analytical modeling strategy is developed and used to derive both frequency and time domain analysis tools.
Date: September 27, 2004
Creator: Ishii, M.; Sunn, X. & Kuran, S.
Partner: UNT Libraries Government Documents Department

Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

Description: The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems
Date: September 30, 2003
Creator: Revankar, Shripad T. & Oh, Seungmin
Partner: UNT Libraries Government Documents Department

A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

Description: This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation.
Date: December 9, 2002
Creator: Chen, Hudong
Partner: UNT Libraries Government Documents Department

Development of Mechanistic Modeling Capabilities for Local Neutronically-Coupled Flow-Induced Instabilities in Advanced Water-Cooled Reactors

Description: The major research objectives of this project included the formulation of flow and heat transfer modeling framework for the analysis of flow-induced instabilities in advanced light water nuclear reactors such as boiling water reactors. General multifield model of two-phase flow, including the necessary closure laws. Development of neurton kinetics models compatible with the proposed models of heated channel dynamics. Formulation and encoding of complete coupled neutronics/thermal-hydraulics models for the analysis of spatially-dependent local core instabilities. Computer simulations aimed at testing and validating the new models of reactor dynamics.
Date: November 30, 2009
Creator: Podowski, Michael
Partner: UNT Libraries Government Documents Department

Criticality safety criteria for the handling, storage, and transportation of LWR fuel outside reactors: ANS-8.17-1984

Description: The potential for criticality accidents during the handling, storage, and transportation of fuel for nuclear reactors represents a health and safety risk to personnel involved in these activities, as well as to the general public. Appropriate design of equipment and facilities, handling procedures, and personnel training can minimize this risk. Even though the focus of the American National Standard, `Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,` ANSI/ANS-8.1-1983, is general criteria for the ensurance of criticality safety, ANS-8.17-1984, provides additional guidance applicable to handling, storage, and transportation of light-water- reactor (LWR) nuclear fuel units in any phase of the fuel cycle outside the reactor core. ANS-8.17 had its origin in the late 1970s when a work group consisting of representatives from private industry, personnel from government contractor facilities, and scientists and engineers from the national laboratories was established. The work of this group resulted in the issuance of ANSI/ANS-8.17 in January 1984. This document provides a discussion of this standard.
Date: September 1, 1996
Creator: Whitesides, G.E.
Partner: UNT Libraries Government Documents Department

Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

Description: This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.
Date: March 1, 1995
Creator: Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E. et al.
Partner: UNT Libraries Government Documents Department

Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

Description: This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.
Date: May 1, 1997
Partner: UNT Libraries Government Documents Department