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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

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

MELCOR model for an experimental 17x17 spent fuel PWR assembly.

Description: A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.
Date: November 1, 2010
Creator: Cardoni, Jeffrey
Partner: UNT Libraries Government Documents Department

Tabulation of Fundamental Assembly Heat and Radiation Source Files

Description: The purpose of this calculation is to tabulate a set of computer files for use as input to the WPLOAD thermal loading software. These files contain details regarding heat and radiation from pressurized water reactor (PWR) assemblies and boiling water reactor (BWR) assemblies. The scope of this calculation is limited to rearranging and reducing the existing file information into a more streamlined set of tables for use as input to WPLOAD. The electronic source term files used as input to this calculation were generated from the output files of the SAS2H/ORIGIN-S sequence of the SCALE Version 4.3 modular code system, as documented in References 2.1.1 and 2.1.2, and are included in Attachment II.
Date: October 25, 2006
Creator: deBues, T. & Ryman, J.C.
Partner: UNT Libraries Government Documents Department

LOFT Monthly Progress Report for December 1980

Description: On December 10, 1980, LOFT conducted its fifth nuclear test in the L3 Series (small break) as well as the first in the L8 Series (core uncovery). The tests, designated L3-6/L8-1, were run in series with each experiment designated to address specific safety questions. Test L3-6 simulated a four-inch break in a commercial pressurized water reactor. The purpose of this experiment was to investigate the influence of main coolant pump operation on the quantity of fluid which leaves the system and the quantity of fluid in the reactor core region during the experiment. The results of L3-6 are being compared with the results of a previous LOFT test, designated L3-5, which was performed on September 29, 1980. On December 9 and 10, a special review group committee, commissioned by the NRC, visited LOFT. The purpose of the committee's presence here was to study and evaluate the LOFT project, and then report recommendations to the NRC. The NRC would then use this information to help determine how the resources of the project could best be utilized, and which kind and how many tests should be conducted over how long a time period. Recommendations of this committee are expected to be announced early in February. The Quick Look Report on experiment L3-6/L8-1, reviewing initial conditions and limited experimental results, was published during the month. More extensive data analysis will continue for the next several months. The current planned LOFT test schedule and target dates were rescheduled in December which resulted in a major budget revision to the baseline. Successful completion and approval is anticapted by mid-January 1981. Overall current costs are in good agreement with budgets and authorized funding levels.
Date: January 1, 1981
Creator: Leach, L. P.
Partner: UNT Libraries Government Documents Department

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

Description: The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.
Date: July 29, 1997
Creator: Wright, Kenneth D.
Partner: UNT Libraries Government Documents Department

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

Description: The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.
Date: July 30, 1997
Creator: Wright, Kenneth D.
Partner: UNT Libraries Government Documents Department

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

Description: The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.
Date: February 8, 2001
Creator: Wilson, Michael L.
Partner: UNT Libraries Government Documents Department

CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

Description: The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies.
Date: September 3, 1997
Creator: Wright, Kenneth D.
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