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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

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

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

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

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

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

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

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

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

Review of computational thermal-hydraulic modeling

Description: Corrosion of heat transfer tubing in nuclear steam generators has been a persistent problem in the power generation industry, assuming many different forms over the years depending on chemistry and operating conditions. Whatever the corrosion mechanism, a fundamental understanding of the process is essential to establish effective management strategies. To gain this fundamental understanding requires an integrated investigative approach that merges technology from many diverse scientific disciplines. An important aspect of an integrated approach is characterization of the corrosive environment at high temperature. This begins with a thorough understanding of local thermal-hydraulic conditions, since they affect deposit formation, chemical concentration, and ultimately corrosion. Computational Fluid Dynamics (CFD) can and should play an important role in characterizing the thermal-hydraulic environment and in predicting the consequences of that environment,. The evolution of CFD technology now allows accurate calculation of steam generator thermal-hydraulic conditions and the resulting sludge deposit profiles. Similar calculations are also possible for model boilers, so that tests can be designed to be prototypic of the heat exchanger environment they are supposed to simulate. This paper illustrates the utility of CFD technology by way of examples in each of these two areas. This technology can be further extended to produce more detailed local calculations of the chemical environment in support plate crevices, beneath thick deposits on tubes, and deep in tubesheet sludge piles. Knowledge of this local chemical environment will provide the foundation for development of mechanistic corrosion models, which can be used to optimize inspection and cleaning schedules and focus the search for a viable fix.
Date: December 31, 1995
Creator: Keefer, R.H. & Keeton, L.W.
Partner: UNT Libraries Government Documents Department

Summary of results for the uranium benchmark problem of the ANS Ad Hoc Committee on Reactor Physics Benchmarks

Description: This paper presents a summary of the results obtained by all of the contributors to the Uranium Benchmark Problem of the ANS Ad hoc Committee on Reactor Physics Benchmarks. The benchmark problem was based on critical experiments which mocked-up lattices typical of PWRs. Three separate cases constituted the benchmark problem. These included a uniform lattice, an assembly-type lattice with water holes and an assembly-type lattice with pyrex rods. Calculated results were obtained from eighteen separate organizations from all over the world. Some organizations submitted more than one set of results based on different calculational methods and cross section data. Many of the most widely used assembly physics and core analysis computer codes and neutron cross section data libraries were applied by the contributors.
Date: December 31, 1998
Creator: Parish, T.A.; Mosteller, R.D.; Diamond, D.J. & Gehin, J.C.
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

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

Heat transfer characteristics of porous sludge deposits and their impact on the performance of commercial steam generators

Description: Steam generator (SG) fouling, in the form of corrosion deposits on the secondary sides of SG tubes, has been known to occur in almost all commercial US nuclear PWR (pressurized water reactor) plants. The level of fouling, as measured by the quantity of corrosion products that form, varies widely from plant to plant. In addition, the effect of SG fouling, as measured by a decrease in effective heat-transfer coefficient, has also varied substantially among commercial US plants. While some have observed large decreases in heat transfer, others have noted little change in performance despite the presence of significant quantities of secondary corrosion layers on their SG tubes. This observation has led to considerable confusion about what role secondary deposits play in causing heat-transfer degradation in SGs. As will become clear later in this report, secondary deposits can have a wide range of effects on heat transfer, from highly resistive to slightly enhancing (reflected by negative fouling). These different behaviors are the result of differences in deposit thickness, composition, and morphology. The main focus of this report is an investigation of the effects of secondary deposits on SG thermal performance. This investigation includes compilation of detailed information on the properties of tube scale at five commercial US nuclear plants and corresponding information characterizing SG thermal performance at these plants.
Date: December 1, 1998
Creator: Kreider, M.A.; White, G.A.; Varrin, R.D. & Ouzts, P.J.
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

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

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

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

Quick look data report for COMET Test U2

Description: Investigations are underway at Forschungszentrum Karlsruhe (FZK) addressing methods to terminate and stabilize a core melt accident situation ex-vessel. In this approach, the molten core-concrete interaction (MCCI) begins erosion of the concrete, and after erosion proceeds to some modest depth, it exposes and unseals an array of tubes. The tubes are connected to a water reservoir pressurized by static water head. Upon unsealing, the tubes direct a flow of water into the bottom of the corium layer. The water is forced up through the melt, cooling the melt and causing it to solidify in a form that allows continued permeation and heat removal by the water. Thus, the accident progression can be halted, and the debris may be permanently cooled. The key aspect of the passive ex-vessel core retention approach described above is the ability of water injected at the bottom of a corium melt layer to quench the melt forming a coolable debris bed in the process. This process has been tested using iron-alumina thermite as a corium simulant with promising results. As a part of a collaborative research agreement between FZK and the US DOE, two scoping tests are being conducted at Argonne National Laboratory to test the FZK core retention concept using real reactor materials. The second of these two tests, denoted COMET Test U2, was successfully conducted on December 17, 1997. The objectives of this data report are to: summarize the experiment facility and operating procedure for COMET Test U2, and present the test data.
Date: January 8, 1998
Creator: Farmer, M. T.; Spencer, B. W.; Kilsdonk, D. J. & Aeschlimann, R. W.
Partner: UNT Libraries Government Documents Department