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Plant systems/components modularization study. Final report. [PWR]

Description: The final results are summarized of a Plant Systems/Components Modularization Study based on Stone and Webster's Pressurized Water Reactor Reference Design. The program has been modified to include evaluation of the most promising areas for modular consideration based on the level of the Sundesert Project engineering design completion and the feasibility of their incorporation into the plant construction effort.
Date: July 1, 1977
Partner: UNT Libraries Government Documents Department

Evaluation of tight-pitch PWR cores

Description: The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - /sup 235/U/UO/sub 2/ : Pu/ThO/sub 2/ : /sup 233/U/ThO/sub 2/ - and the conventional recycle-mode uranium system - /sup 235/U/UO/sub 2/ : Pu/UO/sub 2/. The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) < F/M < 4.0 are limited by the scarcity of experiments with F/M > 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments.
Date: August 1, 1979
Creator: Correa, F.; Driscoll, M.J. & Lanning, D.D.
Partner: UNT Libraries Government Documents Department

PWR decontamination feasibility study

Description: The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.
Date: December 18, 1978
Creator: Silliman, P.L.
Partner: UNT Libraries Government Documents Department

Technical evaluation of the proposed technical specification change for the Arkansas Nuclear Power Station, Unit 2

Description: This report documents the technical evaluation of the request for changes in the Technical Specifications for the Arkansas Nuclear Power Station, Unit 2. These changes were proposed by the licensee in a letter dated November 27, 1979. The basis for review included a report entitled Determination of Plant System Trip Setpoints Valves. The requested changes to the Technical Specifications were found to be acceptable based on information submitted by the licensee.
Date: August 1, 1980
Creator: Victor, R.A.
Partner: UNT Libraries Government Documents Department

Calculated neutron-source spectra from selected irradiated PWR fuel assemblies

Description: The energy spectra of neutrons emitted from a pressurized-water-reactor fuel assembly have been calculated for a variety of exposures and cooling times. They are presented in graphical form. Some effects of initial enrichment are also included. Neutrons from spontaneous fissions were given either a Maxwellian temperature of 1.2 or 1.5 MeV, depending on whether they were due to plutonium and uranium nuclides or curium nuclides. A single (..cap alpha..,n) spectrum was deemed sufficient to represent the neutrons from all the alpha-emitting nuclides. The proportions of the nuclides undergoing spontaneous fission and those emitting alpha particles were determined from calculated atom densities. The particular pressurized-water-reactor fuel assembly assumed for this purpose was of the type used in the H.B. Robinson Unit-2 power plant (740 MWe).
Date: December 1, 1981
Creator: Rinard, P.M.; Bosler, G.E. & Phillips, J.R.
Partner: UNT Libraries Government Documents Department

In-core detector activation rate for a PWR assembly

Description: The in-core detector system is the principal source of information for determining relative assembly powers, and maximum fuel rod powers in a reactor core. The detector signals are used in conjunction with pre-calculated factors, and appropriate normalizations, to obtain measured power values. Considerable reliance is placed on the accuracy of in-core detector inferred power distributions in reactor operations, and in the verification of calculational methods. The objective of this study was to compare results from standard design codes for the in-core detector activation rate (and the fission rate distribution in an assembly), to results obtained from a detailed calculation performed with a continuous energy Monte Carlo program with ENDF/B-V nuclear data.
Date: January 1, 1982
Creator: Todosow, M. & Eisenhart, L.D.
Partner: UNT Libraries Government Documents Department

Piping-reliability analysis for pressurized-water-reactor feedwater lines

Description: This paper presents a piping reliability analysis for feedwater lines at five PWR plants; the analysis is based on probabilistic fracture mechanics. On the basis of observed pipe cracks in these feedwater lines, the crack is modeled with an initial semi-elliptical shape along the pipe inner circumference. Initial crack samples are generated using the Monte Carlo technique in conjunction with an importance sampling scheme. The fatigue model for crack growth employs a Paris-type growth-rate equation. Leak probabilities for feedwater lines of five PWR plants are estimated through the first ten years of the design life. Comparison of estimated leak probabilities and leak data in the five PWR plants led to the conclusion that the piping reliability model used in the present analysis can provide a reasonable estimate of the reliability of PWR feedwater lines.
Date: January 1, 1982
Creator: Woo, H.H. & Chou, C.K.
Partner: UNT Libraries Government Documents Department

Once-through thorium cycle for BWR's

Description: Problems in application of thorium cycles include greater fissile inventory requirements, the blending of highly enriched uranium or plutonium with thorium, and the necessity to recover and recycle the valuable U-233 produced in order to recover the costs of the initial inventory and enrichment. With these problems in mind, a once-through thorium cycle was developed for Boiling Water Reactors which minimizes the effect of thorium on the fissile inventory, which is initiated with ThO/sub 2/ fuel containing no initial fissile material, and which does not require U-233 recovery and recycle to make the application economically competitive. The design makes advantageous use of the inherent lattice heterogeneity and other characteristics of the BWR lattice to produce U-233 in ThO/sub 2/ without power distribution penalties and to improve reactor performance (thermal and transient margins). Standard BWR fuel assembly hardware was used to make the design backfitable with minimum manufacturing impact. Preliminary conclusions are that the once-through thorium application has potential to both reduce uranium ore requirements and increase BWR operating margins.
Date: January 1, 1979
Creator: Townsend, D.B.; Crowther, R.L. & Wolters, R.A.
Partner: UNT Libraries Government Documents Department

Refueling outage availability study. Phase 1 final report

Description: Babcock and Wilcox entered into a contract with the Department of Energy (formerly the Energy Research and Development Administration) for the performance of a refueling outage availability study with the cooperation of Duke Power Company and Arkansas Power and Light Company. The objective was to improve plant availability through reduction of refueling outage time. The conclusions of the study were drawn from data gathered during the 1976 Oconee 3 and 1977 Arkansas Unit One refueling outages. The onsite effort was one of observation and data recording, which included time-lapsed photography and video tape. The collected data were then evaluated and analyzed for potential improvements and to identify in detail where resources were consumed. The overall result was a listing of (1) specific recommendations for implementing improvements in the facilities, equipment, tools, procedures, and techniques for the participating utilities; (2) generic recommendations of immediate benefit to other applicable utilities; and (3) recommendations for further work in the succeeding phases of the DOE program. The results indicate that, by incorporating the recommendations and taking credit for the time savings, an ideal refueling outage length of 21 to 22 days for the nuclear steam system (NSS) could be realized. Additional benefits would be a reduction in man-Rem exposure and manpower requirements.
Date: November 1, 1977
Creator: Thomasson, F.R.
Partner: UNT Libraries Government Documents Department

Steam generator tube integrity program. Phase I report. [PWR]

Description: The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators.
Date: September 1, 1979
Creator: Alzheimer, J.M.; Clark, R.A.; Morris, C.J. & Vagins, M.
Partner: UNT Libraries Government Documents Department

Technology, safety and costs of decommissioning a reference pressurized water reactor power station

Description: Additional analyses of decommissioning of the reference pressurized water reactor (PWR) power station are made that examine some parameters not covered in the initial study report (NUREG/CR-0130). The parameters examined are: (1) the effect of plant size on costs and radiation exposure for dismantlement, (2) the costs and radiaton exposures associated with entombment of the reference PWR, (3) the impact on costs and radiation exposure of higher radiation dose rates throughout the facility than were assumed in the initial study, (4) the effect on costs of using individual contractors to accomplish dismantlement rather than using the utility staff as postulated in the initial study, and (5) the effect on costs of increasing disposal charges at waste disposal facilities.
Date: August 1, 1979
Creator: Smith, R.I. & Polentz, L.M.
Partner: UNT Libraries Government Documents Department

Determination of K-factors for arbitrarily shaped flaws at pressure vessel nozzle corners

Description: Photoelastic and finite element studies are being conducted to determine Mode I stress intensity factor distributions along arbitrarily shaped flaw fronts at pressure vessel nozzle corners. Comparisons of results from NOZ-FLAW, BIGIF, and the photoelastic studies showed that (1) good agreement was obtained between NOZ-FLAW and the photoelastically determined K/sub 1/'s for the deep flaw in an ITV model, (2) good agreement was obtained between NOZ-FLAW BIGIF for shallow and moderately deep flaws in a BWR model, and (3) less satisfactory agreement was obtained between NOZ- FLAW and the photoelastic results for the BWR models, particularly for moderately deep to deep flaws. Attempts are presently being made at understanding and explaining the discrepancies between the two.
Date: January 1, 1979
Creator: Bryson, J.W.
Partner: UNT Libraries Government Documents Department

Oconee 1, cycle 5 design report

Description: The Oconee 1, cycle 5 fuel cycle was designed to irradiate five fuel assemblies to a burnup of approximately 40,000 MWd/mtU. The fuel cycle design was performed using standard Babcock and Wilcox calculational techniques for nuclear fuel cycles. All applicable design criteria were satisfied. Valuable experimental data on the performance characteristics of high-burnup fuel assemblies will be obtained from these assemblies in subsequent post-irradiation examinations.
Date: May 1, 1979
Creator: Coleman, T.A. (ed.)
Partner: UNT Libraries Government Documents Department

Nondestructive examination of Oconee 1 fuel assemblies after three cycles of irradiation

Description: The Babcock and Wilcox Company (B and W) in conjunction with Duke Power Company is participating in a Department of Energy sponsored research and development program to qualify current design pressurized water reactor (PWR) fuel assemblies for extended burnup (>40,000 MWd/mtU). The information obtained from this program will provide a basis for future design improvements in PWR fuel assemblies culminating in an extended burnup assembly having a nominal operating limit of approximately 50,000 MWd/mtU. An extension of the current assembly design to higher burnups will result in the following benefits: (1) lower uranium ore requirements, (2) greater fuel cycle efficiency, (3) reduction in spent fuel storage requirements, and (4) increased flexibility in tailoring fuel batch sizes to better accommodate the varying energy requirements of the utilities.
Date: September 1, 1979
Creator: Pyecha, T.D.; Davis, H.H.; Mayer, J.T.; Guthrie, B.A. III & Larson, J.G.
Partner: UNT Libraries Government Documents Department

Failure analysis of tubes with wastages. [PWR]

Description: A finite element method for large strain calculation using the constitutive relation due to Hill was developed. This constitutive relation relates the co-rotational rate of the Kirchoff stress and deformation rate tensor which leads to a symmetric structure stiffness. This method is used to calculate failure pressures of degraded tubes.
Date: January 1, 1979
Creator: Prachuktam, S.; Reich, M. & Rajan, J.
Partner: UNT Libraries Government Documents Department

Free vibration analysis of a steam generator tube bundle with and without lateral support. [PWR]

Description: The vibrational modes and frequency characteristics of a pressurized water reactor (PWR) steam generator tube bundle assembly with and without lateral support in a fluid environment are analyzed. The idealized half-model was constructed using the SAP-IV finite element code. Free vibration analyses were performed for an in-air case and a submerged in-water case, each with different constraint conditions at steam generator tube bundle assembly support plates 10 and 11. These constraint conditions included having both support plates free, having both support plates fixed, and having support plate 11 free while support plate 10 was fixed. It was found that as the support plate constraints were removed, the frequency range for each case increased significantly.
Date: April 1, 1979
Creator: King, D.M.
Partner: UNT Libraries Government Documents Department

Bond graph modeling of nuclear reactor dynamics. [PWR]

Description: A tenth-order linear model of a pressurized water reactor (PWR) is developed using bond graph techniques. The model describes the nuclear heat generation process and the transfer of this heat to the reactor coolant. Comparisons between the calculated model response and test data from a small-scale PWR show the model to be an adequate representation of the actual plant dynamics. Possible application of the model in an advanced plant diagnostic system is discussed.
Date: January 1, 1981
Creator: Tylee, J.L.
Partner: UNT Libraries Government Documents Department

Fuel Utilization Improvements in a Once-Through PWR Fuel Cycle. Final Report on Task 6

Description: In studying the position of the United States Department of Energy, Non-proliferation Alternative Systems Assessment Program, this report determines the uranium saving associated with various improvement concepts applicable to a once-through fuel cycle of a standard four-loop Westinghouse Pressurized Water Reactor. Increased discharged fuel burnup from 33,000 to 45,000 MWD/MTM could achieve a 12% U/sub 3/O/sub 8/ saving by 1990. Improved fuel management schemes combined with coastdown to 60% power, could result in U/sub 3/O/sub 8/ savings of 6%.
Date: June 1, 1979
Creator: Dabby, D.
Partner: UNT Libraries Government Documents Department

Standardized Technical Specifications for Westinghouse PWRs

Description: This Standard Technical Specification (STS) has been structured for the broadest possible use on Westinghouse plants currently being reviewed for an Operating License. Accordingly, the document contains specifications applicable to plants (1) with either 3 or 4 loops and (2) with and without loop stop valves. In addition, four separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, Ice Condenser, Sub-Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. The format of the STS addresses the categories required by 10 CFR 50 and consists of six sections covering the areas of: Definitions, Safety Limits and Limiting Safety System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls.
Date: June 15, 1978
Partner: UNT Libraries Government Documents Department

Standard technical specifications for Babcock and Wilcox pressurized water reactors

Description: The Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. The format of the STS addresses the categories required by 10 CFR 50 and consists of six sections covering the areas of: Definitions, Safety Limits and Limiting Safety System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls.
Date: June 1, 1978
Partner: UNT Libraries Government Documents Department

Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

Description: The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction.
Date: April 1, 1978
Creator: Mayers, J.B. & Soth, L.G.
Partner: UNT Libraries Government Documents Department

U. S. experience with in-service monitoring of core barrel motion in PWRs using ex-core neutron detectors

Description: Coolant flow forces, pressure pulsations, and reactor primary system mechanical vibrations in a PWR combine to cause pendular (and other more complicated) motions of the reactor core support barrel within its surrounding pressure vessel. These motions are normally quite small (a few thousandths of an inch) and constitute no immediate safety problem, but in view of one past occurrence where some internal structural damage resulted, the U.S. Nuclear Regulatory Commission is considering making routine in-service monitoring for excessive core barrel motion mandatory for all PWRs. Should this be judged necessary, it is our opinion that quantitative in-service monitoring can be performed in a manner that is acceptable both to the Commission and to the nuclear plant operators by decomposing and interpreting the signals from the ex-core, power-range neutron monitors that are already a part of standard PWR instrumentation.
Date: January 1, 1978
Creator: Kryter, R C; Robinson, J C & Thie, J A
Partner: UNT Libraries Government Documents Department

Modeling and diagnostic techniques applicable to the analysis of pressure noise in pressurized water reactors and pressure-sensing systems

Description: Pressure noise data from a PWR are interpreted by means of a computer-implemented model. The model's parameters, namely hydraulic impedances and noise sources, are either calculated or deduced from fits to data. Its accuracy is encouraging and raises the possibility of diagnostic assistance for nuclear plant monitoring. A number of specific applications of pressure noise in the primary system of a PWR and in a pressure sensing system are suggested.
Date: January 1, 1984
Creator: Mullens, J.A. & Thie, J.A.
Partner: UNT Libraries Government Documents Department

Bias identification in PWR pressurizer instrumentation using the generalized liklihood-ratio technique

Description: A method for detecting and identifying biases in the pressure and level sensors of a pressurized water reactor (PWR) pressurizer is described. The generalized likelihood ratio (GLR) technique performs statistical tests on the innovations sequence of a Kalman filter state estimator and is capable of determining when a bias appears, in what sensor the bias exists, and estimating the bias magnitude. Simulation results using a second-order linear, discrete PWR pressurizer model demonstrate the capabilities of the GLR method.
Date: January 1, 1981
Creator: Tylee, J.L.
Partner: UNT Libraries Government Documents Department