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Comparison of ASME Code NB-3200 and NB-3600 results for fatigue analysis of B31.1 branch nozzles

Description: Fatigue analyses wre conducted on two reactor coolant system branch nozzles in an operating PWR designed to the B31.1 Code, for which no explicit fatigue analysis was required by the licensing basis. These analyses were performed as part of resolving issues connected with NRC`s Fatigue Action Plan to determine if the cumulative usage factor (CUF) for these nozzles, using the 1992 ASME Code and representative PWR transients, were comparable to nozzles designed and analyzed to the ASME Code. Both NB-3200 and NB-3600 ASME Code methods were used. NB-3200 analyses included the development of finite element models for each nozzle. Although detailed thermal transients were not available for the plant analyzed, representative transients from similar PWRs were applied in each method. CUFs calculated using NB-3200 methods were significantly less than using NB-3600. The paper points out differences in analysis methods and highlights difficulties and unknowns in performing more detailed analyses to reduce conservative assumptions.
Date: June 1, 1996
Creator: Nitzel, M.E.; Ware, A.G. & Morton, D.K.
Partner: UNT Libraries Government Documents Department

Advanced pressurized water reactor for improved resource utilization, part II - composite advanced PWR concept

Description: This report evaluates the enhanced resource utilization in an advanced pressurized water reactor (PWR) concept using a composite of selected improvements identified in a companion study. The selected improvements were in the areas of reduced loss of neutrons to control poisons, reduced loss of neutrons in leakage from the core, and improved blanket/reflector concepts. These improvements were incorporated into a single composite advanced PWR. A preliminary assessment of resource requirements and costs and impact on safety are presented.
Date: September 15, 1981
Creator: Turner, S.E.; Gurley, M.K.; Kirby, K.D. & Mitchell, W III
Partner: UNT Libraries Government Documents Department

Results of a nuclear power plant Application of a new technique for human error analysis (ATHEANA)

Description: A new method to analyze human errors has been demonstrated at a pressurized water reactor (PWR) nuclear power plant. This was the first application of the new method referred to as A Technique for Human Error Analysis (ATHEANA). The main goals of the demonstration were to test the ATHEANA process as described in the frame-of-reference manual and the implementation guideline, test a training package developed for the method, test the hypothesis that plant operators and trainers have significant insight into the error-forcing-contexts (EFCs) that can make unsafe actions (UAs) more likely, and to identify ways to improve the method and its documentation. A set of criteria to evaluate the {open_quotes}success{close_quotes} of the ATHEANA method as used in the demonstration was identified. A human reliability analysis (HRA) team was formed that consisted of an expert in probabilistic risk assessment (PRA) with some background in HRA (not ATHEANA) and four personnel from the nuclear power plant. Personnel from the plant included two individuals from their PRA staff and two individuals from their training staff. Both individuals from training are currently licensed operators and one of them was a senior reactor operator {open_quotes}on shift{close_quotes} until a few months before the demonstration. The demonstration was conducted over a 5 month period and was observed by members of the Nuclear Regulatory Commission`s ATHEANA development team, who also served as consultants to the HRA team when necessary. Example results of the demonstration to date, including identified human failure events (HFEs), UAs, and EFCs are discussed. Also addressed is how simulator exercises are used in the ATHEANA demonstration project.
Date: October 1, 1997
Creator: Forester, J.A.; Whitehead, D.W.; Kolaczkowski, A.M. & Thompson, C.M.
Partner: UNT Libraries Government Documents Department

Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

Description: This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.
Date: September 26, 1996
Creator: Lata
Partner: UNT Libraries Government Documents Department

Analysis and Measurement of Bubble Dynamics and Associated Flow Field in Subcooled Nucleate Boiling Flows

Description: In recent years, subooled nucleate boiling (SNB) has attrcted expanding research interest owing to the emergence of axial offset anomaly (AOA) or crud-induced power shigt (CIPS) in many operating US PWRs, which is an unexpected deviation in the core axial power distribution from the predicted power curves. Research indicates that the formation of the crud, which directly leads to AOA phenomena, results from the presence of the subcooled nucleate boiling, and is especially realted to bubble motion occurring in the core region.
Date: October 1, 2008
Creator: Jones, Barclay G.
Partner: UNT Libraries Government Documents Department

Analysis of an AP600 intermediate-size loss-of-coolant accident

Description: A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.
Date: April 1, 1995
Creator: Boyack, B.E. & Lime, J.F.
Partner: UNT Libraries Government Documents Department

CASL L1 Milestone report : CASL.P4.01, sensitivity and uncertainty analysis for CIPS with VIPRE-W and BOA.

Description: The CASL Level 1 Milestone CASL.P4.01, successfully completed in December 2011, aimed to 'conduct, using methodologies integrated into VERA, a detailed sensitivity analysis and uncertainty quantification of a crud-relevant problem with baseline VERA capabilities (ANC/VIPRE-W/BOA).' The VUQ focus area led this effort, in partnership with AMA, and with support from VRI. DAKOTA was coupled to existing VIPRE-W thermal-hydraulics and BOA crud/boron deposit simulations representing a pressurized water reactor (PWR) that previously experienced crud-induced power shift (CIPS). This work supports understanding of CIPS by exploring the sensitivity and uncertainty in BOA outputs with respect to uncertain operating and model parameters. This report summarizes work coupling the software tools, characterizing uncertainties, and analyzing the results of iterative sensitivity and uncertainty studies. These studies focused on sensitivity and uncertainty of CIPS indicators calculated by the current version of the BOA code used in the industry. Challenges with this kind of analysis are identified to inform follow-on research goals and VERA development targeting crud-related challenge problems.
Date: December 1, 2011
Creator: Sung, Yixing (Westinghouse Electric Company LLC, Cranberry Township, PA); Adams, Brian M. & Secker, Jeffrey R. (Westinghouse Electric Company LLC, Cranberry Township, PA)
Partner: UNT Libraries Government Documents Department

THERMAL EVALUATION OF PRELIMINARY 21 PWR AUCF DESIGN

Description: This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8, 4-11, 4-24, 5-1, and 5-13, Ref. 5.10) and the Waste Package Plan (pp.3-15, 3-17, and 3-24, Ref. 5.9). The design data request addressed herein is: Characterize the preliminary 21 pressurized water reactor (PWR) advanced (A) uncanistered fuel (UCF) waste package (WP) to show that the design is feasible for use in the MGDS environment. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the UCF WP do not preclude UCF WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the preliminary UCF WP design under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation and to provide the required guidance to determining the major design issues for future design efforts. Future design efforts will focus on UCF design changes as further design and operations information becomes available.
Date: May 10, 1996
Creator: Wang, H.
Partner: UNT Libraries Government Documents Department

THERMAL EVALUATION OF THE CONCEPTUAL 12 PWR UNCANISTERED FUEL (UCF) TUBE BASKET DESIGN DISPOSAL CONTAINER

Description: This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8, 4-11, 4-24, 5-1, and 5-13. Ref. 5.10) and Waste package Plan (pp. 3-15, 3-17, and 3-24, Ref. 5.9). The design data request addressed herein is: Characterize the conceptual 12 pressurized water reactor (PWR) uncanistered fuel (UCF) waste package (WP) to show that the design is feasible for use in the MGDS environment. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the UCF WP do not preclude UCF WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual UCF WP design under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation and to provide the required guidance to determining the major design issues for future design efforts. Future design efforts will focus on UCF design changes as further design and operations information becomes available.
Date: March 1, 1996
Creator: Wang, H.
Partner: UNT Libraries Government Documents Department

SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

Description: This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.
Date: August 29, 1996
Creator: Davis, J.W.
Partner: UNT Libraries Government Documents Department

UCF WP TIPOVER ANALYSIS

Description: The purpose of this analysis is to determine the structural response of the 21 pressurized water reactor (PWR) uncanistered fuel (UCF) waste package (WP) to a tipover design basis event (DBE) dynamic load; the results will be reported in terms of stress magnitudes. Finite-element solution was performed by making use of the commercially available ANSYS finite-element code. A finite-element model of the waste package was developed and analyzed for a tipover DBE dynamic load. The results of this analysis were provided in tables and were also plotted in terms of the maximum stress contours to determine their locations.
Date: April 28, 1998
Creator: Ceylan, Z.
Partner: UNT Libraries Government Documents Department

Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

Description: The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only.
Date: January 29, 2001
Creator: Mastilovic, S.; Scheider, A. & Bennett, S.M.
Partner: UNT Libraries Government Documents Department

Characteristics of Spent Fuel from Plutonium Disposition Reactors. Vol. 3: A Westinghouse Pressurized-Water Reactor Design

Description: This report discusses the results of a simulation study involving the burnup of mixed-oxide (MOX) fuel in a Westinghouse pressurized-water reactor (PWR). The MOX was composed of uranium and plutonium oxides, where the plutonium was of weapons-grade composition. The study was part of the Fissile Materials Disposition Program and considered the possibility of fueling commercial reactors with weapons plutonium. The isotopic composition, the activities, and the decay heat, together with the gamma and neutron dose rates are discussed for the spent fuel. For the steady-state situation involving this PWR burning MOX fuel, two burn histories are reported. In one case, an assembly is burned in the reactor for two cycles, and in the second case and assembly is burned for three cycles. Furthermore, assemblies containing wet annular burnable absorbers (WABAs) and assemblies that do not contain WABAs are considered in all cases. The two-cycle cases have a burnup of 35 GWd/t, and the three-cycle cases have a burnup of 52.5 GWd/t.
Date: July 1, 1997
Creator: Murphy, B.D.
Partner: UNT Libraries Government Documents Department

Validation of a method for prediction of isotopic concentrations in burnup credit applications

Description: Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of the spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. This paper describes a set of experimentally characterized pressurized-water-reactor (PWR) fuel samples and provides a comparison to results of SCALE-4 depletion calculations. An approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations.
Date: September 1, 1995
Creator: DeHart, M.D.; Hermann, O.W. & Parks, C.V.
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

Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

Description: This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor.
Date: March 1, 1997
Partner: UNT Libraries Government Documents Department

Fracture toughness evaluation of a low upper-shelf weld metal from the Midland Reactor using the master curve

Description: The primary objective of the Heavy-Section Steel Irradiation (HSSI) Program Tenth Irradiation Series was to develop a fracture mechanics evaluation of weld metal WF-70, which was taken from the beltline and nozzle course girth weld joints of the Midland Reactor vessel. This material became available when Consumers Power Company of Midland, Michigan, decided to abort plans to operate their nuclear power plant. WF-70 is classified as a low upper-shelf steel primarily due to the Linde 80 flux that was used in the submerged-arc welding process. The master curve concept is introduced to model the transition range fracture toughness when the toughness is quantified in terms of K{sub Jc} values. K{sub Jc} is an elastic-plastic stress intensity factor calculated by conversion from J{sub c}; i.e., J-integral at onset of cleavage instability.
Date: March 1997
Creator: McCabe, D. E.; Sokolov, M. A. & Nanstad, R. K.
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

Development of a standard for calculation and measurement of the moderator temperature coefficient of reactivity in water-moderated power reactors

Description: The contents of ANS 19.11, the standard for ``Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity in Water-Moderated Power Reactors,`` are described. The standard addresses the calculation of the moderator temperature coefficient (MTC) both at standby conditions and at power. In addition, it describes several methods for the measurement of the at-power MTC and assesses their relative advantages and disadvantages. Finally, it specifies a minimum set of documentation requirements for compliance with the standard.
Date: December 1, 1998
Creator: Mosteller, R.D.; Hall, R.A.; Apperson, C.E. Jr.; Lancaster, D.B.; Young, E.H.; Gavin, P.H. et al.
Partner: UNT Libraries Government Documents Department

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

Description: The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.
Date: February 2, 2001
Creator: Ellis, RJ
Partner: UNT Libraries Government Documents Department

Spent fuel test project, Climax granitic stock, Nevada Test Site

Description: The Spent Fuel Test-Climax (SFT-C) is a test of dry geologic storage of spent nuclear reactor fuel. The SFT-C is located at a depth of 420 m in the Climax granitic stock at the Nevada Test Site. Eleven canisters of spent commercial PWR fuel assemblies are to be stored for 3 to 5 years. Additional heat is supplied by electrical heaters, and more than 800 channels of technical information are being recorded. The measurements include rock temperature, rock displacement and stress, joint motion, and monitoring of the ventilation air volume, temperature, and dewpoint.
Date: October 24, 1980
Creator: Ramspott, L.D.
Partner: UNT Libraries Government Documents Department

Proceedings of the USNRC/EPRI/ANL heated crevice seminar.

Description: An international Heated Crevice Seminar, sponsored by the Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Argonne National Laboratory, and the Electric Power Research Institute, was held at Argonne National Laboratory on October 7-11, 2002. The objective of the seminar was to provide a working forum for the exchange of information by contributing experts on current issues related to corrosion in heated crevices, particularly as it relates to the integrity of PWR steam generator tubes. Forty-five persons from six countries attended the seminar, including representatives from government agencies, private industry and consultants, government research laboratories, nuclear vendors, and electrical utilities. The seminar opened with keynote talks on secondary-side crevice environments associated with IGA and IGSCC of mill-annealed Alloy 600 steam generator tubes and the submodes of corrosion in heat transfer crevices. This was followed by technical sessions on (1) Corrosion in Crevice Geometries, (2) Experimental Methods, (3) Results from Experimental Studies, and (4) Modeling. The seminar concluded with a panel discussion on the present understanding of corrosive processes in heated crevices and future research needs.
Date: August 31, 2003
Creator: Park, J. Y.; Fruzzetti, K.; Muscara, J.; Diercks, D. R.; Technology, Energy; EPRI et al.
Partner: UNT Libraries Government Documents Department