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Sample records for review pwr bwr

  1. Proceedings: 2001 PWR/BWR Plant Chemistry Meeting

    SciTech Connect

    2001-05-01

    This report presents proceedings of EPRI's 2001 Plant Chemistry Conference, which brought together approximately 100 industry representatives to discuss experiences and issues regarding nuclear plant chemistry at both pressurized water reactor (PWR) and boiling water reactor (BWR) plants.

  2. PWR and BWR spent fuel assembly gamma spectra measurements

    NASA Astrophysics Data System (ADS)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  3. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGESBeta

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; et al

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  4. Beta and gamma dose calculations for PWR and BWR containments

    SciTech Connect

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 /times/ 10/sup 8/ rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 /times/ 10/sup 8/ rad equipment qualification test region. 8 refs., 23 figs., 12 tabs.

  5. Experience in PWR and BWR mixed-oxide fuel management

    SciTech Connect

    Schlosser, G.J.; Krebs, W.; Urban, P. )

    1993-04-01

    Germany has adopted the strategy of a closed fuel cycle using reprocessing and recycling. The central issue today is plutonium recycling by the use of U-Pu mixed oxide (MOX) in pressurized water reactors (PWRs) and boiling water reactors (BWRs). The design of MOX fuel assemblies and fuel management in MOX-containing cores are strongly influenced by the nuclear properties of the plutonium isotopes. Optimized MOX fuel assembly designs for PWRs currently use up to three types of MOX fuel rods having different plutonium contents with natural uranium or uranium tailings as carrier material but without burnable absorbers. The MOX fuel assembly designs for BWRs use four to six rod types with different plutonium contents and Gd[sub 2]O[sub 3]/UO[sub 2] burnable absorber rods. Both the PWR and the BWR designs attain good burnup equivalence and compatibility with uranium fuel assemblies. High flexibility exists in the loading schemes relative to the position and number of MOX fuel assemblies in the reloads and in the core as a whole. The Siemens experience with MOX fuel assemblies is based on the insertion of 318 MOX fuel assemblies in eight PWRs and 168 in BWRs and pressurized heavy water reactors so far. The primary operating results include information on the cycle length, power distribution, reactivity coefficients, and control rod worth of cores containing MOX fuel assemblies.

  6. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104). [PWR; BWR

    SciTech Connect

    Kress, T. S.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time.

  7. Decontamination as a precursor to decommissioning. Status report Task 2: process evaluation. [PWR; BWR

    SciTech Connect

    Divine, J.R.; Woodruff, E.M.; McPartland, S.A.; Zima, G.E.

    1983-05-01

    As part of the US Nuclear Regulatory Commission's program to reduce occupational exposure and waste volumes, the Pacific Northwest Laboratory is studying decontamination as a precursor to decommissioning. Eleven processes or solvents were examined for their behavior in decontaminating BWR carbon steel samples. The solvents included NS-1, a proprietary solvent of Dow Chemical Corporation, designed for BWR use, and AP-Citrox, a well-known, two-step process designed for PWR stainless steel; it was used to provide a reference for later comparison to other systems and processes. The decontamination factors observed in the tests performed in a small laboratory scale recirculating loop ranged from about 1 (no effect) to 222 (about 99.6% of the initial activity removed. Coordinated corrosion measurements were made using twelve chemical solvents and eight metal alloys found in a range of reactor types.

  8. Decay heat removal systems: design criteria and options. [PWR; BWR

    SciTech Connect

    Berry, D.L.

    1980-01-01

    Design criteria and alternate decay heat removal system concepts which have evolved in several different countries throughout the world were compared. The conclusion was reached that the best way to improve the reliability of pressurized water reactor (PWR) decay heat removal is first to focus on improving the reliability of the auxiliary feedwater and high pressure injection systems to cope with certain loss of feedwater transients and small loss of coolant accidents and then to assess how well these systems can handle special emergencies (e.g., sabotage, earthquake, airplane crash). For boiling water reactors (BWRs), it was concluded that emphasis should be placed first on improving the reliability of the residual heat removal and high pressure service water systems to cope with a loss of suppression pool cooling following a loss of feedwater transient and then to assess how well these systems can handle special emergencies. It was found that, for both PWRs and BWRs, a design objective for alternate decay heat removal systems should be at least an order of magnitude reduction in core meltdown probability.

  9. Code System for PWR & BWR Multicompartment Containment Analysis, Versions MOD5

    1999-06-02

    CONTEMPT4/MOD6 describes the response of multicompartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user-supplied descriptions of compartments,more » inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. CONTEMPT4/MOD6 also provides analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation to accommodate degraded core type accidents.« less

  10. Modelling of molten fuel/concrete interactions. [PWR; BWR

    SciTech Connect

    Muir, J. F.; Benjamin, A. S.

    1980-01-01

    A computer program modelling the interaction between molten core materials and structural concrete (CORCON) is being developed to provide quantitative estimates of fuel-melt accident consequences suitable for risk assessment of light water reactors. The principal features of CORCON are reviewed. Models developed for the principal interaction phenomena, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. Alternative models for the controlling phenomenon, heat transfer from the molten pool to the surrounding concrete, are presented. These models, formulated in conjunction with the development of CORCON, are characterized by the presence or absence of either a gas film or viscous layer of molten concrete at the melt/concrete interface. Predictions of heat transfer based on these models compare favorably with available experimental data.

  11. Relationship of fire protection research to plant safety. [PWR; BWR

    SciTech Connect

    Berry, D.L.

    1983-01-01

    For several years, Sandia National Laboratories has been responsible for numerous tests of fire protection systems and concepts. Tests of fire retardant cables, cable coatings, cable tray covers, penetration seals, fire barriers, and spatial separation have been reported and summarized. Other tests involving the effectiveness of suppression systems and the vulnerability of electrical cabinets have been completed with reports in preparation. The following questions constitute the central theme of current fire research by Sandia and the NRC: under what conditions is spatial separation of redundant safety systems adequate; what are the temperature, smoke, humidity, or corrosive vapor damage thresholds of cable and safety equipment exposed to fire or suppression activities; what is the safety significance of fires involving control room cabinets or remote shutdown panels; and what is the relative importance of fire to nuclear power plant safety, as compared to other types of anticipated or postulated accidents. Evidence of why these questions seem important and a description of work being undertaken to address each question are reviewed in the following paragraphs.

  12. Remote Gamma Scanning System for Characterization of BWR and PWR Fuel Rod Sections

    SciTech Connect

    Crowell, Shannon L.; Alzheimer, James M.

    2011-08-08

    Sometimes challenges with the design and deployment of automated equipment in remote environments deals more with the constraints imposed by the remote environment than it does with the details of the automation. This paper discusses the development of a scanning system used to provide gamma radiation profiles of irradiated fuel rod segments. The system needed the capability to provide axial scans of cut segments of BWR and PWR fuel rods. The scanning location is A-Cell at the Radiochemical Processing Laboratory (RPL) at the Hanford site in Washington State. The criteria for the scanning equipment included axial scanning increments of a tenth of an inch or less, ability to scan fuel rods with diameters ranging from 3/8 inch to 5/8 inch in diameter, and fuel rod segments up to seven feet in length. Constraints imposed by the environment included having the gamma detector and operator controls on the outside of the hot cell and the scanning hardware on the inside of the hot cell. This entailed getting a narrow, collimated beam of radiation from the fuel rod to the detector on the outside of the hot cell while minimizing the radiation exposure caused by openings for the wires and cables traversing the hot cell walls. Setup and operation of all of the in-cell hardware needed to accommodate limited access ports and use of hot cell manipulators. The radiation levels inside the cell also imposed constraints on the materials used.

  13. Radiation dose rates from commercial PWR and BWR spent fuel elements

    SciTech Connect

    Willingham, C.E.

    1981-10-01

    Data on measurements of gamma dose rates from commercial reactor spent fuel were collected, and documented calculated gamma dose rates were reviewed. As part of this study, the gamma dose rate from spent fuel was estimated, using computational techniques similar to previous investigations into this problem. Comparison of the measured and calculated dose rates provided a recommended dose rate in air versus distance curve for PWR spent fuel.

  14. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    SciTech Connect

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A.

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  15. Standard- and extended-burnup PWR (pressurized-water reactor) and BWR (boiling-water reactor) reactor models for the ORIGEN2 computer code

    SciTech Connect

    Ludwig, S.B.; Renier, J.P.

    1989-12-01

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs.

  16. Pre-Phase 1 Aging Assessment of the BWR and PWR Accumulators

    SciTech Connect

    Buckely, G. D.

    1995-08-01

    Accumulators are important components used in many systems at commercial boiling water reactors (BWRs) and pressurized water reactors in the United States. The accumulators are vessels attached to fluid systems to provide 1) a limited backup source of stored fluid energy for hydraulic/pneumatic mechanical equipment, 2) a damping effect on pressure pulses in fluid systems, and 3) a volume of fluid to be injected passively into a fluid system. Accumulators contain a gas that is compressed or expanded as the fluid from the system enters or exits the accumulator. The gas and fluid in accumulators are usually separated from each other by a piston or bladder. In support of the U.S. Nuclear Regulatory Commission's Nuclear Aging Research Program (NPAR), the Pacific Northwest Laboratory conducted an analysis of available industry databases to determine if accumulator components already had been studied in other NPAR assessments and to evaluate each accumulator type for applicable aging issues. The results of this preliminary study indicate that two critical uses of accumulators have been previously evaluated by the NPAR program. NUREGICR-5699, Aging and Service Wear of Control Rod Drive Mechanisms for BUT Nuclear Plants (Greene 199 I), identified two hydraulic control unit components subject to aging failures: accumulator nitrogen-charging cartridge valves and the scram water accumulator. In addition, NUREGICR-6001, Aging Assessment of BWR Standby Liquid Control Systems (Buckley et al. 1992), identified two predominant aging-related accumulator failures that result in a loss of the nitrogen blanket pressure: (charging) valve wear and failure of the gas bladder. The present study has identified five prevalent aging-related accumulator failures: rupture of the accumulator bladder separation of the metal disc from the bottom of the bladder leakage of the gas from the charging valve leakage past the safety injection tank manway cover gasket leakage past O-rings. An additional

  17. Main-coolant-pump shaft-seal guidelines. Volume 3. Specification guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria to aid in the generation of procurement specifications for Main Coolant Pump Shaft Seals. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies, a review of pump and shaft seal literature and discussions with pump and seal designers. This report is preliminary in nature and could be expanded and finalized subsequent to completion of further design, test and evaluation efforts.

  18. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  19. Iodine volatility. [PWR; BWR

    SciTech Connect

    Beahm, E.C.; Shockley, W.E.

    1984-01-01

    The ultimate aim of this program is to couple experimental aqueous iodine volatilities to a fission product release model. Iodine partition coefficients, for inorganic iodine, have been measured during hydrolysis and radiolysis. The hydrolysis experiments have illustrated the importance of reaction time on iodine volatility. However, radiolysis effects can override hydrolysis in determining iodine volatility. In addition, silver metal in radiolysis samples can react to form silver iodide accompanied by a decrease in iodine volatility. Experimental data are now being coupled to an iodine transport and release model that was developed in the Federal Republic of Germany.

  20. BWR internal cracking issues

    SciTech Connect

    Carpenter, C.E. Jr.; Lund, A.L.

    1999-07-01

    The regulatory issues associated with cracking of boiling water reactor (BWR) internals is being addressed by the Nuclear Regulatory Commission (NRC) staff and is the subject of a voluntary industry initiative. The lessons learned from this effort will be applied to pressurized water reactor (PWR) internals cracking issues.

  1. Kohonen mapping of the crack growth under fatigue loading conditions of stainless steels in BWR environments and of nickel alloys in PWR environments

    NASA Astrophysics Data System (ADS)

    Urquidi-Macdonald, Mirna

    2008-09-01

    In this study, crack growth rate data under fatigue loading conditions generated by Argonne National Laboratories and published in 2006 were analyzed [O.K. Chopra, B. Alexandreanu, E.E. Gruber, R.S. Daum, W.J. Shack, Argonne National Laboratory, NUREG CR 6891-series ANL 04/20, Crack Growth Rates of Austenitic Stainless Steel Weld Heat Affected Zone in BWR Environments, January, 2006; B. Alexandreanu, O.K. Chopra, H.M. Chung, E.E. Gruber, W.K. Soppet, R.W. Strain, W.J. Shack, Environmentally Assisted Cracking in Light Water Reactors, vol. 34 in the NUREG/CR-4667 series annual report of Argonne National Laboratory program studies for Calendar (Annual Report 2003). Manuscript Completed: May 2005, Date Published: May 2006], and reported by DoE [B. Alexandreanu, O.K. Chopra, W.J. Shack, S. Crane, H.J. Gonzalez, NRC, Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments, NUREG/CR-6964, May 2008]. The data collected were measured on austenitic stainless steels in BWR (boiling water reactor) environments and on nickel alloys in PWR (pressurized water reactor) environments. The data collected contained information on material composition, temperature, conductivity of the environment, oxygen concentration, irradiated sample information, weld information, electrochemical potential, load ratio, rise time, hydrogen concentration, hold time, down time, maximum stress intensity factor ( Kmax), stress intensity range (Δ Kmax), crack length, and crack growth rates (CGR). Each position on that Kohonen map is called a cell. A Kohonen map clusters vectors of information by 'similarities.' Vectors of information were formed using the metal composition, followed by the environmental conditions used in each experiments, and finally followed by the crack growth rate (CGR) measured when a sample of pre-cracked metal is set in an environment and the sample is cyclically loaded. Accordingly

  2. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    NASA Astrophysics Data System (ADS)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  3. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    SciTech Connect

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  4. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    SciTech Connect

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  5. Review of inservice inspections of greased tendons in prestressed-concrete containments. [PWR; BWR

    SciTech Connect

    Dougan, J.R.; Ashar, H.

    1983-01-01

    Prestressed-concrete containments in the United States using greased prestressing tendons are inspected periodically to ensure structural integrity and to identify and correct problem areas before they become critical. An analysis of the available utility inspection data and an evaluation of the current and proposed guidelines were conducted to provide a measure of the reliability of the inspection process. Comments from utility and industry personnel were factored into the analysis. The results indicated that the majority of the few incidences of problems or abnormalities which occurred were minor in nature and did not threaten the structural integrity of the containment.

  6. PNL technical review of pressurized thermal-shock issues. [PWR

    SciTech Connect

    Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J.; Simonen, E.P.; Simonen, F.A.; Stevens, D.L.; Taylor, T.T.

    1982-07-01

    Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.

  7. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    NASA Astrophysics Data System (ADS)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  8. Master information data acquisition system. [PWR; BWR

    SciTech Connect

    Stratton, R.C.

    1981-06-01

    MIDAS provides real-time work and component status as a function of the Work Control Log (WCL). Status is maintained and tracked with regard to Tag-Out status, Out-of-Service status, and document status. Component integration and technical and safety information is provided by the MIDAS Component Index (MCI). This information is provided directly to the user upon request. This information is also provided directly to the WCL as a function of component input to the WCL document prior to the documents release for execution.

  9. Impact analyses after pipe rupture. [PWR; BWR

    SciTech Connect

    Chun, R.C.; Chuang, T.Y.

    1983-12-13

    Two of the French pipe whip experiments are reproduced with the computer code WIPS. The WIPS results are in good agreement with the experimental data and the French computer code TEDEL. This justifies the use of its pipe element in conjunction with its U-bar element in a simplified method of impact analyses.

  10. Fuel performance under normal PWR conditions: A review of relevant experimental results and models

    NASA Astrophysics Data System (ADS)

    Charles, M.; Lemaignan, C.

    1992-06-01

    Experiments conducted at Grenoble (CEA/DRN) over the past 20 years in the field of nuclear fuel behaviour are reviewed. Of particular concern is the need to achieve a comprehensive understanding of and subsequently overcome the limitations associated with high burnup and load-following conditions (pellet-cladding interaction (PCI), fission gas release (FGR), water-side corrosion). A general view is given of the organization of research work as well as some experimental details (irradiation, postirradiation examination — PIE). Based on various experimental programmes (Cyrano, Medicis, Anemone, Furet, Tango, Contact, Cansar, Hatac, Flog, Decor), the main contributions of the thermomechanical behaviour of a PWR fuel rod are described: thermal conductivity, in-pile densification, swelling, fission gas release in steady state and moderate transient conditions, gap thermal conductance, formation of primary and secondary ridges under PCI conditions. Specific programmes (Gdgrif, Thermox, Grimox) are devoted to the behaviour of particular fuels (gadolinia-bearing fuel, MOX fuel). Moreover, microstructure-based studies have been undertaken on fission gas release (fine analysis of the bubble population inside irradiated fuel samples), and on cladding behaviour (PCI related studies on stress-corrosion cracking (SCO, irradiation effects on zircaloy microstructure).

  11. Preliminary study on direct recycling of spent BWR fuel in BWR system

    NASA Astrophysics Data System (ADS)

    Waris, A.; Sumbono, Prayudhatama, Dythia; Novitrian, Su'ud, Zaki

    2012-06-01

    Spent fuel management is considered to be one of the main problems in energy nuclear utilization. Recycling after reprocessing is one of the options for dealing with nuclear reactor spent fuel. Reprocessing is very costly and needs remote handling since spent fuel is very hazard high level waste. On top of that, only a small number of countries can manage a reprocessing plant. If country likes Indonesia decide to "go nuclear", it should find another way to deal with the nuclear spent fuel. Korea has proposed the DUPIC (Direct Utilization of Spent PWR fuel In CANDU) concept. Nevertheless, DUPIC concept requires two types of nuclear power plants, i.e., pressurized water reactor (PWR) and CANadian Deuterium Uranium reactor (CANDU). In this study, we evaluate a scheme of direct recycling of spent BWR fuel in BWR system, under the concept that we have called as a SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario. Several spent BWR fuel compositions in loaded BWR fuel has been evaluated to achieve the criticality of reactor.

  12. Peer Review of NRC Standardized Plant Analysis Risk Models

    SciTech Connect

    Anthony Koonce; James Knudsen; Robert Buell

    2011-03-01

    The Nuclear Regulatory Commission (NRC) Standardized Plant Analysis Risk (SPAR) Models underwent a Peer Review using ASME PRA standard (Addendum C) as endorsed by NRC in Regulatory Guide (RG) 1.200. The review was performed by a mix of industry probabilistic risk analysis (PRA) experts and NRC PRA experts. Representative SPAR models, one PWR and one BWR, were reviewed against Capability Category I of the ASME PRA standard. Capability Category I was selected as the basis for review due to the specific uses/applications of the SPAR models. The BWR SPAR model was reviewed against 331 ASME PRA Standard Supporting Requirements; however, based on the Capability Category I level of review and the absence of internal flooding and containment performance (LERF) logic only 216 requirements were determined to be applicable. Based on the review, the BWR SPAR model met 139 of the 216 supporting requirements. The review also generated 200 findings or suggestions. Of these 200 findings and suggestions 142 were findings and 58 were suggestions. The PWR SPAR model was also evaluated against the same 331 ASME PRA Standard Supporting Requirements. Of these requirements only 215 were deemed appropriate for the review (for the same reason as noted for the BWR). The PWR review determined that 125 of the 215 supporting requirements met Capability Category I or greater. The review identified 101 findings or suggestions (76 findings and 25 suggestions). These findings or suggestions were developed to identify areas where SPAR models could be enhanced. A process to prioritize and incorporate the findings/suggestions supporting requirements into the SPAR models is being developed. The prioritization process focuses on those findings that will enhance the accuracy, completeness and usability of the SPAR models.

  13. Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

    SciTech Connect

    Wagner, J.C.

    2000-10-12

    Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and

  14. Prediction of quench and rewet temperatures. [PWR; BWR

    SciTech Connect

    Gunnerson, F. S.

    1980-01-01

    Many postulated nuclear reactor accidents result in high-temperature dryout or film boiling within the nuclear core. In order to mitigate potential fuel rod damage or rod failure, safe or lower fuel rod temperatures must be reestablished by promoting coolant/cladding contact. This process is commonly referred to as quenching or rewetting, and often, these terms are not differentiated. All theoretical predictions of the cooling process by various models based on single or multidimensional analytical and numerical studies require a knowledge of either the quenching or the rewetting temperature. The purpose of this paper is to define quench and rewet temperatures and present a method whereby each may be estimated.

  15. HSST pressurized-thermal-shock experiment, PTSE-1. [PWR; BWR

    SciTech Connect

    Bryan, R.H.; Bass, B.R.; Robinson, G.C.; Merkle, J.G.; Whitman, G.D.; Pugh, C.E.

    1984-01-01

    The first pressurized-thermal-shock experiment (PTSE-1) in the Heavy-Section Steel Technology (HSST) Program is the most recent of a long successtion of fracture-mechanics experiments that are on a scale that allows important aspects of fracture behavior of reactor pressure vessels to be simulated. Such experiments are the means by which theoretical models of fracture behavior can be evaluated for possible aplication to fracture analysis of vessels in nuclear plants. The principal issues of concern in the pressurized-thermal-shock experiments are: (1) warm prestressing phenomena, (2) crack propagation from brittle to ductile regions, (3) transient crack stabilization in ductile regions, and (4) crack shape changes in bimetallic zones of clad vessels. PTSE-1 was designed to investigate the first three issues under conditions relevant to a flawed reactor vessel during an overcooling accident.

  16. Fuel performance annual report for 1981. [PWR; BWR

    SciTech Connect

    Bailey, W.J.; Tokar, M.

    1982-12-01

    This annual report, the fourth in a series, provides a brief description of fuel performance during 1981 in commercial nuclear power plants. Brief summaries of fuel operating experience, fuel problems, fuel design changes and fuel surveillance programs, and high-burnup fuel experience are provided. References to additional, more detailed information and related NRC evaluations are included.

  17. Stress-corrosion cracking in BWR and PWR piping

    SciTech Connect

    Weeks, R.W.

    1983-07-01

    Intergranular stress-corrosion cracking of weld-sensitized wrought stainless steel piping has been an increasingly ubiquitous and expensive problem in boiling-water reactors over the last decade. In recent months, numerous cracks have been found, even in large-diameter lines. A number of potential remedies have been developed. These are directed at providing more resistant materials, reducing weld-induced stresses, or improving the water chemistry. The potential remedies are discussed, along with the capabilities of ultrasonic testing to find and size the cracks and related safety issues. The problem has been much less severe to date in pressurized-water reactors, reflecting the use of different materials and much lower coolant oxygen levels.

  18. Fire Protection Research Program at Sandia Laboratories. [BWR; PWR

    SciTech Connect

    Klamerus, L.J.

    1980-01-01

    Sandia Laboratories is executing a program for the Nuclear Regulatory Commission to provide data needed for confirmation of the suitability of current design standards and regulatory guides for fire protection and control in water reactor power plants. This paper summarizes the activities of this ongoing program through December 1979. Characterization of electrically initiated fires revealed a margin of safety in the separation criteria of Regulatory Guide 1.75 for such fires in IEEE-383 qualified cable. However, tests confirmed that these guidelines and standards are not sufficient, in themselves, to protect against exposure fires. This paper describes both small and full scale tests to assess the adequacy of fire retardant coatings and full scale tests on fire shields to determine their effectiveness. It also describes full scale tests to determine the effects of walls and ceilings on fire propagation between cable trays.

  19. Planning guidance for nuclear-power-plant decontamination. [PWR; BWR

    SciTech Connect

    Munson, L.F.; Divine, J.R.; Martin, J.B.

    1983-06-01

    Direct and indirect costs of decontamination are considered in the benefit-cost analysis. A generic form of the benefit-cost ratio is evaluated in monetary and nonmonetary terms, and values of dollar per man-rem are cited. Federal and state agencies that may have jurisiction over various aspects of decontamination and waste disposal activities are identified. Methods of decontamination, their general effectiveness, and the advantages and disadvantages of each are outlined. Dilute or concentrated chemical solutions are usually used in-situ to dissolve the contamination layer and a thin layer of the underlying substrate. Electrochemical techniques are generally limited to components but show high decontamination effectiveness with uniform corrosion. Mechanical agents are particularly appropriate for certain out-of-system surfaces and disassembled parts. These processes are catagorized and specific concerns are discussed. The treatment, storage, and disposal or discharge or discharge of liquid, gaseous, and solid wastes generated during the decontamination process are discussed. Radioactive and other hazardous chemical wastes are considered. The monitoring, treatment, and control of radioactive and nonradioactive effluents, from both routine operations and possible accidents, are discussed. Protecting the health and safety of personnel onsite during decontamination is of prime importance and should be considered in each facet of the decontamination process. The radiation protection philosophy of reducing exposure to levels as low as reasonably achievable should be stressed. These issues are discussed.

  20. End effects on elbows subjected to moment loadings. [PWR; BWR

    SciTech Connect

    Rodabaugh, E.C.; Moore, S.E.

    1982-01-01

    So-called end effects for moment loadings on short-radius and long-radius butt welding elbows of various arc lengths are investigated with a view toward providing more accurate design formulas for critical piping systems. Data developed in this study, along with published information, were used to develop relatively simple design equations for elbows attached at both ends to long sections of straight pipe. These formulas are the basis for an alternate ASME Code procedure for evaluating the bending moment stresses in Class 1 nuclear piping (ASME Code Case N-319). The more complicated problems of elbows with other end conditions, e.g., flanges at one or both ends, are also considered. Comparisons of recently published experimental and theoretical studies with current industrial code design rules for these situations indicate that these rules also need to be improved.

  1. Interfacial transfer in annular dispersed flow. [PWR; BWR

    SciTech Connect

    Ishii, M.; Kataoka, I.

    1982-01-01

    The interfacial drag, droplet entrainment, droplet deposition and droplet-size distributions are important for detailed mechanistic modeling of annular dispersed two-phase flow. In view of this, recently developed correlations for these parameters are presented and discussed in this paper. The onset of droplet entrainment significantly alters the mechanisms of mass, momentum, and energy transfer between the film and gas core flow as well as the transfer between the two-phase mixture and the wall. By assuming the roll wave entrainment mechanism, the correlations for the amount of entrained droplet as well as for the droplet-size distribution have been obtained from a simple model in collaboration with a large number of data. Then the rate equations for entrainment and deposition have been developed. The drag correlations relevant to the droplet transfer is also presented. The comparison of the correlations to various data show satisfactory agreement.

  2. Hydrodynamics of annular-dispersed flow. [PWR; BWR

    SciTech Connect

    Ishii, M.; Kataoka, I.

    1982-01-01

    The interfacial drag, droplet entrainment, and droplet size distributions are important for detailed mechanistic modeling of annular dispersed two-phase flow. In view of this, recently developed correlations for these parameters are presented and discussed in this paper. The drag correlations for multiple fluid particle systems have been developed from a similarity hypothesis based on the mixture viscosity model. The results show that the drag coefficient depends on the particle Reynolds number and droplet concentration. The onset on droplet entrainment significantly alters the mechanisms of mass, momentum, and energy transfer between the film and gas core flow as well as the transfer between the two-phase mixture and the wall. By assuming the roll wave entrainment mechanism, the correlations for the amount of entrained droplet as well as for the droplet size distribution have been obtained from a simple model in collaboration with a large number of data.

  3. Critical discharge of initially subcooled water through slits. [PWR; BWR

    SciTech Connect

    Amos, C N; Schrock, V E

    1983-09-01

    This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits. The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure. Two new analytical models, which allow for the generation of a metastable liquid phase, are developed. Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model.

  4. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOEpatents

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  5. BWR AXIAL PROFILE

    SciTech Connect

    J. Huffer

    2004-09-28

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I.

  6. The role of Hydrogen and Creep in Intergranular Stress Corrosion Cracking of Alloy 600 and Alloy 690 in PWR Primary Water Environments ? a Review

    SciTech Connect

    Rebak, R B; Hua, F H

    2004-07-12

    Intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of Alloy 600 in PWR steam generator environment has been extensively studied for over 30 years without rendering a clear understanding of the essential mechanisms. The lack of understanding of the IGSCC mechanism is due to a complex interaction of numerous variables such as microstructure, thermomechanical processing, strain rate, water chemistry and electrochemical potential. Hydrogen plays an important role in all these variables. The complexity, however, significantly hinders a clearer and more fundamental understanding of the mechanism of hydrogen in enhancing intergranular cracking via whatever mechanism. In this work, an attempt is made to review the role of hydrogen based on the current understanding of grain boundary structure and chemistry and intergranular fracture of nickel alloys, effect of hydrogen on electrochemical behavior of Alloy 600 and Alloy 690 (e.g. the passive film stability, polarization behavior and open-circuit potential) and effect of hydrogen on PWSCC behavior of Alloy 600 and Alloy 690. Mechanistic studies on the PWSCC are briefly reviewed. It is concluded that further studies on the role of hydrogen on intergranular cracking in both inert and primary side environments are needed. These studies should focus on the correlation of the results obtained at different laboratories by different methods on materials with different metallurgical and chemical parameters.

  7. Review of a Metdology for Tracking the Neutron Exposure of Pwr Pressure Vessels during the License Renewal Period

    NASA Astrophysics Data System (ADS)

    Anderson, Stan L.; Fero, Arnold H.; Roberts, George K.

    2003-06-01

    The neutron fluence associated with each material in the pressure vessel beltline region is determined on a plant specific basis at each surveillance capsule withdrawal. Based on an assumed mode of operation, fluence projections to account for future operation are then made for use in vessel integrity evaluations. The applicability of these assumed projections is normally verified and updated, if necessary, at each subsequent surveillance capsule withdrawal. However, following the last scheduled withdrawal of a surveillance capsule, there is generally no formal mechanism in place to assure that fluence projections for the remainder of plant operating lifetime remain valid. This paper provides a review of a methodology that can be efficiently used in conjunction with future fuel loading patterns or on-line core power distribution monitoring systems to track the actual fluence accrued by each of the pressure vessel beltline materials in the operating period following the last capsule withdrawal.

  8. Synergistic Failure of BWR Internals

    SciTech Connect

    Ware, Arthur Gates; Chang, T-Y

    1999-10-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.

  9. The electrochemistry in 316SS crevices exposed to PWR-relevant conditions

    NASA Astrophysics Data System (ADS)

    Vankeerberghen, M.; Weyns, G.; Gavrilov, S.; Henshaw, J.; Deconinck, J.

    2009-04-01

    The chemical and electrochemical conditions within a crevice of Type 316 stainless steel in boric acid-lithium hydroxide solutions under PWR-relevant conditions were modelled with a computational electrochemistry code. The influence of various variables: dissolved hydrogen, boric acid, lithium hydroxide concentration, crevice length, and radiation dose rate was studied. It was found with the model that 25 ccH 2/kg (STP) was sufficient to remain below an electrode potential of -230 mV she, commonly accepted sufficient to prevent stress corrosion cracking under BWR conditions. In a PWR plant various operational B-Li cycles are possible but it was found that the choice of the cycle did not significantly influence the model results. It was also found that a hydrogen level of 50 ccH 2/kg (STP) would be needed to avoid substantial lowering of the pH inside a crevice.

  10. PWR depressurization analyses

    SciTech Connect

    Brownson, D.A.; Dobbe, C.A.; Knudson, D.L.

    1992-12-31

    Early containment failure resulting from direct containment heating (DCH) has been identified as a potential contributor to the risk of operating a pressurized water reactor (PWR). One important factor needed to evaluate the contribution of DCH to risk is the conditional probability that, given a core melt, the primary system will be at high pressure when the reactor vessel lower head fails. Two mechanisms that could reduce the pressure during a station blackout core melt accident are discussed. First, natural circulation in the reactor coolant system (RCS) could cause a temperature-induced failure of the RCS pressure boundary, which could result in unintentional (without operator action) depressurization. Second, plant operators could open relief valves in an attempt to intentionally depressurize the RCS prior to. lower head failure. Results from analytical studies of these two depressurization mechanisms for select PWRs are presented.

  11. PWR depressurization analyses

    SciTech Connect

    Brownson, D.A.; Dobbe, C.A.; Knudson, D.L.

    1992-01-01

    Early containment failure resulting from direct containment heating (DCH) has been identified as a potential contributor to the risk of operating a pressurized water reactor (PWR). One important factor needed to evaluate the contribution of DCH to risk is the conditional probability that, given a core melt, the primary system will be at high pressure when the reactor vessel lower head fails. Two mechanisms that could reduce the pressure during a station blackout core melt accident are discussed. First, natural circulation in the reactor coolant system (RCS) could cause a temperature-induced failure of the RCS pressure boundary, which could result in unintentional (without operator action) depressurization. Second, plant operators could open relief valves in an attempt to intentionally depressurize the RCS prior to. lower head failure. Results from analytical studies of these two depressurization mechanisms for select PWRs are presented.

  12. Decontamination of BWR fuel bundles

    SciTech Connect

    Ocken, H.

    1988-01-01

    Decontamination of individual systems in operating reactors, such as recirculation piping in boiling water reactors (BWRs) and steam generators in pressurized water reactors, is becoming an accepted technique to reduce radiation fields and occupational radiation exposure. Because a significant inventory of radioactivity resides on the reactor core, a longer term goal is to effect full plant decontamination with the fuel in place. Full plant decontamination has proved effective in CANDU and steam-generating heavy water reactor plants, but only recently have US plants begun to consider seriously the merits of such an approach. Clearly, a first step is to show that exposure to commercial decontamination solvents of highly irradiated core components will not induce any adverse effects. This paper describes a study of the application of the LOMI and CANDECON solvents to three-cycle discharged fuel bundles from the Quad Cities-2 BWR. Highly irradiated stainless steel specimens cut from a section of a LaCrosse BWR control blade also were decontaminated at the same time as the fuel bundles. CANDECON was selected as being representative of dilute chelant process and LOMI as representative of more strongly reducing processes. Both processes were preceded by the application of an oxidizing alkaline permanganate (AP) oxidizing step to help dissolve chromium.

  13. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    SciTech Connect

    Wagner, J.C.; Parks, C.V.

    2000-09-01

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration ({approximately}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ({le} 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of

  14. BWR Assembly Optimization for Minor Actinide Recycling

    SciTech Connect

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  15. 44 BWR Waste Package Loading Curve Evaluation

    SciTech Connect

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  16. Simplified compact containment BWR plant

    SciTech Connect

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-07-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  17. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  18. BWR Refill-Reflood Program. Final report

    SciTech Connect

    Myers, L L

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests.

  19. BWR pipe crack remedies evaluation

    SciTech Connect

    Shack, W.J.; Kassner, T.F.; Maiya, P.S.; Park, J.Y.; Ruther, W.E.

    1986-10-01

    This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-craking (SCC) susceptibility of Types 304, 316NG, and 347 stainless (SS); (b) fracture-mechanics crack-growth-rate measurements on these materials and weld overlay specimens in different environments; and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on Types 304 and 316NG SS under crevice and non-crevice conditions in 289/sup 0/C water containing 0.25 ppM dissolved oxygen with low sulfate concentrations indicate that SCC initiates at very low strains (<3%) in the nuclear grade material. Crack growth measurements on fracture-mechanics-type specimens, under low-frequency cyclic loading, show that the Type 316NG steel cracks at a somewhat lower rate (approx.40%) than sensitized Type 304 SS in an impurity environment with 0.25 ppM dissolved-oxygen; however, the latter material stops cracking when sulfate is removed from the water. Crack growth in both materials ceases under simulated hydrogen-water chemistry conditions (<5 ppB oxygen) even with 100 ppB sulfate present in the water. An unexpected result was obtained in the test on a weld overlay specimen in the impurity environment, viz., the crack grew to the overlay interface at a nominal rate, branched at 90/sup 0/ in both directions, and then grew at high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones.

  20. PWR fuel behavior: lessons learned from LOFT. [PWR

    SciTech Connect

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior.

  1. Assessment of two BWR accident management strategies

    SciTech Connect

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs.

  2. BWR stability analysis at Brookhaven National Laboratory

    SciTech Connect

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-12-31

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  3. Institutional implications of establishing safety goals for nuclear power plants. [PWR; BWR

    SciTech Connect

    Morris, F.A.; Hooper, R.L.

    1983-07-01

    The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants. The report emphasizes one particular category of institutional problems: the possible use of safety goals as a basis for legal challenges to NRC actions, and the resolution of such challenges by the courts. Three types of legal issues are identified and analyzed. These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof. Second is the particular formulation of goals. Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism. Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.

  4. Correlation and spectral measurements of fluctuating pressures and velocities in annular turbulent flow. [PWR; BWR

    SciTech Connect

    Wilson, R.J.; Jones, B.G.; Roy, R.P.

    1980-02-01

    An experimental study of the fluctuating velocity field, the fluctuating static wall pressure and the in-stream fluctuating static pressure in an annular turbulent air flow system with a radius ratio of 4.314 has been conducted. The study included direct measurements of the mean velocity profile, turbulent velocity field; fluctuating static wall pressure and in-stream fluctuating static pressure from which the statistical values of the turbulent intensity levels, power spectral densities of the turbulent quantities, the cross-correlation between the fluctuating static wall pressure and the fluctuating static pressure in the core region of the flow and the cross-correlation between the fluctuating static wall pressure and the fluctuating velocity field in the core region of the flow were obtained.

  5. Heat transfer to water from a vertical tube bundle under natural-circulation conditions. [PWR; BWR

    SciTech Connect

    Gruszczynski, M.J.; Viskanta, R.

    1983-01-01

    The natural circulation heat transfer data for longitudinal flow of water outside a vertical rod bundle are needed for developing correlations which can be used in best estimate computer codes to model thermal-hydraulic behavior of nuclear reactor cores under accident or shutdown conditions. The heat transfer coefficient between the fuel rod surface and the coolant is the key parameter required to predict the fuel temperature. Because of the absence of the required heat transfer coefficient data base under natural circulation conditions, experiments have been performed in a natural circulation loop. A seven-tube bundle having a pitch-to-diameter ratio of 1.25 was used as a test heat exchanger. A circulating flow was established in the loop, because of buoyancy differences between its two vertical legs. Steady-state and transient heat transfer measurements have been made over as wide a range of thermal conditions as possible with the system. Steady state heat transfer data were correlated in terms of relevant dimensionless parameters. Empirical correlations for the average Nusselt number, in terms of Reynolds number, Rayleigh number and the ratio of Grashof to Reynolds number are given.

  6. One-dimensional time-dependent debris bed model. [PWR; BWR

    SciTech Connect

    Gorham-Bergeron, E.

    1982-01-01

    The dryout process is described for a particle bed using a time-dependent one-dimensional porous bed model. The model is based on momentum, energy and mass conservation equations for separated flow. The model is applicable to the case in which capillary forces can be neglected. For the case in which only laminar flow is considered exact algebraic solutions to the equations can be obtained. These are presented. Distinct regimes for the parameterized solutions can be identified and associated with moving fronts in the bed. Extension to the full turbulent and laminar equations is made with the aid of insights gained from solution of the laminar case. Comparison with recent experimental results and theoretical predictions is made. The model is seen to encompass and extend the theoretical models. It suggests additional experiments.

  7. Source terms: an investigation of uncertainties, magnitudes, and recommendations for research. [PWR; BWR

    SciTech Connect

    Levine, S.; Kaiser, G. D.; Arcieri, W. C.; Firstenberg, H.; Fulford, P. J.; Lam, P. S.; Ritzman, R. L.; Schmidt, E. R.

    1982-03-01

    The purpose of this document is to assess the state of knowledge and expert opinions that exist about fission product source terms from potential nuclear power plant accidents. This is so that recommendations can be made for research and analyses which have the potential to reduce the uncertainties in these estimated source terms and to derive improved methods for predicting their magnitudes. The main reasons for writing this report are to indicate the major uncertainties involved in defining realistic source terms that could arise from severe reactor accidents, to determine which factors would have the most significant impact on public risks and emergency planning, and to suggest research and analyses that could result in the reduction of these uncertainties. Source terms used in the conventional consequence calculations in the licensing process are not explicitly addressed.

  8. Application of the TEMPEST computer code for simulating hydrogen distribution in model containment structures. [PWR; BWR

    SciTech Connect

    Trent, D.S.; Eyler, L.L.

    1982-09-01

    In this study several aspects of simulating hydrogen distribution in geometric configurations relevant to reactor containment structures were investigated using the TEMPEST computer code. Of particular interest was the performance of the TEMPEST turbulence model in a density-stratified environment. Computed results illustrated that the TEMPEST numerical procedures predicted the measured phenomena with good accuracy under a variety of conditions and that the turbulence model used is a viable approach in complex turbulent flow simulation.

  9. Experiment data report for Multirod Burst Test (MRBT) bundle B-6. [PWR; BWR

    SciTech Connect

    Chapman, R H; Longest, A W; Crowley, J L

    1984-07-01

    A reference source of MRBT bundle B-6 test data is presented with minimum interpretation. The primary objective of this 8 x 8 multirod burst test was to investigate cladding deformation in the alpha-plus-beta-Zircaloy temperature range under simulated light-water-reactor (LWR) loss-of-coolant accident (LOCA) conditions. B-6 test conditions simulated the adiabatic heatup (reheat) phase of an LOCA and produced very uniform temperature distributions. The fuel pin simulators were electrically heated (average linear power generation of 1.42 kW/m) and were slightly cooled with a very low flow (Re approx. 140) of low-pressure superheated steam. The cladding temperature increased from the initial temperature (330/sup 0/C) to the burst temperature at a rate of 3.5/sup 0/C/s. The simulators burst in a very narrow temperature range, with an average of 930/sup 0/C. Cladding burst strain ranged from 21 to 56%, with an average of 31%. Volumetric expansion over the heated length of the cladding ranged from 16 to 32%, with an average of 23%. 23 references.

  10. Fire-protection research program for the US Nuclear Regulatory Commission, 1975-1981. [PWR; BWR

    SciTech Connect

    Dube, D.A.

    1983-04-01

    Since early 1975, Sandia National Laboratories has been conducting fire-protection research for the US Nuclear Regulatory Commission. Testing has been done on grouped electrical-cable fires including electrical initiation, fire propagation, the effects of fire-retardant coatings and barriers, suppression, and characterization of the damage-ability of electrical cables. In addition, several studies of a more-generic nature such as fire detection, ventilation, and fire-hazards analysis methodologies were performed. This report condenses all of the test results, reports, papers, and research findings of the past seven years. Research conducted by contractors to Sandia National Laboratories is also summarized.

  11. Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes. [PWR; BWR

    SciTech Connect

    Ware, A.G.; Thinnes, G.L.

    1984-04-01

    EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping.

  12. Effect of bundle size on cladding deformation in LOCA simulation tests. [PWR; BWR

    SciTech Connect

    Chapman, R.H.; Crowley, J.L.; Longest, A.W.

    1982-01-01

    Two LOCA simulation tests were conducted to investigate the effects of temperature uniformity and radial restraint boundary conditions on Zircaloy cladding deformation. In one of the tests (B-5), boundary conditions typical of a large array were imposed on an inner 4 x 4 square array by two concentric rings of interacting guard fuel pin simulators. In the other test (B-3), the boundary conditions were imposed on a 4 x 4 square array by a non-interacting heated shroud. Test parameters conducive to large deformation were selected in order to favor rod-to-rod interactions. The tests showed that rod-to-rod interactions play an important role in the deformation process.

  13. Aging of electronics with application to nuclear power plant instrumentation. [PWR; BWR

    SciTech Connect

    Johnson, Jr, R T; Thome, F V; Craft, C M

    1983-01-01

    A survey to identify areas of needed research to understand aging mechanisms for electronics in nuclear power plant instrumentation has been completed. The emphasis was on electronic components such as semiconductors, capacitors, and resistors used in safety-related instrumentation in the reactor containment area. The environmental and operational stress factors which may produce degradation during long-term operation were identified. Some attention was also given to humidity effects as related to seals and encapsulants, and failures in printed circuit boards and bonds and solder joints. Results suggest that neutron as well as gamma irradiations should be considered in simulating the aging environment for electronic components. Radiation dose-rate effects in semiconductor devices and organic capacitors need to be further investigated, as well as radiation-voltage bias synergistic effects in semiconductor devices and leakage and permeation of moisture through seals in electronics packages.

  14. Experimental facility for containment sump reliability studies (Generic Task A-43). [PWR; BWR

    SciTech Connect

    Durgin, W. W.; Padmanabhan, M.; Janik, C. R.

    1980-12-01

    On July 3, 1979, Sandia National Laboratories (Sandia) contracted the Alden Research Laboratory (ARL) to conduct tests on unresolved safety issues associated with containment sump performance during the recirculation mode (Generic Task A-43). This report describes the test facility constructed and completed under Phase I, Task III of the contract. Sump performance is determined through the observation of vortex formation in the main tank and the measurement of swirl, pressure gradient, and entrained air in the suction pipes. The use of electrically operated valves and a sophisticated data acquisition system, with computer interface, allows the test flow parameters to be set and test data to be taken (with the exception of vortex observations) from a single central office.

  15. Main-coolant-pump shaft-seal guidelines. Volume 2. Operational guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria for improving main coolant pump shaft seal operational reliability. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies. Usage procedures/practices and operational environment influence on seal life and reliability from the most recent such survey are summarized. The shaft seal and its auxiliary supporting systems are discussed both from technical and operational related viewpoints.

  16. Parametric calculations of fatigue-crack growth in piping. [PWR; BWR

    SciTech Connect

    Simonen, F.A.; Goodrich, C.W.

    1983-03-01

    A major objective of this program is to provide data that can be used to formulate recommended revisions to ASME Section XI and regulatory requirements for inservice inspection of piping and pressure vessels. This study presents calculations of the growth of piping flaws produced by fatigue. Flaw growth was predicted as a function of the initial flaw size, the level and number of stress cycles, the piping material, and environmental factors.

  17. Experimental study on natural-convection boiling burnout in an annulus. [PWR; BWR

    SciTech Connect

    Mishima, K.; Ishii, M.

    1982-01-01

    An experimental study was performed on burnout heat flux at low flow rates for low-pressure steam-water upward flow in an annulus. The data indicated that a premature burnout occurred due to flow-regime transition from churn-turbulent to annular flow. It is shown that the burnout observed in the experiment is essentially a flooding-limited burnout and the burnout heat flux can be well reproduced by a nondimensional correlation derived from the previously obtained criterion for flow-regime transition. It is also shown that the conventional correlations for burnout heat flux at low mass velocities agree well with the data on circulation and entrainment-limited burnout.

  18. Full-scale turbine-missile-casing tests. Final report. [PWR; BWR

    SciTech Connect

    Yoshimura, H.R.; Schamaun, J.T.

    1983-01-01

    Results are presented of two full-scale tests simulating the impact of turbine disk fragments on simple ring and shell structures that represent the internal stator blade ring and the outer housing of an 1800-rpm steam turbine casing. The objective was to provide benchmark data on both the energy-absorbing mechanisms of the impact process and, if breakthrough occured, the exit conditions of the turbine missile. A rocket sled was used to accelerate a 1527-kg (3366-lb) segment of a turbine disk, which impacted a steel ring 12.7 cm (5 in.) thick and a steel shell 3.2 cm (1.25 in.) thick. The impact velocity of about 150 m/s (492 ft/s) gave a missile kinetic energy corresponding to the energy of a fragment from a postulated failure at the design overspeed (120% of operating speed). Depending on the orientation of the missile at impact, the steel test structure either slowed the missile to 60% of its initial translational velocity or brought it almost to rest (an energy reduction of 65 and 100%, respectively). The report includes structural and finite element analysis and data interpretation, estimates of energy during impact, missile displacement and velocity histories, and selected strain gage data.

  19. Analytical modeling of the buffeting of a rod in axial flow. [PWR; BWR

    SciTech Connect

    Lin, W.H.; Wamsganss, M.W.

    1981-12-01

    Turbulent buffeting of a circular, flexible rod in axial flows is reported. The main excitation mechanisms are turbulent wall-pressure fluctuations and the motion-dependent force field caused by the rod motion. On the assumption that the turbulent wall-pressure fluctuations are independent of rod motion, a linear forced vibration model is proposed to compute the buffeting displacement of the rod with the aid of empirical constants determined from experimental measurements of wall-pressure fluctuations. Predicted and measured values of the root-mean-square rod displacement are shown to be in reasonably good agreement.

  20. Application of RELAP5 to a pipe blowdown experiment. [PWR; BWR

    SciTech Connect

    Carlson, K.E.; Ransom, V.H.; Wagner, R.J.

    1980-01-01

    The application of the RELAP5 computer program to a pipe blowdown experiment is described in this paper. The basic hydrodynamic model, constitutive relations, and special process models included in RELAP5 are also briefly discussed. The results of this application confirm the effectiveness of using a choked flow model.

  1. Hydrogen combustion and control studies in intermediate scale. Final report. [PWR; BWR

    SciTech Connect

    Torok, R.; Siefert, K.; Wachtler, W.; Gay, R.R.; Gloski, D.M.; Wanless, J.W.

    1983-06-01

    Experiments were conducted to examine the combustion behavior of hydrogen under containment conditions which might occur in a postulated degraded-core accident. Parameters included hydrogen concentration, hydrogen and steam flow rates, water-vapor concentration, igniter location, and water-spray characteristics. Both quiescent (premixed) and dynamic (continuous injection) tests were conducted in a vessel of 213 cm (7 ft) internal diameter by 518 cm (17 ft) high, having a volume of 17,800 liters (630 ft/sup 3/). Measurements were made of temperatures, pressure, flamefront propagation, and gas-constituent concentration. The maximum overepressure increase in the 17 dynamic injection tests was 193 kPa (28 psi) for a case without steam addition to the hydrogen flow and without water sprays or fog. The presence of steam injected with hydrogen generally lowered the maximum pressure observed. Equipment response scoping tests were performed on a representative sample of safety-related Class 1E equipment and cable typically used in nuclear-reactor containments.

  2. Experimental damping data for dynamic analysis of nuclear power plant piping systems. [PWR; BWR

    SciTech Connect

    Ware, A.G.

    1983-01-01

    A summary of damping values reported in some recent piping system damping experiments and best estimate values for those systems is presented. The majority of the data is from tests conducted at the Heissdampfreaktor (HDR) in the Federal Republic of Germany. Data from the Kuosheng plant (Taiwan) and the LaSalle and Indian Point plants (US) are also included. From the data surveyed, the most significant influence on damping was the type of supports used. Other influential parameters were excitation level and response frequency. Remaining effects were minor or could not be determined from the available data. Rayleigh curve fits generally represent the data adequately and can be used in many structural codes. The USNRC Regulatory Guide 1.61 recommended damping values are shown to provide a conservative lower bound to the best estimate values reported, especially at frequencies below 20 Hz, for systems supported by seismic restraints.

  3. Overview of the use of prestressed concrete in US nuclear power plants. [PWR; BWR

    SciTech Connect

    Ashar, H.; Naus, D.J.

    1983-01-01

    In the United States it is required that the condition and functional capability of the ungrouted post-tensioning systems of prestressed-concrete nuclear-power-plant containments be periodically assessed. This is accomplished, in part, systematically through an inservice tendon inspection program which must be developed and implemented for each containment. An overview of the essential elements of the inservice inspection requirements is presented, and the effectiveness of these requirements is demonstrated through presentation of some of the potential problem areas which have been identified through the periodic assessments of the structural integrity of containments. Also, a summary of general problems which have been encountered with prestressed-concrete construction at nuclear-power-plant containments in the United States is presented: that is, dome delamination, cracking of anchorheads, settlement of bearing plates, etc. The paper will conclude with an assessment of the overall effectiveness of the prestressed-concrete containments.

  4. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    SciTech Connect

    J.M. Scaglione

    2004-08-25

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials.

  5. Physics of hydride fueled PWR

    NASA Astrophysics Data System (ADS)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  6. Bi-content Gadolinia as Burnable Absorber in PWR to Improve the Reactor Core Behaviour

    SciTech Connect

    Zheng, S.

    2007-07-01

    The gadolinia product is one of the standard burnable absorbers used in the PWR long and low leakage fuel cycle in order to control the radial power distribution and to hold down the initial core reactivity. This product presents a large number of advantages such as the high efficiency with only a small number of gadolinia-bearing rods, the easy adjustment between the number and the content of the gadolinia-bearing rods according to the cycle length need and the initial reactivity hold-down, no increasing of boron concentration versus cycle depletion, no additional increasing of internal pressure in poisoned rods, very low additional manufacture cost. On the other hand, some unfavourable phenomena are also observed during the utilization of the gadolinia: amplification of the asymmetrical power distribution and more negative axial offset. Based on the correlation between the gadolinia burnout and its content, the use of gadolinia bi-content will improve the parameters indicated here above. The gadolinia bi-content have been used in BWR for more than 20 years. In this paper, the comparison of the main reactor core physical parameters in PWR, calculated with the AREVA NP standard neutronic code package SCIENCE, is made by using the mono- and bi-content of the gadolinia products in the same fuel assembly. The results show that the asymmetrical axial and azimuthal power distribution can be improved in the case of the bi-content gadolinia product. (authors)

  7. Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2013-07-01

    The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

  8. RIA Limits Based On Commercial PWR Core Response To RIA

    SciTech Connect

    Beard, Charles L.; Mitchell, David B.; Slagle, William H.

    2006-07-01

    Reactivity insertion accident (RIA) limits have been under intense review by regulators since 1993 with respect to what should be the proper limit as a function of burnup. Some national regulators have imposed new lower limits while in the United States the limits are still under review. The data being evaluated with respect to RIA limits come from specialized test reactors. However, the use of test reactor data needs to be balanced against the response of a commercial PWR core in setting reasonable limits to insure the health and safety of the public without unnecessary restrictions on core design and operation. The energy deposition limits for a RIA were set in the 1970's based on testing in CDC (SPERT), TREAT, PBF and NSRR test reactors. The US limits given in radially averaged enthalpy are 170 cal/gm for fuel cladding failure and 280 cal/gm for coolability. Testing conducted in the 1990's in the CABRI, NSRR and IGR test reactors have demonstrated that the cladding failure threshold is reduced with burnup, with the primary impact due to hydrogen pickup for in-reactor corrosion. Based on a review of this data very low enthalpy limits have been proposed. In reviewing proposed limits from RIL-0401(1) it was observed that much of the data used to anchor the low allowable energy deposition levels was from recent NSRR tests which do not represent commercial PWR reactor conditions. The particular characteristics of the NSRR test compared to commercial PWR reactor characteristics are: - Short pulse width: 4.5 ms vs > 8 ms; - Low temperature conditions: < 100 deg. F vs 532 deg. F. - Low pressure environment: atmospheric vs {approx} 2200 psi. A review of the historical RIA database indicates that some of the key NSRR data used to support the RIL was atypical compared to the overall RIA database. Based on this detailed review of the RIA database and the response of commercial PWR core, the following view points are proposed. - The Failure limit should reflect local fuel

  9. A simplified spatial model for BWR stability

    SciTech Connect

    Berman, Y.; Lederer, Y.; Meron, E.

    2012-07-01

    A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

  10. Assessment of severe accident prevention and mitigation features: BWR (boiling water reactor), Mark I containment design

    SciTech Connect

    Pratt, W.T.; Eltawila, F.; Perkins, K.R.; Fitzpatrick, R.G.; Luckas, W.J.; Lehner, J.R.; Davis, P.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark I containments (BWR Mark I's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Peach Bottom plant and from assessment of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark I to severe accident containment loads were also identified. In addition, those features of a BWR Mark I, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Peach Bottom and other Mark I plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance.

  11. LBB application in Swedish BWR design

    SciTech Connect

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  12. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    SciTech Connect

    Putney, J.M.; Preece, R.J.

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  13. Evaluation of BWR emergency procedure guidelines for BWR ATWS using RAMONA-3B code

    SciTech Connect

    Neymotin, L.; Slovik, G.; Cazzoli, E.; Saha, P.

    1985-01-01

    An MSIV Closure ATWS calculation for a typical BWR/4 (Browns Ferry, Unit 1) was performed using the RAMONA-3B code which is a BWR systems transient code combining three-dimensional neutronic core representation with multi-channel one-dimensional thermal hydraulics. The main objective of the study was to perform a best-estimate evaluation of the recently proposed Emergency Procedure Guidelines for Anticipated Transients Without Scram (ATWS). Emphasis was placed on evaluating the effects of lowering the downcomer water level to the Top of Active Fuel (TAF) and vessel depressurization. The calculation was run up to approximately 1200 seconds. Both actions, namely, lowering the water level and vessel depressurization, lowered the reactor power to some extent. However, the pressure suppression pool water temperature still reached approximately 90/sup 0/C (potential High Pressure Coolant Injection (HPCI) pump seal failure temperature) in twenty minutes. Thus, other actions such as boron injection and/or manual control rod insertion are necessary to mitigate a BWR/4 Main Steam Isolation Valve (MSIV) closure ATWS. 19 refs., 14 figs., 3 tabs.

  14. BWR control-rod cobalt-alloy replacement. Final report

    SciTech Connect

    Aldred, P.

    1982-03-01

    Cobalt base pin and roller alloys in BWR Control Rods are a source for the Co-60 isotope which contributes to radiation buildup in the BWR core, the recirculation piping system and the spent fuel pool. It thereby influences personnel radiation exposure during BWR plant maintenance. The program objectives were (a) to identify non-cobalt alloys which could potentially replace the cobalt alloys, (b) evaluate the alloys by testing to qualify them for in-reactor surveillance testing, and (c) to initiate reactor tests at 2 BWRs. Wear resistance, an important requirement for pins and rollers, was measured in a simulated BWR environment (excluding irradiation). Prototypic wear tests were emphasized and a prototype control rod drive test facility was used to evaluate several pin and roller alloy combinations during prototype control rod operations.

  15. Development of long operating cycle simplified BWR

    SciTech Connect

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-07-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  16. BWR Core Heat Transfer Code System.

    1999-04-27

    Version 00 MOXY is used for the thermal analysis of a planar section of a boiling water reactor (BWR) fuel element during a loss-of-coolant accident (LOCA). The code emplyoys models that describe heat transfer by conduction, convection, and thermal radiation, and heat generation by metal-water reaction and fission product decay. Models are included for considering fuel-rod swelling and rupture, energy transport across the fuel-to-cladding gap, and the thermal response of the canister. MOXY requires thatmore » time-dependent data during the blowdown process for the power normalized to the steady-state power, for the heat-transfer coefficient, and for the fluid temperature be provided as input. Internal models provide these parameters during the heatup and emergency cooling phases.« less

  17. Advanced Construction of Compact Containment BWR

    SciTech Connect

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-07-01

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  18. The Effect of Improved Water Chemistry on Corrosion Cracking of BWR Piping: Workshop Proceedings

    SciTech Connect

    1989-12-01

    Implementation of the EPRI BWR water chemistry guidelines by utilities has significantly improved the chemistry of BWRs. Water chemistry improvements extend the service life of BWR piping and provide a technical justification for increased intervals of in-service inspections of BWR piping.

  19. BWR Source Term Generation and Evaluation

    SciTech Connect

    J.C. Ryman

    2003-07-31

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the

  20. Design study status of compact containment BWR

    SciTech Connect

    Heki, H.; Nakamaru, M.; Kuroki, M.; Kojima, Y.; Arai, K.; Tahara, M.; Hoshi, T.

    2006-07-01

    The reactor concept considered in this paper has a relatively mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Japan Atomic Power Company has been taking initiative in developing the concept of the Compact Containment Boiling Water Reactor (CCR). The CCR., which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's relatively mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, upper entry control rod drives (CRDs) and simplified safety system with high pressure resistible containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The upper entry CRDs enable a simplified safety system followed by in-vessel retention (IVR) capability with the compact primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration of RPV below the top of the core height, passive cooling system by isolation condenser (IC). The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. Further core design study has been carried out taking into account compact reactor size and reduction of fuel

  1. Status update of the BWR cask simulator

    SciTech Connect

    Lindgren, Eric R.; Durbin, Samuel G.

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of

  2. An overview of the BWR ECCS strainer blockage issues

    SciTech Connect

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

    1996-03-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, {open_quotes}Containment Emergency Sump Performance,{close_quotes} and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts.

  3. BWR full integral simulation test (FIST) program. TRAC-BWR model development. Volume 3. Developmental assessment for plant application

    SciTech Connect

    Cheung, Y.K.; Andersen, J.G.M.; Chu, K.H.; Shaug, J.C.

    1985-11-01

    The TRACB04 computer code has been developed under the model development tasks in the FIST Program. This report describes two developmental assessment calculations performed on BWR plants with TRACB04. A BWR/2 Design Basis Accident (DBA) including the containment response and a BWR/4 DBA with Low Pressure Coolant Injection (LPCI) water injected into the lower plenum were calculated and results of these cases were documented. These cases serve to test some of the new features of the TRACB04 (air field, containment model, ''water packing'' fixes and faster numerics in the three dimensional vessel component) and to demonstrate that the code has been assembled properly. They also provide best estimate LOCA results for the two plant types.

  4. Horizontal Drop of 21- PWR Waste Package

    SciTech Connect

    A.K. Scheider

    2007-01-31

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design.

  5. Assessment of void swelling in austenitic stainless steel PWR core internals.

    SciTech Connect

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and

  6. PWR secondary water chemistry guidelines: Revision 3. Final report

    SciTech Connect

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239).

  7. A review of plant decontamination methods: 1988 Update: Final report

    SciTech Connect

    Remark, J.F.

    1989-01-01

    This document updates the state-of-the-art in decontamination technology since the publication of the previous review (EPRI NP- 1128) in May 1981. A brief description of the corrosion-film characteristics is presented as well as corrosion film differences between a BWR and PWR. The generation transportation, activation, and deposition of the radioisotopes found throughout the reactor coolant system is also discussed. Successful, well executed, decontamination campaigns are always preceded by meticulous planning and careful procedure preparation which include contingency operations. The Decontamination Planning and Preparation Section describes the technical planning steps as well as the methodology that should be followed in order to select the optimum decontamination technique for a specific application. A review of a number of the decontamination methods commercialized since 1980 is presented. The basic mechanism for each process is described as well as specific applications of the technology in the fields. Where possible, results obtained in the field are presented. The information was obtained from industry vendors as well as personnel at the plant locations that have utilized the technology. 72 refs., 5 tabs.

  8. BWR Full Integral Simulation Test (FIST) Program. TRAC-BWR model development. Volume 2. Models

    SciTech Connect

    Chu, K.H.; Andersen, J.G.M.; Cheung, Y.K.; Shaug, J.C.

    1985-11-01

    TRAC-BWR (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a Boiling Water Reactor system. In this report, the development of new models and the implementation of the balance of plant models leading to the creation of the TRACB04 version of the code, is described. The new models include an improved model for boron transport which accounts for non-uniform mixing and stratification, and a model for the interfacial heat transfer at two-phase levels. The balance of plant models (turbine, containment and heat exchanger) developed at INEL were evaluated, adapted, and implemented into TRACB04 to provide complete transient analysis capability. In addition, a model for air or a noncondensible gas as an additional field in the system of equations was adapted to the two step numerical method and incorporated into TRACB04.

  9. Technical considerations related to interim source-term assumptions for emergency planning and equipment qualification. [PWR; BWR

    SciTech Connect

    Niemczyk, S.J.; McDowell-Boyer, L.M.

    1982-09-01

    The source terms recommended in the current regulatory guidance for many considerations of light water reactor (LWR) accidents were developed a number of years ago when understandings of many of the phenomena pertinent to source term estimation were relatively primitive. The purpose of the work presented here was to develop more realistic source term assumptions which could be used for interim regulatory purposes for two specific considerations, namely, equipment qualification and emergency planning. The overall approach taken was to adopt assumptions and models previously proposed for various aspects of source term estimation and to modify those assumptions and models to reflect recently gained insights into, and data describing, the release and transport of radionuclides during and after LWR accidents. To obtain illustrative estimates of the magnitudes of the source terms, the results of previous calculations employing the adopted assumptions and models were utilized and were modified to account for the effects of the recent insights and data.

  10. Preliminary results of thermal igniter experiments in H/sub 2/-air-steam environments. [PWR; BWR

    SciTech Connect

    Lowry, W.

    1981-01-01

    Thermal igniters (glow plugs), proposed by the Tennessee Valley Authority for intentional ignition of hydrogen in nuclear reactor containment, have been tested for functionability in mixtures of air, hydrogen, and steam. Test environments included 6% to 16% hydrogen concentrations in air, and 8%, 10%, and 12% hydrogen in mixtures with 30% and 40% steam fractions. All were conducted in a 10.6 ft/sup 3/ insulated pressure vessel. For all of these tests the glow plug successfully initiated combustion. Dry air/hydrogen tests exhibited a distinct tendency for complete combustion at hydrogen concentrations between 8% and 9%. Steam suppressed both peak pressures and completeness of combustion. No combustion could be initiated at or above a 50% steam fraction. Circulation of the mixture with a fan increased the completeness of combustion. The glow plug showed no evidence of performance degradation throughout the program.

  11. Critical heat-flux experiments under low-flow conditions in a vertical annulus. [PWR; BWR; LMFBR

    SciTech Connect

    Mishima, K.; Ishii, M.

    1982-03-01

    An experimental study was performed on critical heat flux (CHF) at low flow conditions for low pressure steam-water upward flow in an annulus. The test section was transparent, therefore, visual observations of dryout as well as various instrumentations were made. The data indicated that a premature CHF occurred due to flow regime transition from churn-turbulent to annular flow. It is shown that the critical heat flux observed in the experiment is essentially similar to a flooding-limited burnout and the critical heat flux can be well reproduced by a nondimensional correlation derived from the previously obtained criterion for flow regime transition. The observed CHF values are much smaller than the standard high quality CHF criteria at low flow, corresponding to the annular flow film dryout. This result is very significant, because the coolability of a heater surface at low flow rates can be drastically reduced by the occurrence of this mode of CHF.

  12. Comparison of methods for uncertainty analysis of nuclear-power-plant safety-system fault-tree models. [PWR; BWR

    SciTech Connect

    Martz, H F; Beckman, R J; Campbell, K; Whiteman, D E; Booker, J M

    1983-04-01

    A comparative evaluation is made of several methods for propagating uncertainties in actual coupled nuclear power plant safety system faults tree models. The methods considered are Monte Carlo simulation, the method of moments, a discrete distribution method, and a bootstrap method. The Monte Carlo method is found to be superior. The sensitivity of the system unavailability distribution to the choice of basic event unavailability distribution is also investigated. The system distribution is also investigated. The system distribution is especially sensitive to the choice of symmetric versus asymmetric basic event distributions. A quick-and dirty method for estimating percentiles of the system unavailability distribution is developed. The method identifies the appropriate basic event distribution percentiles that should be used in evaluating the Boolean system equivalent expression for a given fault tree model to arrive directly at the 5th, 10th, 50th, 90th, and 95th percentiles of the system unavailability distribution.

  13. Experimental stress analysis and fatigue tests of five 24-in. NPS ANSI Standard B16. 9 tees. [PWR; BWR

    SciTech Connect

    Moore, S.E.; Hayes, J.K.; Weed, R.A.

    1985-03-01

    Experimental stress analyses and low-cycle fatigue tests of five 24-in. nominal pipe size American National Standards Institute (ANSI) Standard B16.9 forged tees are documented in this report. The tees, designated as Oak Ridge National Laboratory tees T10, T11, T12, T13, and T16, were tested under subcontract at Combustion Engineering, Inc. in Chattanooga, Tennessee. Experimental stress analyses were conducted for 12 individual loadings on each tee. Each test model was instrumented with approx. 225, 1/8-in. three-gage, 45/sup 0/ strain rosettes on the inside and outside surfaces; and 6 linear variable differential transformers mounted on special nonflexible holding frames for measuring deflections and rotations of the pipe extensions. Following completion of the strain-gate tests, each tee was fatigue tested to failure with either a fully reversed displacement controlled in-plane bending moment on the branch or a cyclic internal pressure that ranged from a value slightly above zero to about 90% of the nominal yield pressure of the pipe extensions.

  14. Procedures for using expert judgment to estimate human-error probabilities in nuclear power plant operations. [PWR; BWR

    SciTech Connect

    Seaver, D.A.; Stillwell, W.G.

    1983-03-01

    This report describes and evaluates several procedures for using expert judgment to estimate human-error probabilities (HEPs) in nuclear power plant operations. These HEPs are currently needed for several purposes, particularly for probabilistic risk assessments. Data do not exist for estimating these HEPs, so expert judgment can provide these estimates in a timely manner. Five judgmental procedures are described here: paired comparisons, ranking and rating, direct numerical estimation, indirect numerical estimation and multiattribute utility measurement. These procedures are evaluated in terms of several criteria: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Situational constraints such as the number of experts available, the number of HEPs to be estimated, the time available, the location of the experts, and the resources available are discussed in regard to their implications for selecting a procedure for use.

  15. One- and two-dimensional STEALTH simulations of fuel-pin transient response. Final report. [BWR; PWR

    SciTech Connect

    Wahi, K.K.

    1980-08-01

    This report presents an assessment of the adaptability of EPRI's one- and two-dimensional STEALTH computer codes to perform transient fuel rod analysis. The ability of the STEALTH code to simulate transient mechanical or thermomechanical loss-of-coolant accident is described. Analytic models of one- and two-dimensional formulations and features included in the two-dimensional simulation are discussed.

  16. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    SciTech Connect

    Langenbuch, S.; Velkov, K.; Lizorkin, M.

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  17. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    SciTech Connect

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degree}C and whether the cladding of the stored spent fuel ever exceeds 350{degree}C. Limiting the borehole to temperatures of 97{degree}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degree}C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degree}C for the full 1000-yr analysis period.

  18. Main-coolant-pump shaft-seal guidelines. Volume 1. Maintenance-manual guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and a listing of information and data which should be included in maintenance manuals and procedures for Main Coolant Pump Shaft Seals. The noted guidelines and data listing are developed from EPRI sponsored nuclear plant seal operating experience studies. The maintenance oriented results of the most recent such study is summarized. The shaft seal and its auxiliary supporting systems are discussed from both technical and maintenance related viewpoints.

  19. Heavy-section steel technology program. Volume 1. Quarterly progress report, January-March 1983. [PWR; BWR

    SciTech Connect

    Pugh, C.E.

    1983-09-01

    A thermal-strain modification was made to the deformation-plasticity model in the ADINA-ORVIRT fracture-mechanics analysis system in order to be more applicable to combined pressure and thermal loadings. Subcontractors continued studies on crack arrest, cleavage fracture transition, and environmentally assisted crack growth. Charpy testing of state-of-the-art weld specimens in the Fourth HSST Irradiation Series was performed on unirradiated specimens and on a few irradiated specimens for scoping purposes. Finite-flaw capabilities were incorporated into the OCA-II computer code, and parametric studies were carried out to compare fracture predictions with two-dimensional and specific finite flaws. Preparations continued for thermal-shock experiment TSE-7 to be conducted in May. Preparations for the first pressurized-thermal-shock experiment continued.

  20. Assessment of Biasi and Columbia University CHF correlations with GE 3x3 rod bundle experiment. [PWR; BWR

    SciTech Connect

    Chen, B.C.J.; Chien, T.H.; Sha, W.T.; Kim, J.H.

    1984-01-01

    The critical heat flux (CHF), at which a sudden degradation of heat transfer occurs without corresponding decrease in heat generation, is one of the limiting parameters for safe operation of nuclear reactors. Reactor operation beyond the CHF causes a rapid rise in fuel cladding temperature and thus should be avoided to maintain the fuel element integrity. Reactor power limits are therefore set so that a prescribed safety margin below the CHF is maintained. Two CHF correlations are evaluated for reactor core thermal hydraulic analysis: the Biasi correlation and the Columbia University correlation. The BODYFIT-2PE computer code is used for this assessment. The CHF predicted by the BODYFIT-2PE using the two correlations is compared with GE 3x3 rod bundle CHF experiment.

  1. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail.

  2. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual. [PWR; BWR

    SciTech Connect

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code.

  3. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces.

  4. Identification of seismically risk-sensitive systems and components in nuclear power plants: feasibility study. [PWR; BWR

    SciTech Connect

    Azarm, M; Boccio, J; Farahzad, P

    1983-06-01

    An approach for the identification of risk-sensitive components in a nuclear power plant during and after a seismic event is described. Application of the methodology to two hypothetical power plants - a Boiling Water Reactor and a Pressurized Water Reactor - are presented and the results are given in tabular and graphical form. Conclusions drawn and lessons learned through the course of this study, based on the relative importance of various accident scenarios and sensitivity analyses, are discussed. In addition, the areas that may need further investigation are identified.

  5. Plant heat cycles, vessel internal arrangement, and auxiliary systems. Volume five

    SciTech Connect

    Not Available

    1986-01-01

    This volume covers nuclear power plant heat cycles (type of nuclear power cycles, power cycle refinements, BWR/PWR power cycle, BWR/PWR reactor coolant system), reactor vessel internal arrangement (reactor vessel features, BWR/PWR reactor vessel and internals, BWR/PWR reactor core), reactor auxiliary systems (purpose of reactor auxiliary systems, PWR and BWR reactor auxiliary systems, PWR and BWR control rod drive mechanisms).

  6. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    SciTech Connect

    Redding, J.R.

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  7. Generic aging management programs for license renewal of BWR reactor coolant systems components.

    SciTech Connect

    Shah, V.N.; Liu, Y.Y.

    2002-02-15

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  8. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    SciTech Connect

    Shah, V.N.; Liu, Y.Y.

    2002-07-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  9. The Development of the Evolutionary BWR

    SciTech Connect

    Murase, A.; Nakamaru, M.; Kuroki, M.; Kojima, Y.; Yokoyama, S.

    2006-07-01

    Considering the delay of the fast breeding reactor (FBR) development, it is expected that the light water reactor will still play the main role of the electric power generation in the 2030's. Accordingly, Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to improve countermeasure against severe accident (SA). At the same time, we have made the simplification of the overall plant systems in order to improve economy. The simplification of the AB1600 is based on the proven technologies. To retain the safety performance superior or equivalent to the current ABWR and to strengthen the countermeasure against SA, the AB1600 has introduced the passive systems such as the passive containment cooling system (PCCS), the gravity driven core cooling system (GDCS) and the isolation condenser (IC). While we retain the safety performance superior or equivalent to the current ABWR, we have made the simplification of the safety systems. We could eliminate the high pressure core flooder system (HPCF) and the reactor core isolation system (RCIC) by extending the height of reactor pressure vessel (RPV) two meters. To achieve simplification of reactor systems, we have reduced the number of fuel bundles and the number of control rods by adopting large bundle that has a bundle pitch 1.2 times wider than that of the current ABWR. In the 1600 MWe class, the number of fuel bundles could be reduced to 600 from 872 of the current ABWR, and the number of control rods could be reduced to 137 from 205 of the current ABWR. Because the reactor internal pump (RIP) of the current ABWR has sufficient performance capacity and the improvement of fuel characteristics from the current fuel enables the operation at lower core flow, the number of RIPs could be decreased from ten to eight. Furthermore, we have reduced the number of divisions of emergency core cooling system (ECCS

  10. Some Aspects of Cost/ Benefit Analysis for In-Service Inspection of PWR Steam Generators

    SciTech Connect

    Zima, G. E.; Lyon, G. H.; Doctor, P. G.; Hoenes, G. R.; Petty, S. E.; Weakley, S. A.

    1981-05-01

    This report discusses a number of aspects of cost/benefit (C/B) analysis for in-service inspection (lSI} of pressurized water reactor (PWR) steam generators (SGs) and identifies several problem areas that must be addressed prior to a full C/B analysis capability. Following a brief review of the impact of SG problems on the productivity of PWR units and of the scope and variability of SG problems among U.S. PWRs, various occupational implications of SG lSI are considered, namely manpower, time, and rad exposure. The opportunities provided by refueling outages in respect to lSI frequency and work time windows are reviewed. Indices for characterizing the nondestructive testing {NDT) information, rad exposure, $ impact, and manpower and time attributes of single ISIs and a series of ISIs over an arbitrary evaluation period are presented and calculated for a number of lSI cases using SG parameters for three typical PWR units. A comparison of the $ impact of unscheduled outages attributable to SG problems with the $ cost of ambitious lSI strategies indicates that the $ cost is virtually negligible for well-planned ISis. Considering the ALARA constraint on occupational rad exposure, the skilled manpower pool for NDT work appears to be the principal factor limiting lSI scope and frequency. Analysis of the manpower and time requirements for inspection of a 40-unit PWR population indicates, however, that an lSI strategy embodying two campaigns per year and a total population inspection within a 2-year interval is not far beyond current capabilities.

  11. BWR refill-reflood program: core spray distribution experimental task plan

    SciTech Connect

    Eckert, T.

    1981-02-01

    An experimental task plan for the BWR/4 core spray task of the Refill-Reflood Test Program is presented. The test program will provide core spray distribution data for a 30 degree sector of the BWR/4 and 5-218 design. This design uses different nozzle types and different sparger elevations than the BWR/6-218 design which was tested previously. Test parameter ranges are specified; individual tests are defined; and measurement and data utilization plans are defined.

  12. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    SciTech Connect

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y.; Makigami, T.

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  13. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    SciTech Connect

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  14. Application of the gamma thermometer to BWR core monitoring

    SciTech Connect

    Martin, C.L.

    1996-12-31

    Boiling water reactor (BWR) core monitoring systems rely on in-core instrumentation to help determine the precise axial and radial power distribution of the core. Recently, it has been proposed to replace the existing traversing in-core probe (TIP) system with a system based on fixed in-core gamma thermometers. In this paper, the author describes the type of gamma thermometer (GT) that could be used in the proposed system and provides results from an ongoing implant test program.

  15. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    SciTech Connect

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  16. Operating the plant, quality assurance, and the job of the operating staff, Volume Twelve

    SciTech Connect

    Not Available

    1986-01-01

    Subject matter includes operating the plant (the role of the operator, the control room, plant technical specifications, plant operating procedures, initial startup program, BWR/PWR plant startup, BWR/PWR steady state power operation, BWR/PWR transient operation, emergency operation), quality assurance (what is quality, what is quality control, quality assurance includes quality control, government regulation and quality assurance, administrative controls for nuclear power plants, the necessity of reviews and audits, practical quality assurance), and the job of the operating staff (the plant operating staff, plant safety, first aid and resuscitation, general plant hazards, personnel protective equipment, handling chemicals, handling compressed gas, equipment repair and maintenance, communicating with others.

  17. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  18. The BWR advanced fuel design experience using Studsvik CMS

    SciTech Connect

    DiGiovine, A.S.; Gibbon, S.H.; Wiksell, G.

    1996-12-31

    The current trend within the nuclear industry is to maximize generation by extending cycle lengths and taking outages as infrequently as possible. As a result, many utilities have begun to use fuel designed to meet these more demanding requirements. These fuel designs are significantly more heterogeneous in mechanical and neutronic detail than prior designs. The question arises as to how existing in-core fuel management codes, such as Studsvik CMS perform in modeling cores containing these designs. While this issue pertains to both pressurized water reactors (PWRs) and boiling water reactors (BWRs), this summary focuses on BWR applications.

  19. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    SciTech Connect

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O/sub 2/ fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO/sub 2/ fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O/sub 2/-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O/sub 2/-fueled BWR should perform similar to a UO/sub 2/-fueled BWR under all operating conditions. A (Pu/Th)O/sub 2/-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO/sub 2/-fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths.

  20. Physics review on inherently safe features of ESBWR

    SciTech Connect

    Chiang, R. T.; Fawcett, R. M.; Cheung, Y. K.; Chung, A. K.

    2006-07-01

    The scope of this physics review includes: 1) the major differences among ESBWR, ABWR and conventional BWR cores, 2) the reason why ESBWR operation is inherently safe based on stability analysis, 3) an innovative wide-blades control rod conceptual core design to reduce cost by reducing number of control rod drives by near 50% for a natural circulation BWR, and 4) an innovative top-entry control rod conceptual core design to take advantage of additional space in the chimney area in order to reduce the plant size and cost for a natural circulation BWR. (authors)

  1. A PWR Thorium Pin Cell Burnup Benchmark

    SciTech Connect

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  2. High Cycle Thermal Fatigue in French PWR

    SciTech Connect

    Blondet, Eric; Faidy, Claude

    2002-07-01

    Different fatigue-related incidents which occurred in the world on the auxiliary lines of the reactor coolant system (SIS, RHR, CVC) have led EDF to search solutions in order to avoid or to limit consequences of thermodynamic phenomenal (Farley-Tihange, free convection loop and stratification, independent thermal cycling). Studies are performed on mock-up and compared with instrumentation on nuclear power stations. At the present time, studies allow EDF to carry out pipe modifications and to prepare specifications and recommendations for next generation of nuclear power plants. In 1998, a new phenomenal appeared on RHR system in Civaux. A crack was discovered in an area where hot and cold fluids (temperature difference of 140 deg. C) were mixed. Metallurgic studies concluded that this crack was caused by high cycle thermal fatigue. Since 1998, EDF is making an inventory of all mixing areas in French PWR on basis of criteria. For all identified areas, a method was developed to improve the first classifying and to keep back only potential damage pipes. Presently, studies are performing on the charging line nozzle connected to the reactor pressure vessel. In order to evaluate the load history, a mock-up has been developed and mechanical calculations are realised on this nozzle. The paper will make an overview of EDF conclusions on these different points: - dead legs and vortex in a no flow connected line; - stratification; - mixing tees with high {delta}T. (authors)

  3. BWR containments license renewal industry report; revision 1. Final report

    SciTech Connect

    Smith, S.; Gregor, F.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR`s). License renewal applicants may choose to reference these IR`s in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers.

  4. Chemical decontamination of BWR fuel and core materials

    SciTech Connect

    Beauregard, R.J. )

    1989-09-01

    A previous EPRI project decontaminated two discharged BWR fuel assemblies using the AP-LOMI and AP-CAN-DECON processes at Commonwealth Edison's Quad-Cities Nuclear Power Site. The two decontaminated assemblies and a third control assembly were shipped to the B W Hot Cell Facility in Lynchburg, Virginia. The three assemblies were partially disassembled in the hot cells and several rods extracted for nondestructive oxide measurement and visual examination. Various components were removed from the two decontaminated fuel assemblies for destructive examination to search for possible deleterious effects of chemical cleaning. The AP-LOMI process removed essentially all of the crud which normally covers a BWR bundle and channel. The AP-CAN-DECON process removed most of the crud, but left a thin layer on the rods and components in the central region of the bundle between the top and bottom spacer grids. Neither decontamination process appeared to damage the Zircaloy-2 fuel and water rods, or the Zircaloy-4 channels and spacers. An adherent zirconium oxide layer still covered all of the Zircaloy surfaces which were examined. The increase in hydrogen content of the channels and fuel rods was low. The AP-LOMI process did not appear to damage the Inconel X-750 fuel rod expansion springs, spacer lantern springs or channel finger spring. A thin, adherent oxide layer was found on all components.

  5. Proceedings: PWR Primary Startup/Shutdown Chemistry Workshop

    SciTech Connect

    2000-08-01

    This workshop summary outlines the proceedings of the EPRI-sponsored PWR Primary Startup/Shutdown Workshop held in San Antonio, Texas on April 25-27, 2000 to support the next revision of current EPRI PWR Chemistry Guidelines. Information was exchanged to assess the effectiveness of the guidelines. The workshop also helped identify issues needing further study before the next revision. Approximately 50 utility and industry representatives attended the workshop with utility personnel chairing four sessions. The workshop provided an opportunity for utility representatives to express an opinion as to the effectiveness of the existing PWR Primary Water Chemistry Guidelines: Volume 2, Revision 4. Potential improvements and additions to the Guidelines are outlined in this report.

  6. Timing analysis of PWR fuel pin failures

    SciTech Connect

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. ); Straka, M. )

    1992-09-01

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively.

  7. PWR full-reactor coolant system decontamination

    SciTech Connect

    Aspden, R.G.; Pessall, N.; Grand, T.F. )

    1992-01-01

    The overall objective of the current program is to identify and address all aspects of full system decontamination with the purpose of qualifying at least one process for PWR use. The objective of the current study is to provide baseline data on the performance of materials on the primary side after exposure to one cycle of the LOMI fault testing. This data supplements prior information obtained after exposure to three cycles of LOMI testing. The technical significance of this excursion will be determined in a subsequent task. The general corrosion characteristics of over 39 materials were evaluated for some combinations of material, type of specimen (coupon and creviced coupons), and loop velocity (0, 5, 20 and 150 ft/sec). At velocities of less than or equal to 20 ft/sec, sixteen types of specimens were employed to evaluate localized corrosion and stress corrosion cracking. Specimens were examined after one cycle. Also included in this exposure were specimens added to provide more information on the effect of LOMI fault exposure one: (1) surface roughening of Stellite 156; (2) crevice corrosion of chromium plated 304 stainless steel with the open end gap increased from 3 to {approximately} 9 mils; (3) susceptibility of Inconel X-750 (HTH) to subsequent stress corrosion cracking, (4) loss of chromium plate from threads of 304 stainless steel bolts torqued into stainless steel collars; (5) crack initiation in an Alloy 600 tube known to be susceptible to primary water stress corrosion cracking; and (6) surface alternation of stressed Inconel X-750 springs with the spring temper.

  8. Leak before break application in French PWR plants under operation

    SciTech Connect

    Faidy, C.

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  9. Effects of some little noticed water impurities on stress corrosion cracking of BWR construction materials

    SciTech Connect

    Ljungberg, L.G.; Cubicciotti, D.; Trolle, M.

    1988-01-01

    The effects of some little noticed dissolved impurities in simulated BWR water on environmental cracking for some BWR pressure bearing construction materials were studied by constant elongation rate tensile (CERT) tests. Fluoride, silica and thiosulfate were found to be harmful. Phosphate and perchlorate in concentration up to 1 ppm had no effect in simulated hydrogen water chemistry. Organic acids and zinc were found to be generally beneficial, except when they occurred in combination.

  10. Enriched boric acid for PWR application: Cost evaluation study for a twin-unit PWR

    SciTech Connect

    Battaglia, J.A.; Waters, R.M.; von Hollen, J.M.; Lamatia, L.A.; Bergmann, C.A.; Ditommaso, S.M. . Nuclear and Advanced Technology Div.)

    1989-09-01

    In the nuclear industry boric acid dissolved in the reactor coolant is used as a soluble reactivity control agent. Reactivity control in nuclear plants is also provided by neutron absorbing control rods. This neutron absorbing duty is distributed between the control rods and soluble boric acid in such a way as to provide the most economical split. Typically, the control rods take care of rapid reactivity changes and the boric acid handles the slower long term control of reactivity by varying the boric acid concentrations within the reactor coolant. In PWR reactor plants the dissolved boric acid is referred to as a soluble poison or chemical shim due to the high capacity for thermal neutron capture exhibited by the boron-10 isotope contained in the boric acid molecule. This slow reactivity change or chemical shim control would otherwise have to be performed using control rods, a much more expensive proposition. Reactivity changes are controlled by the B-10 isotope by virtue of its very high cross section (3837 barns) for thermal neutron absorption. However, natural boron contains only 20 atom percent of the B-10 isotope and essentially all the remaining 80 percent as the B-11 isotope. The B-11 isotope of cross section .005 barns is essentially of no use as a neutron absorber. Since B-11 makes up the bulk of the total boron present and contributes little to the nuclear operation it would seem logical to eliminate this isotope of boron from the boric acid molecule. In so doing boric acid concentration in operating PWR plants need only be a fraction of that existing to accomplish identical nuclear operations. However, to achieve the elimination of B-11 from NBA (Natural Boric Acid) an isotope separation must be performed. 4 refs., 25 figs., 17 tabs.

  11. BWR Full Integral Simulation Test (FIST). Phase I test results

    SciTech Connect

    Hwang, W S; Alamgir, M; Sutherland, W A

    1984-09-01

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report.

  12. BWR plant analyzer development at BNL (Brookhaven National Laboratory)

    SciTech Connect

    Wulff, W.; Cheng, H.S.; Mallen, A.N.

    1986-01-01

    An engineering plant analyzer has been developed at BNL for realistically and accurately simulating transients and severe abnormal events in BWR power plants. Simulations are being carried out routinely with high fidelity, high simulation speed, at low cost and with unsurpassed user convenience. The BNL Plant Analyzer is the only operating facility which (a) simulates more than two orders-of-magnitude faster than the CDC-7600 mainframe computer, (b) is accessible and fully operational in on-line interactive mode, remotely from anywhere in the US, from Europe or the Far East (Korea), via widely available IBM-PC compatible personal computers, standard modems and telephone lines, (c) simulates both slow and rapid transients seven times faster than real-time in direct access, and four times faster in remote access modes, (d) achieves high simulation speed without compromising fidelity, and (e) is available to remote access users at the low cost of $160 per hour.

  13. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect

    P.M. O'Leary; Dr. M.L. Pitts

    2000-08-21

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers.

  14. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  15. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    SciTech Connect

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  16. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

    SciTech Connect

    Ward, Andrew; Downar, Thomas J.; Xu, Y.; March-Leuba, Jose A; Thurston, Carl; Hudson, Nathanael H.; Ireland, A.; Wysocki, A.

    2015-04-22

    The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, the capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.

  17. Advanced BWR core component designs and the implications for SFD analysis

    SciTech Connect

    Ott, L.J.

    1997-02-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B{sub 4}C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities.

  18. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

    DOE PAGESBeta

    Ward, Andrew; Downar, Thomas J.; Xu, Y.; March-Leuba, Jose A; Thurston, Carl; Hudson, Nathanael H.; Ireland, A.; Wysocki, A.

    2015-04-22

    The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

  19. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

  20. The nature and behavior of particulates in PWR primary coolant

    SciTech Connect

    Bridle, D.A.; Butter, K.R.; Cake, P.; Comley, G.C.W.; Mitchell, C.R. )

    1989-12-01

    A study of particle size distributions, nature and behavior of insoluble species carried by PWR coolants has been carried out over a four year period in Belgian reactors. Comparative data was obtained by the use of improved sampling systems designed to overcome the inadequacies of standard facilities. Coolant data is presented from commissioning and early operation of new plant to that in established PWR circuits. Results arising from reactors transients are also included, which refer to shutdown and start-up phases, power changes and scram situations. The information obtained includes chemical and radiochemical characteristics of particulates and their contribution to total activity carried by reactor coolant. The implications for plant operations are discussed. 16 refs., 55 figs., 24 tabs.

  1. Metal cation inhibitors for controlling denting corrosion in steam generators. Final report. [PWR

    SciTech Connect

    Leidheiser, H. Jr.; Granata, R.D.; Simmons, G.W.; Music, S.; Vedage, H.L.

    1982-12-01

    Metal cations of arsenic, antimony, tin, manganese, zinc, cadmium, indium, and thallium have been evaluated in a preliminary way as possible3 inhibitors for controlling denting corrision observed in steam generators used with pressurized water reactors (PWR). The rationale for this approach was based upon the well-known inhibition effects of metal cations on corrosion rates in electrolyte/metal systems. A review of corrosion inhibition by metal cations (H. Leidheiser, Jr., Corrosion 36, 339 (1982)) has identified eleven inhibition mechanisms. The major test methods used for this evaluation were: (1) Isothermal capsule tests of carbon/steel/Inconel 600 tube bulging rates at temperatures up to 288/sup 0/C in seawater/copper-nickel chloride bulge-accelerating solutions. (2) Immersion weight-loss tests of steel coupled to Inconel 600 in boiling (102/sup 0/C) 3% sodium chloride solutions. In addition, electrochemical measuremens and surface analyses were performed. The major findings of this investigation are presented.

  2. BWR Anticipated Transients Without Scram Leading to Instability

    SciTech Connect

    Cheng L. Y.; Baek J.; Cuadra, A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor power decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).

  3. Comet solutions to a stylized BWR benchmark problem

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2012-07-01

    In this paper, a stylized 3-D BWR benchmark problem was used to evaluate the performance of the coarse mesh radiation transport method COMET. The benchmark problem consists of 560 fuel bundles at 3 different burnups and 3 coolant void states. The COMET solution was compared with the corresponding Monte Carlo reference solution using the same 2-group material cross section library for three control blade (rod) configurations, namely, all rods out (ARO), all rods in (ARI) and some rods in (SRJ). The differences in the COMET and MCNP eigenvalues were 43 pcm, 66 pcm and 32 pcm for the ARO, ARI and SRI cases, respectively. These differences are all within 3 standard deviations of the COMET uncertainty. The average relative differences in the bundle averaged fission densities for these three cases were 0.89%, 1.24%, and 1.05%, respectively. The corresponding differences in the fuel pin averaged fission densities were 1.24%, 1.84% and 1.29%, respectively. It was found that COMET is 3,000 times faster than Monte Carlo, while its statistical uncertainty in the fuel pin fission density is much lower than that of Monte Carlo (i.e., {approx}40 times lower). (authors)

  4. Westinghouse BWR Fuel Reliability - Recent Experience and Analyses

    SciTech Connect

    Ryttersson, Kristina; Helmersson, Sture; Wright, Jonathan; Hallstadius, Lars

    2007-07-01

    Fuel reliability and failure free fuel has always been one of the most important objectives in the development work at Westinghouse Electric Sweden. An important step in tailoring remedies against both primary and secondary fuel failures is to understand the failure mechanisms. Studies of the mechanisms behind both primary and secondary failures have been performed. For primary failures the recent efforts have been focused on debris fretting failures, since this has been the only mechanism that causes failures in Westinghouse BWR fuel for several years. A statistical analysis of debris fretting failures was performed. The results showed a strong dependency on flow velocity which could be related to a working hypothesis coupling to the excitation of vibrations and the pressure drop over an object in a flow. To increase the understanding of the secondary degradation mechanism, two test reactor studies have been performed. Also, trends related to residence time in core, burnup and power have been evaluated based on the Westinghouse fuel failure database. No clear trends could be seen regarding residence time or burnup up to {approx}40 MWd/kgU. Beyond {approx}40 MWd/kgU the secondary degradation seems to be less severe. One trend that could be identified was an increase in the severity of secondary degradation with increasing rod power. (authors)

  5. Calculation of a BWR partial ATWS using RAMONA-3B

    SciTech Connect

    Garber, D.I.; Diamond, D.J.; Cheng, H.S.

    1982-01-01

    The RAMONA-3B code has been used to simulate a boiling water reactor (BWR) transient initiated by the closure of the main steam line isolation valves in which all the control rods in one-half the core fail to scram after reactor trip. The modeling of the nuclear steam supply system included three-dimensional neutron kinetics and parallel hydraulic channels (including a bypass channel). The transient is characterized by an initial pressure spike and then by oscillations in the pressure due to the opening and closing of relief valves. These oscillations in turn affect all thermohydraulic properties in the vessel. The simulation was continued for 7 minutes of reactor time at which point boron began to accumulate in the core. The calculation demonstrates the importance of using three-dimensional neutron kinetics in conjunction with the modeling of the nuclear steam supply system for this type of transient. RAMONA-3B is unique in its ability to do this type of calculation.

  6. Effect of aging on the PWR Chemical and Volume Control System

    SciTech Connect

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K.

    1995-06-01

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  7. Modeling a nuclear reactor for experimental purposes. [PWR

    SciTech Connect

    Berta, V T

    1980-01-01

    The Loss-of-Fluid Test (LOFT) Facility is a scale model of a commercial PWR and is as fully functional and operational as the generic commercial counterpart. LOFT was designed and built for experimental purposes as part of the overall NRC reactor safety research program. The purpose of LOFT is to assess the capability of reactor safety systems to perform their intended functions during occurrences of off-normal conditions in a commercial nuclear reactor. Off-normal conditions arising from large and small break loss-of-coolant accidents (LOCA), operational transients, and anticipated transients without scram (ATWS) were to be investigated. This paper describes the LOFT model of the generic PWR and summarizes the experiments that have been conducted in the context of the significant findings involving the complex transient thermal-hydraulics and the consequent effects on the commercial reactor analytical licensing techniques. Through these techniques the validity of the LOFT model as a scaled counterpart of the generic PWR is shown.

  8. PWR Cross Section Libraries for ORIGEN-ARP

    SciTech Connect

    McGraw, Carolyn; Ilas, Germina

    2012-01-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  9. Design study of long-life PWR using thorium cycle

    SciTech Connect

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-06

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that {sup 231}Pa better than {sup 237}Np as burnable poisons in thorium fuel system. Thorium oxide system with 8%{sup 233}U enrichment and 7.6{approx} 8%{sup 231}Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1%{Delta}k/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53%{Delta}k/k and reduced power peaking during its operation.

  10. Design study of long-life PWR using thorium cycle

    NASA Astrophysics Data System (ADS)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-01

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that 231Pa better than 237Np as burnable poisons in thorium fuel system. Thorium oxide system with 8% 233U enrichment and 7.6˜ 8% 231Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1% Δk/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53% Δk/k and reduced power peaking during its operation.

  11. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    SciTech Connect

    Clayton, Dwight A.; Poore, III, Willis P.

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  12. System Integral Test by BWR Drywell Cooler Applied as Phase-II Accident Management

    SciTech Connect

    Nagasaka, Hideo; Tobimatsu, Toshimi; Tahara, Mika; Yokobori, Seiichi; Akinaga, Makoto

    2002-07-01

    This paper deals with the system interaction performance using the BWR drywell local cooler (DWC) in combination with containment spray as a Japanese Phase-II accident management (AM). By using almost full height simulation test facility (GIRAFFE-DWC) with scaling ratio of 1/600, the system integral tests simulating BWR low pressure vessel failure sequence were accomplished during about 14 hours. In case of DWC application, the containment pressure increase was found milder due to DWC heat removal performance. Initial spray timing was delayed about 3 hours and each spray period was reduced almost by half. It was concluded that the application of a BWR DWC to Phase-II AM measure is quite promising from the point of delaying or preventing the containment venting. (authors)

  13. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    SciTech Connect

    Marshall, William BJ J

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  14. A Program for Risk Assessment Associated with IGSCC of BWR Vessel Internals

    SciTech Connect

    A. G. Ware; D. K. Morton; J. D. Page; M. E. Nitzel; S. A. Eide; T. -Y. Chang

    1999-08-01

    A program is being carried out for the US Nuclear Regulatory Commission (NRC) by the Idaho National Engineering and Environmental Laboratory (INEEL), to conduct an independent risk assessment of the consequences of failures initiated by intergranular stress corrosion cracking (IGSCC) of the reactor vessel internals of boiling water reactor (BWR) plants. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, both singly and in combination with the failures of others, with specific consideration given to potential cascading and common mode effects on system performance. This paper presents a description of the overall program that is underway to modify an existing probabilistic risk assessment (PRA) of the BWR/4 plant to include IGSCC-initiated failures, subsequently to complete a quantitative PRA.

  15. Code System for Best-Estimate Analysis of LOCA in BWR.

    2001-07-23

    Version 00 TRAC-BD1 performs best estimate analyses of loss-of-coolant accidents (LOCA) and other transients in boiling water reactors (BWRs). The program provides LOCA analysis capability for BWRs and for many BWR-related thermal-hydraulic experimental facilities. The program features a three-dimensional treatment of the BWR pressure vessel, a detailed model of a BWR fuel bundle including multi-rod, multi-bundle, radiation heat transfer, and leakage path modeling capability; flow-regime-dependent constitutive equation treatment; reflood tracking capability both for falling filmsmore » and bottom flood quench fronts; and consistent treatment of the entire accident sequence. Dump/restart capabilities are also provided.« less

  16. Technical basis for the initiation and cessation of environmentally-assisted cracking of low-alloy steels in elevated temperature PWR environments

    SciTech Connect

    James, L.A.

    1997-10-01

    The Section 11 Working Group on Flaw Evaluation of the ASME B and PV Code Committee is considering a Code Case to allow the determination of the conditions under which environmentally-assisted cracking of low-alloy steels could occur in PWR primary environments. This paper provides the technical support basis for such an EAC Initiation and Cessation Criterion by reviewing the theoretical and experimental information in support of the proposed Code Case.

  17. Photoelectrochemical protection of stainless alloys from the stress-corrosion cracking in BWR primary coolant environment

    SciTech Connect

    Akashi, Masatsune; Iso-o, Hiroyuki; Kubota, Nobuhiko; Fukuda, Takanori; Ayabe, Muneo; Hirano, Kenji

    1995-12-31

    The feasibility of counteracting or preventing the stress-corrosion cracking in the BWR core internals by the photoelectrochemical method has been examined. For the purpose TiO{sub 2} semiconductor is noted for its capability of photo electrochemically inducing the water-oxidizing anodic reaction in low enough potential domain if supplied with a light of a wavelength shorter than 410 nm. This paper offers an empirical proof by showing that Type 304 stainless steel and Alloy 600 stainless alloy that have been plasma-spray coated with TiO{sub 2} film will do quite well in environments of BWR primary coolant.

  18. Control Rod Pattern Planning of a BWR using Enhanced Nelder-Mead Method

    SciTech Connect

    Yoko Kobayashi; Eitaro Aiyoshi

    2004-07-01

    We propose a new optimization algorithm for the short-term planning of control rod patterns in an operating BWR. This algorithm is based on the enhanced Nelder-Mead simplex method in which convergence ability is improved for constrained problems in several ways. The main characteristic of this approach is it uses continuous values for the axial positions of control rods. Through calculations in an actual BWR plant, we showed that the new algorithm is effective for automation of short-term planning and reduction of the engineer's workload. (authors)

  19. STEALTH: a Lagrange explicit finite difference code for solids, structural, and thermohydraulic analysis. Volume 1B: user's manual - input instructions. Computer code manual. [PWR; BWR

    SciTech Connect

    Hofmann, R.

    1981-11-01

    A useful computer simulation method based on the explicit finite difference technique can be used to address transient dynamic situations associated with nuclear reactor design and analysis. This volume is divided into two parts. Part A contains the theoretical background (physical and numerical) and the numerical equations for the STEALTH 1D, 2D, and 3D computer codes. Part B contains input instructions for all three codes. The STEALTH codes are based entirely on the published technology of the Lawrence Livermore National Laboratory, Livermore, California, and Sandia National Laboratories, Albuquerque, New Mexico.

  20. STEALTH: a Lagrange explicit finite difference code for solids, structural, and thermohydraulic analysis. Volume 2: sample and verification problems. Computer code manual. [PWR; BWR

    SciTech Connect

    Hofmann, R.

    1982-08-01

    STEALTH sample and verification problems are presented to help users become familiar with STEALTH capabilities, input, and output. Problems are grouped into articles which are completely self-contained. The pagination in each article is A.n, where A is a unique alphabetic-character article identifier and n is a sequential page number which starts from 1 on the first page of text for each article. Articles concerning new capabilities will be added as they become available. STEALTH sample and verification calculations are divided into the following general categories: transient mechanical calculations dealing with solids; transient mechanical calculations dealing with fluids; transient thermal calculations dealing with solids; transient thermal calculations dealing with fluids; static and quasi-static calculations; and complex boundary interaction calculations.

  1. STEALTH: a Lagrange explicit finite difference code for solids, structural, and thermohydraulic analysis. Volume 1A: user's manual - theoretical background and numerical equations. Computer code manual. [PWR; BWR

    SciTech Connect

    Hofmann, R.

    1981-11-01

    A useful computer simulation method based on the explicit finite difference technique can be used to address transient dynamic situations associated with nuclear reactor design and analysis. This volume is divided into two parts. Part A contains the theoretical background (physical and numerical) and the numerical equations for the STEALTH 1D, 2D, and 3D computer codes. Part B contains input instructions for all three codes. The STEALTH codes are based entirely on the published technology of the Lawrence Livermore National Laboratory, Livermore, California, and Sandia National Laboratories, Albuquerque, New Mexico.

  2. STEALTH: a Lagrange explicit finite difference code for solids, structural, and thermohydraulic analysis. Volume 7: implicit hydrodynamics. Computer code manual. [PWR; BWR

    SciTech Connect

    McKay, M.W.

    1982-06-01

    STEALTH is a family of computer codes that solve the equations of motion for a general continuum. These codes can be used to calculate a variety of physical processes in which the dynamic behavior of a continuum is involved. The versions of STEALTH described in this volume were designed for the calculation of problems involving low-speed fluid flow. They employ an implicit finite difference technique to solve the one- and two-dimensional equations of motion, written for an arbitrary coordinate system, for both incompressible and compressible fluids. The solution technique involves an iterative solution of the implicit, Lagrangian finite difference equations. Convection terms that result from the use of an arbitrarily-moving coordinate system are calculated separately. This volume provides the theoretical background, the finite difference equations, and the input instructions for the one- and two-dimensional codes; a discussion of several sample problems; and a listing of the input decks required to run those problems.

  3. STEALTH: a Lagrange explicit finite difference code for solids, structural, and thermohydraulic analysis. Volume 3: programmer's manual. Computer code manual. [PWR; BWR

    SciTech Connect

    Hofmann, R.

    1981-11-01

    This volume contains a description of a programming and documentation structure for the STEALTH finite difference computer programs based on general principles applicable to most large scientific computer programs. Program modularization (as well as documentation format) is based entirely on the theoretical elements of analysis of a physical system that were presented in Volume 1. FORTRAN programming and naming conventions are also described. Among the programming formats presented is a FORTRAN manual (Appendix FTN) which can be used as the basis for developing portable codes. STEALTH was developed on a CDC 7600. However, it has been designed so that it can be installed on most large scientific computers. Installation documentation exists for some facilities and can be generated easily for others.

  4. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods; Revision 1

    SciTech Connect

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degrees}C and whether the cladding of the stored spent fuel ever exceeds 350{degrees}C. Limiting the borehole to temperatures of 97{degrees}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degrees}C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 {times} 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degrees}C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350{degrees}C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft {times} 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40{degrees}C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation.

  5. Results of vortex-suppressor tests, single-outlet sump tests and miscellaneous sensitivity tests. Containment sump reliability studies generic task A-43. [PWR; BWR

    SciTech Connect

    Padmanabhan, M.

    1982-09-01

    Full scale tests of flow conditions in Containment Recirculation Sumps for nuclear power stations were conducted at the Alden Research Laboratory (ARL) of Worcester Polytechnic Institute (WPI) to provide sump hydraulic design and performance data for use in resolving the Unresolved Safety Issue, A-43, Containment Sump Performance. This document is a report of the results of investigations conducted as a part of Phase II of the test program, including: (a) vortex suppressor tests to study in detail the hydraulic behavior of two commonly used suppressors; namely, cubic cage and horizontal floor grating; (b) single outlet sump tests to ascertain the hydraulic performance of single outlet sumps compared to double outlet sumps; and (c) tests to study the effects on the hydraulic performance of a solid partition wall in a double outlet sump, pump overspeed (i.e., higher flow), outlet pipe diameter, and bellmouth entrances.

  6. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    SciTech Connect

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  7. Monticello BWR spent fuel assembly decay heat predictions and measurements

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Heeb, C.M.; Creer, J.M.

    1986-06-01

    This report compares pre-calorimetry predictions of rates of six 7 x 7 boiling water reactor (BWR) spent fuel assemblies with measured decay heat rates. The assemblies were from Northern States Power Company's Monticello Nuclear Generating Plant and had burnups of 9 to 21 GWd/MTU and cooling times of 9 to 10 years. Conclusions are: The agreement between ORIGEN2 predictions and decay heat measurements of Monticello spent fuel is dependent on the method used to calibrate the calorimeter and to make the decay heat measurements. The agreement between predictions and measurements of decay heat rates of Monticello fuel is the same as that for Cooper and Dresden fuel if the same measurement method is used. The predictions are within a standard deviation of +-15 W of the measurements. Using a different measurement method, ORIGEN2 underpredicts the measured decay heat output of Monticello fuel assemblies by a constant 20 +- 2 W. The 20-W offset appears to be an artifact of the calibration procedure. The constant term in the calibration curve (i.e., q/sub DH/ = mx + b) can account for measurement differences of 40 W based on the 1983, 1984, and 1985 calibration curves. The difference between ORIGEN2 predictions and calorimeter decay heat measurements does not appear to be dependent on the magnitude of decay heat output. Predicted axial decay heat profiles are in good agreement with measured axial gamma radiation profiles. Recommendations are: Predictions using other decay heat codes should be compared to experimental data contained in this report, to evaluate prediction capabilities. The source of the differences that exist among calorimeter calibration curves needs to be determined. Calorimeter operational methods need to be investigated further to determine cause and effect relationships between operational method and calorimeter precision and accuracy.

  8. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    SciTech Connect

    H. Wang

    1997-01-23

    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: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF 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, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  9. Conditioning of BWR Control - Elements Using the New MOSAIK 80T/SWR-SE Cask - Concept

    SciTech Connect

    Oldiges, O.; Blenski, H.-J.; Engelage, H.; Behrens, W.; Majunke, J.; Schwarz, W.; Hallfarth, Dr.

    2002-02-27

    During the operation of Boiling Water Reactors, Control - Elements are used to control the neutron flux inside the reactor vessel. After the end of the lifetime, the Control - Elements are usually stored in the fuel - elements - pool of the reactor. Up to now, in Germany no conditioning of Control - Elements has been done in a BWR under operation.

  10. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    SciTech Connect

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in

  11. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These maymore » be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section

  12. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect

    J.M. Scaglione

    2004-12-17

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation.

  13. TRU transmutation in thorium-based heterogeneous PWR core

    SciTech Connect

    Bae, Kang-Mok; Lim, Jae-Yong; Kim, Myung-Hyun

    2004-07-01

    A thorium-based seed and blanket design concept for a conventional pressurized light water reactor (PWR) was proposed to enhance the proliferation resistance potential and fuel cycle economics. The KTF core was satisfied with neutronic and thermal-hydraulic design limit of conventional PWR, APR-1400. In order to evaluate transmutation capability of a thorium-based KTF core, U/Zr seed fuel mixed with 10% TRU which come from 1,000 MWe power reactor after 10 years decay was proposed and analyzed by transmutation indices such as D{sub j}, TEX and SR. KTF core showed an extended fuel cycle burnup; average burnup of seed was 79.5 MWd/kgHM and blanket was 94.6 MWd/kgHM. It means that residence time of TRU in the core could be long enough for transmutation when TRU is mixed in seed fuel. The amount of TRU production from conventional PWR could be transmuted in the KTF-TRU core, especially Am-241 isotope is remarkably transmuted by capture reaction. Even isotopes of curium were cumulated in the core during the burnup, however, KTF-TRU core could reduce the TRU in spent fuel by using well-thermalized neutron spectrum. Proliferation resistance potential of thorium based transmutation fuel is slightly increased. About 31% reduction of TRU amount was measured from reduced plutonium production from U-238. Total amount of Am-241 was reduced significantly, but total amount of minor actinide (MA) was reduced by 28% of its initial loading mass. (authors)

  14. Estimating probable flaw distributions in PWR steam generator tubes

    SciTech Connect

    Gorman, J.A.; Turner, A.P.L.

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  15. Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

    SciTech Connect

    S. Mastilovic; A. Scheider; S.M. Bennett

    2001-01-29

    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.

  16. Characterization of PWR steam generator deposits

    SciTech Connect

    Varrin, R. Jr.

    1996-02-01

    Restoring the thermal performance of the steam generators often requires the utility to remove deposits by expensive chemical means. This work demonstrates that careful characterization of secondary side deposit samples can reveal their chemical and physical properties which in turn contribute to an overall assessment of the need for and extent of steam generator inspection and maintenance. More specifically, knowledge of deposit characteristics can contribute to: (1) determination of the source of corrosion products, (2) assessment of feedwater chemistry control strategies, (3) prediction of rates of tube degradation, and (4) evaluation of degraded heat transfer performance or flow instabilities. Despite the relationships between deposits and steam generator operation and performance, few utilities elect to perform the types of characterizations which are suitable for the determination of the specific chemical and physical nature of their particular deposits. One of the principal goals of this document is to encourage utilities to consider deposit characterization an integral part of an overall effort to assess and maintain the material condition of the steam generators at their plant. This document includes a review of the nature of deposits and relates deposit characteristics to a variety of secondary side phenomena including corrosion and fouling. Candidate techniques for revealing relevant deposit properties are provided so that inferences regarding the role of deposits in promoting or causing these phenomena at their plant can be developed.

  17. Enhancing BWR proliferation resistance fuel with minor actinides

    NASA Astrophysics Data System (ADS)

    Chang, Gray S.

    2009-03-01

    To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in

  18. A comprehensive in-pile test of PWR fuel bundle

    NASA Astrophysics Data System (ADS)

    Kang, Rixin; Zhang, Shucheng; Chen, Dianshan

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3 × 3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 μm. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation.

  19. VERA Core Simulator Methodology for PWR Cycle Depletion

    SciTech Connect

    Kochunas, Brendan; Collins, Benjamin S; Jabaay, Daniel; Kim, Kang Seog; Graham, Aaron; Stimpson, Shane; Wieselquist, William A; Clarno, Kevin T; Palmtag, Scott; Downar, Thomas; Gehin, Jess C

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  20. 21-PWR Waste Package Side and End Impacts

    SciTech Connect

    V. Delabrosse

    2003-02-27

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  1. 21-PWR Waste Package Side and End Impacts

    SciTech Connect

    T. Schmitt

    2005-08-29

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  2. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    SciTech Connect

    Arai, Kenji; Ebata, Shigeo

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  3. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    NASA Astrophysics Data System (ADS)

    Trianti, Nuri; Su'ud, Zaki; Arif, Idam; Riyana, EkaSapta

    2014-09-01

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  4. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    SciTech Connect

    Trianti, Nuri E-mail: szaki@fi.itba.c.id; Su'ud, Zaki E-mail: szaki@fi.itba.c.id; Arif, Idam E-mail: szaki@fi.itba.c.id; Riyana, EkaSapta

    2014-09-30

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  5. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    SciTech Connect

    Ade, Brian J; Marshall, William BJ J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez-Gonzalez, Jesus S

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  6. Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez, J. S.

    2015-01-01

    [Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

  7. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    SciTech Connect

    Wang, Jy-An John; Jiang, Hao

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  8. Enhancing BWR Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect

    Gray S. Chang

    2008-07-01

    reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm3) to the top (0.35 g/cm3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. We concluded that the concept of MARA, which involves the use of transuranic nuclides (237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy rennaissance.

  9. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

    SciTech Connect

    A. K. MAJI; B. MARSHALL; ET AL

    2000-10-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  10. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    SciTech Connect

    Ade, Brian J; Marshall, William BJ J; Martinez-Gonzalez, Jesus S

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  11. Benchmark Validation of Tort Code Using Kkm Measurement and its Application to 800 Mwe Bwr

    NASA Astrophysics Data System (ADS)

    Tsukiyama, Toshihisa; Hayashi, Katsumi; Kurosawa, Masahiko; Hayashida, Yoshihisa; Asano, Kyoichi; Koyabu, Ken

    2003-06-01

    To estimate the applicability of the TORT code, a benchmark calculation was performed using the measured neutron flux in a 375MWe BWR in Switzerland. The calculated neutron flux was compared with the measured neutron fluxes at 27 locations between the shroud and the RPV. The reaction rates of thermal and fast dosimeters calculated by TORT agreed well with the measured data. As a next step, the TORT code was applied to estimate the neutron flux distribution in Japanese 800MWe BWR plants and compared with the measured radioactivity of a few pieces of the top guide beam, shroud and in-core monitor guide tube. Because a reasonable C/M value was obtained, we conclude that we can obtain reasonable neutron distribution profiles with TORT.

  12. Estimating boiling water reactor decommissioning costs. A user`s manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    SciTech Connect

    Bierschbach, M.C.

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  13. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    SciTech Connect

    Bierschbach, M.C.

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  14. Modeling local chemistry in PWR steam generator crevices

    SciTech Connect

    Millett, P.J.

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  15. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    SciTech Connect

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  16. BWR reactor pressure vessel license renewal industry report; revision 1. Final report

    SciTech Connect

    Braden, D.; Stancavage, P.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) Reactor Pressure Vessels (RPVs). The report includes the following RPV components: attachment welds, closure studs, nozzles and safe ends, penetrations, vessel shell and flanges, top and bottom heads and vessel support skirt. The scope is limited to domestic BWRs designated by GE Nuclear Energy as BWR/2 through BWR/6.

  17. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    SciTech Connect

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data.

  18. Bifurcation Analysis of Nuclear-Coupled Thermal Hydraulics of BWR Using BIFDD

    SciTech Connect

    Zhou, Quan; Uddin, Rizwan

    2002-07-01

    Stability and bifurcation analyses of nuclear-coupled thermal hydraulic instability in BWR has been performed using a semi-analytical method. The BWR model used in this study consists of three parts: neutron kinetics, fuel rod heat conduction and single and two-phase heated channel thermal hydraulics. Point reactor model is currently being used for neutron kinetics and will be extended in the future to higher order lambda or omega-mode. In the heat conduction part, a piecewise quadratic approximation to radial temperature distribution in fuel pellet and cladding is assumed. ODEs for the expansion coefficients of the quadratic spatial profiles are developed by applying variational principle. Similar to the heat conduction model, the spatial enthalpy distribution in the single phase region and steam quality in the two-phase region in the BWR core are approximated by quadratic polynomials. Two-phase flow is modeled using the homogeneous equilibrium model. A bifurcation analysis code, BIFDD, is then used to perform the analysis for the stability boundary (SB) and the nature of Poincar Andronov-Hopf bifurcation (PAH-B). Results in control-rod-induced-reactivity inlet-subcooling-number space show that both super or sub-critical bifurcation can occur along the SB he subcritical bifurcation occurs for very small or very large subcooling number values; super-critical PAH-B occurs for intermediate values of subcooling number. (authors)

  19. Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

    SciTech Connect

    1996-03-01

    On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff`s basis for issuing GL 94-03, as well as the staff`s assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date.

  20. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    SciTech Connect

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

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

    SciTech Connect

    Turner, S.E.; Gurley, M.K.; Kirby, K.D.; Mitchell, W III

    1981-09-15

    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.

  2. Identification and evaluation of PWR in-vessel severe accident management strategies

    SciTech Connect

    Dukelow, J S; Harrison, D G; Morgenstern, M

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  3. Impact of makeup water system performance on PWR steam generator corrosion. Final report

    SciTech Connect

    Bell, M.J.; Pearl, W.L.; Sawochka, S.G.; Smith, L.A.

    1985-06-01

    The objectives of this project were to review makeup system design and performance and assess the possible relation of pressurized water reactor (PWR) steam generator corrosion to makeup water impurity ingress at fresh water sites. Project results indicated that makeup water transport of most ionic impurities can be expected to have a significant impact on secondary cycle chemistry only if condenser inleakage and other sources of impurities are maintained at very low levels. Since makeup water oxygen control techniques at most study plants were not consistent with state-of-the-art technology, oxygen input to the cycle via makeup can be significant. Leakage of colloidal silica and organics through makeup water systems can be expected to control blowdown silica levels and organic levels throughout the cycle at many plants. Attempts to correlate makeup water quality to steam generator corrosion observations were unsuccessful since (1) other impurity sources were significant compared to makeup at most study plants, (2) many variables are involved in the corrosion process, and (3) in the case of IGA, the variables have not been clearly established. However, in some situations makeup water can be a significant source of contaminants suspected to lead to both IGA and denting.

  4. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    NASA Astrophysics Data System (ADS)

    Ratnayake, Ruwan Kumara

    While best estimate computer codes provide the licensing basis for nuclear power facilities, they also serve as analytical tools in overall plant and component design procedures. An ideal best estimate code would comprise of universally applicable mechanistic models for all its components. However, due to the limited understanding in these specific areas, many of the models and correlations used in these codes reflect high levels of empiricism. As a result, the use of such models is strictly limited to the range of parameters within which the experiments have been conducted. Disagreements between best estimate code predictions and experimental results are often explained by the mechanistic inadequacies of embedded models. Significant mismatches between calculated and experimental critical power values are common observations in the analyses of Boiling Water Reactors (BWR). Based on experimental observations and calculations, these mismatches are attributed to the additional entrainment and deposition caused by spacer grids in BWR fuel assemblies. In COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid); a state of the art industrial best estimate code, these disagreements are hypothesized to occur due the absence of an appropriate spacer grid model. In this thesis, development of a suitably detailed spacer grid model and integrating it to COBRA-TF is documented. The new spacer grid model is highly mechanistic so that the applicability of it is not seriously affected by geometric variations in different spacer grid designs. COBRA-TF (original version) simulations performed on single tube tests and BWR rod bundles with spacer grids showed that single tube predictions were more accurate than those of the rod bundles. This observation is understood to arise from the non-availability of a suitable spacer grid model in COBRA-TF. Air water entrainment experiments were conducted in a test section simulating two adjacent BWR sub channels to visualize the flow behavior at

  5. Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids

    SciTech Connect

    Ratnayake, Ruwan K.; Hochreiter, L.E.; Ivanov, K.N.; Cimbala, J.M.

    2002-07-01

    Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained from operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces

  6. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    SciTech Connect

    Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  7. Reviews

    NASA Astrophysics Data System (ADS)

    2001-11-01

    BOOK REVIEW Search for Life BOOK REVIEW Health Physics BOOK REVIEW Language and Literacy in Science Education BOOK REVIEW Science Web Reader—Physics Correction GCSE BOOK REVIEW Physics for Higher Tier GCSE BOOK REVIEW Modular Science GCSE BOOK REVIEW Modular Science for AQA: Foundation level and Higher level GCSE BOOK REVIEW Physics for OCR A GCSE BOOK REVIEW Physics Matters, 3rd edition GCSE BOOK REVIEW Physics GCSE BOOK REVIEW Science Foundations: Physics (new edition) GCSE BOOK REVIEW Target Science: Physics Foundation Tier GCSE BOOK REVIEW Target Science: Physics Foundation Tier: AQA WEB WATCH Medical physics organizations

  8. Barium silicate glass/Inconel X-750 interaction. [PWR

    SciTech Connect

    Kelsey, Jr., P. V.; Siegel, W. T.; Miley, D. V.

    1980-01-01

    Water reactor safety programs at the Idaho National Engineering Laboratory have required the development of specialized instrumentation. An example is the electrical conductivity-sensitive liquid level transducer developed for use in pressurized-water reactors (PWRs) in which the operation of the sensing probe relies upon the passage of current through the water between the center pin of the electrode and its shell such that when water is present the resulting voltage is low, and conversely, when water is absent the voltage is high. The transducer's ceramic seal is a hot-pressed glass ceramic; its metal housing is Inconel X-750. The ceramic material provides an essential dielectric barrier between the center pin and the outer housing. The operation of the probe as well as the integrity of the PWR environment requires a hermetically-bonded seal between the ceramic and the metal. However, during testing, an increasing number of probe assemblies failed owing to poor glass-to-metal seals as well as void formation within the ceramic. Therefore, a program was initiated to characterize the metallic surface with respect to pre-oxidation treatment and determine optimum conditions for wetting and bonding of the metal by the glass to obtain baseline data relevant to production of acceptable transducer seals.

  9. Development of large-capacity main steam isolation valves and safety relief valves for next-generation BWR plant

    SciTech Connect

    Mitsugu Nishimura; Shin-ichi Furukawa; Gen Itoh; Kikuo Takeshima

    2002-07-01

    A study was made of high capacity main steam isolation valves (MSIV) and safety relief valves (SRV) for the main steam line of a boiling water reactor (BWR). The next-generation BWR plants, which are planned to have higher thermal power, have raised concerns relating to the main steam line of an increase in maintenance work to SRVs and erosion of the MSIV valve seat due to the increased main steam flow velocity. In this research project, the capacity of the MSIV and SRV was increased and the valve configuration was changed in an attempt to solve these problems. (authors)

  10. A theoretical and numerical investigation of turbulent steam jets in BWR steam blowdown.

    SciTech Connect

    NguyenLe, Q.

    1998-06-26

    The preliminary results of PHOENICS and RELAP5 show that the current numerical models are adequate in predicting steam flow and stratification patterns in the upper Drywell of a BWR containment subsequent to a blow-down event. However, additional modeling is required in order to study detailed local phenomena such as condensation with non-condensables, natural convection, and stratification effects. Analytically, the intermittence modified similarity solutions show great promise. Once {gamma} is accounted for, the jet's turbulent shear stress can be determined with excellent accuracy.

  11. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    SciTech Connect

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  12. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    NASA Astrophysics Data System (ADS)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  13. Simulating the Effect on Criticality of Simultaneous Matrix Degradation and Assembly Collapse for the 21 PWR Waste Package

    SciTech Connect

    A.A. Alsaed

    1999-09-23

    The purpose of this calculation is to evaluate the effects of fission products loss on the reactivity of commercial pressurized water reactor (PWR) spent nuclear fuel (SNF) in 21 PWR waste packages (WPs) in the event of simultaneous fuel matrix degradation and assembly collapse.

  14. Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

    SciTech Connect

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z.; Sekimoto, H.

    2010-06-22

    A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required {sup 233}U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium and uranium confinement in PWR.

  15. The effects of aging on BWR core isolation cooling systems

    SciTech Connect

    Lee, B.S.

    1994-10-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling (RCIC) system in commercial Boiling Water Reactors (BWRs). This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The failure data from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failures causes. Current inspection, surveillance, and monitoring practices were also reviewed.

  16. WG-MOX Fuel Zr-tube Neutron Spectrum Comparison in ATR and PWR

    SciTech Connect

    Gray S. Chang

    2005-02-01

    An experiment containing WG-MOX fuel has been designed and irradiated from 1998 to 2004 in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). Important neutronics parameters were computed using novel Monte Carlo methods. The purpose of this summary is to compare the Weapons-Grade Mixed Oxide fuel (WG-MOX) Zr-tube’s neutron spectrum in ATR and PWR. The results indicate that the Zrtube’s neutron spectrum in ATR are softer than in PWR.

  17. WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

    SciTech Connect

    J.W. Davis

    1996-07-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so.

  18. Categorization of PWR accident sequences and guidelines for fault trees: seismic initiators

    SciTech Connect

    Kimura, C.Y.

    1984-09-01

    This study developed a set of dominant accident sequences that could be applied generically to domestic commercial PWRs as a standardized basis for a probabilistic seismic risk assessment. This was accomplished by ranking the Zion 1 accident sequences. The pertinent PWR safety systems were compared on a plant-by-plant basis to determine the applicability of the dominant accident sequences of Zion 1 to other PWR plants. The functional event trees were developed to describe the system functions that must work or not work in order for a certain accident sequence to happen, one for pipe breaks and one for transients.

  19. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    SciTech Connect

    Trianti, N.; Su'ud, Z.; Riyana, E. S.

    2012-06-06

    A preliminary design study for the utilization of thorium added with {sup 231}Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of {sup 233}U to {sup 231}Pa in burn-up process. Optimizations of the content of {sup 231}Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 {approx} 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  20. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    NASA Astrophysics Data System (ADS)

    Trianti, N.; Su'ud, Z.; Riyana, E. S.

    2012-06-01

    A preliminary design study for the utilization of thorium added with 231Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of 233U to 231Pa in burn-up process. Optimizations of the content of 231Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 ˜ 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  1. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    SciTech Connect

    Hodge, S.A.

    1991-04-15

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR (boiling water reactor) in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed.

  2. Thermal Response of the 44-BWR Waste Package to a Hypothetical Fire Accident

    SciTech Connect

    J.R. Smotrel; H. Marr; M.J. Anderson

    2001-04-05

    The purpose of this calculation is to determine the thermal response of the 44-boiling water reactor (BWR) waste package (WP) to the hypothetical regulatory fire accident. The objective is to calculate the temperature response of the waste package materials to the hypothetical short-term fire defined in 10 CFR 7 1, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the BWR thermal load are determined. The associated activity is the development of engineering evaluations to support the Licensing Application (LA) design activities.

  3. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    SciTech Connect

    Cheng L. Y.; Baek J.; Cuadra,A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  4. Final results of the XR2-1 BWR metallic melt relocation experiment

    SciTech Connect

    Gauntt, R.O.; Humphries, L.L.

    1997-08-01

    This report documents the final results of the XR2-1 boiling water reactor (BWR) metallic melt relocation experiment, conducted at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The objective of this experiment was to investigate the material relocation processes and relocation pathways in a dry BWR core following a severe nuclear reactor accident such as an unrecovered station blackout accident. The imposed test conditions (initial thermal state and the melt generation rates) simulated the conditions for the postulated accident scenario and the prototypic design of the lower core test section (in composition and in geometry) ensured that thermal masses and physical flow barriers were modeled adequately. The experiment has shown that, under dry core conditions, the metallic core materials that melt and drain from the upper core regions can drain from the core region entirely without formation of robust coherent blockages in the lower core. Temporary blockages that suspended pools of molten metal later melted, allowing the metals to continue draining downward. The test facility and instrumentation are described in detail. The test progression and results are presented and compared to MERIS code analyses. 6 refs., 55 figs., 4 tabs.

  5. TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

    DOE PAGESBeta

    L. Y. Cheng; Baek, J. S.; Cuadra, A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2016-06-06

    A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculationmore » pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.« less

  6. BWR fuel design options for self-sustainable Th-{sup 233}U fuel cycle

    SciTech Connect

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2012-07-01

    In this work, we investigate a number of fuel assembly design options for a BWR core operating in a closed self-sustainable Th-{sup 233}U fuel cycle. The designs rely on axially heterogeneous fuel assembly structure in order to improve fertile to fissile conversion ratio. One of the main assumptions of the current study was to restrict the fuel assembly geometry to a single axial fissile zone 'sandwiched' between two fertile blanket zones. The main objective was to study the effect of the most important design parameters, such as dimensions of fissile and fertile zones and average void fraction, on the net breeding of {sup 233}U. The main design challenge in this respect is that the fuel breeding potential is at odds with axial power peaking and therefore limits the maximum achievable core power rating. The calculations were performed with BGCore system, which consists of MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly with reflective radial boundaries was modeled applying simplified restrictions on maximum central line fuel temperature and Critical Power Ratio. It was found that axially heterogeneous fuel assembly design with single fissile zone can potentially achieve net breeding. In this case however, the achievable core power density is roughly one third of the reference BWR core. (authors)

  7. BWRVIP-101: BWR Vessel and Internals Project: Proceedings: BWRVIP Symposium, Orlando, Florida, December 6-7, 2001

    SciTech Connect

    2002-04-01

    The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on BWR vessel and internals issues. This BWRVIP symposium--held December 6-7, 2001--provided an overview of products completed to date and how they are being implemented at individual plants.

  8. RELAP5-3D Analysis of Pressure Perturbation at the Peach Bottom BWR During Low-Flow Stability Tests

    SciTech Connect

    Lombardi Costa, Antonella; Petruzzi, Alessandro; D'Auria, Francesco

    2006-07-01

    Experimental and theoretical studies about the BWR (Boiling Water Reactor) stability have been performed to design a stable core configuration. BWR instabilities can be caused by inter-dependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In the present work, the pressure perturbation is considered in order to study in detail this type of transient. To simulate this event, including the strong feedback effects between core neutronic and reactor thermal-hydraulics, and to verify core behavior and evaluate parameters related to safety, RELAP5-3D code has been used in the analyses. The simulation was performed making use of Peach Bottom-2 BWR data to predict the dynamics of a real reactor during this type of event. Stability tests were conducted in the Peach Bottom 2 BWR, in 1977, and were done along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. The calculated results are herein compared against the available experimental data. (authors)

  9. Reviews

    NASA Astrophysics Data System (ADS)

    2004-03-01

    WEB WATCH (204) Try unearthing some interesting information about archaeology BOOK REVIEWS (206) Teaching and assessing practical skills Book Review: Learn to drive with Sir Isaac Newton DVD REVIEW (207) Bring some sunshine into the classroom EQUIPMENT REVIEWS (208) Robust air puck takes a kicking Flowlog offers sensing options plus multimode datalogging Mastering Chladni figures takes practice but it offers surprises

  10. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE MULTI-PURPOSE CANISTER (MPC) WITH ACD DISPOSAL CONTAINER (SCPB: N/A)

    SciTech Connect

    T.L. Lotz

    1995-11-13

    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: (1) Characterize the conceptual 40 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond a concern that the long-term disposal thermal issues for the Multi-Purpose Canister (MPC) Subsystem Design, if used with SNF designed for a MOX fuel cycle, do not preclude MPC compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual MPC design with disposal container which is loaded with BWR MOX SNF 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, to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded MPC performance is similar to an MPC loaded with commercial BWR SNF. Future design efforts will focus on specific MPC vendor designs and BWR MOX SNF designs when they become available.

  11. Crack growth rates of nickel alloy welds in a PWR environment.

    SciTech Connect

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.; Energy Technology

    2006-05-31

    In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of {approx}5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

  12. Performance evaluation of two-stage fuel cycle from SFR to PWR

    SciTech Connect

    Fei, T.; Hoffman, E.A.; Kim, T.K.; Taiwo, T.A.

    2013-07-01

    One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with an average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)

  13. Effect of hydrogen injection on hydrogen uptake by BWR fuel cladding. Final report

    SciTech Connect

    Cox, B.

    1983-06-01

    The hydrogen uptake rates reported for zirconium alloys in BWRs, PWRs, HWRs and various experimental in-reactor loops have been surveyed. The scatter in the data is large and arises from a variety of sources: variability in material properties; variations in temperature, irradiation flux, and water chemistry at different points in the reactor; and a lack of accurate knowledge of the sources of hydrogen which ends up in the zirconium-alloy cladding. An attempt has been made to assess the significance of these sources of variability in order to estimate the baseline for hydrogen uptake from the outside of good-quality cladding under normal BWR operating conditions. The probable effect of continuous hydrogen injection on this baseline has been estimated. It is concluded that severe localized hydrogen uptake, which might lead to hydride blisters is very improbable, but that an increase in the uniform rate of hydrogen absorption at the outside of the cladding may be expected.

  14. Study on acoustic resonance and its damping of BWR steam dome

    SciTech Connect

    Ohtsuka, Masaya; Fujimoto, Kiyoshi; Takahashi, Shirou; Hirokawa, Fumihito; Tsubaki, Masaaki

    2006-07-01

    Acoustic resonance characteristics in a BWR steam dome are investigated analytically and experimentally to evaluate the acoustic vibration of a steam dryer. Acoustic modes and frequencies of the ABWR, which represents the BWRs in this study, are calculated by the SYSNOISE code. The lowest mode (32 Hz) is a half stand wave anti-symmetric mode to the center line of the steam dome at normal condition. Acoustic pressure distributions and phases on the steam dryer surface are analyzed for evaluating the vibration driving force of the structure. Acrylic 1/11 scale model tests are performed to verify the acoustic analysis and to develop the acoustic damping system. The experimental frequencies and modes agree with analysis ones for low frequencies. Experimentally, the acoustic pressure amplitude is significantly lowered using the Helmholtz resonators after tuning up the acoustic resonant frequency of the resonator to the acoustic resonant frequency of the main system. (authors)

  15. Equipment environmental monitoring: Perspective from the BWR license renewal lead plant

    SciTech Connect

    Bailey, T.L.

    1991-06-01

    Northern States Power`s Monticello Nuclear Plant is the BWR License Renewal Lead Plant. NSP is currently evaluating plant components and programs such that a license renewal application may be submitted by the end of 1991. As part of the justification to extend the license of the plant, NSP will be required to demonstrate for components important to license renewal that either: evaluations show degradation of the component is not significant during the license renewal term or demonstrate that programs are in place which manage potentially significant degradation mechanisms. This paper identifies NSP`s perspective of the environmental monitoring activities which are expected to be utilized by the project to support the technical evaluations.

  16. The BWR lower head response during a large-break LOCA with core damage

    SciTech Connect

    Alammar, M.A.

    1996-12-31

    Some of the important issues in severe accident management guidelines development deal with estimating the time to lower head vessel failure after core damage and the time window available for water injection that would prevent vessel failure. These issues are obviously scenario dependent, but bounding estimates are needed. The scenario chosen for this purpose was a design-basis accident (DBA) loss-of-coolant accident (LOCA) because it was one of the contributors to the Oyster Creek containment failure frequency. Oyster Creek is a 1930-MW(thermal) boiling water reactor (BWR)-2. The lower head response models have improved since the Three Mile Island unit 2 (TMI-2) vessel investigation project (VIP) results became known, specifically the addition of rapid- and slow-cooling models. These mechanisms were found to have taken place in the TMI-2 lower head during debris cooldown and were important contributors in preventing vessel failure.

  17. Evaluation of a passive containment cooling system for a simplified BWR (boiling water reactor)

    SciTech Connect

    Otonari, J.; Arai, K. ); Oikawa, H.; Nagasaka, H. )

    1989-11-01

    Simplified boiling water reactors (BWRs) are characterized for the adoption of a passive containment cooling system (PCCS) and a passive emergency core cooling system (ECCS). TOSPAC, which had been developed as the preliminary design code for several PCCS concepts, was compared with TRAC for verification. TOSPAC analyses were also performed to show the effectiveness of the isolation condenser (IC) as a PCCS over a wide range of break spectra. The selected reference plant for the analysis is a natural circulation BWR plant with 1,800-MW(thermal) power. The ECCS consists of a gravity-driven cooling system (GDCS) and depressurization valves. The IC and drywell cooler are considered for the PCCS. The IC units and drywell coolers are placed in the IC pool and GDCS pool, respectively.

  18. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    SciTech Connect

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  19. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    SciTech Connect

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z.

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  20. Calculation of MCPR (minimum critical power ratio) for BWR transients using the BNL plant analyzer

    SciTech Connect

    Horak, W.C.; Diamond, D.J.

    1987-06-01

    The critical power ratio (CPR) is used for determining the thermal limits of boiling water reactors. In this study, critical power ratios for a series of transients run on the Brookhaven Plant Analyzer (BPA) (1) have been calculated. The transients include nominal base case simulations, simulations with variations in relief valve setpoints and the number of failed feedwater heaters, simulations at the 100% power, 75% flow point on the extended load line of the MEOD, and a simulation with partial feedwater heating. The plant represented with the BPA is a BWR/4 rated at 3293 MW with a 6.38 m (251'') vessel. Data were obtained by the Plant Analyzer Development Group at BNL from a variety of sources describing the Browns Ferry Plant.

  1. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    SciTech Connect

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  2. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    SciTech Connect

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  3. Peach Bottom Turbine Trip Simulations with RETRAN Using INER/TPC BWR Transient Analysis Method

    SciTech Connect

    Kao Lainsu; Chiang, Show-Chyuan

    2005-03-15

    The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations are described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT

  4. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    SciTech Connect

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  5. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    SciTech Connect

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  6. Comparative assessment of selected PWR auxiliary feedwater system reliability analyses

    SciTech Connect

    Youngblood, R.; Fresco, A.; Papazoglou, I.A.; Tsao, J.

    1985-01-01

    This paper presents a sample of results obtained in reviewing utility submittals of Auxiliary Feedwater System reliability studies. These results are then used to illustrate a few general points regarding such studies. The submittals and reviews for operating license applications are quite significant in that they represent an application of probabilistic risk assessment techniques in the licensing process.

  7. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  8. Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant.

    PubMed

    Ródenas, J; Abarca, A; Gallardo, S

    2011-08-01

    BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool.

  9. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    SciTech Connect

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken.

  10. Reviews

    ERIC Educational Resources Information Center

    Schodde, P.; Ed.

    1976-01-01

    Reviews 17 books and curriculum materials of interest to secondary science teachers. Topics include plant science, pollution, fishes, science investigations, general zoology, neurobiology, electronics, and the environment. (MLH)

  11. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    SciTech Connect

    Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.; Gauld, Ian C.; Ilas, Germina; Marshall, William BJ J.

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  12. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis: User's guide

    SciTech Connect

    Rettig, W.H.; Wade, N.L. )

    1992-06-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODI version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MOD1 version produces results consistent with previous versions. Assessment calculations using the two TRAC-BFI versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  13. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis, Model description

    SciTech Connect

    Borkowski, J.A.; Wade, N.L.; Giles, M.M.; Rouhani, S.Z.; Shumway, R.W.; Singer, G.L.; Taylor, D.D.; Weaver, W.L. )

    1992-08-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODl version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MODI version produces results consistent with previous versions. Assessment calculations using the two TRAC-BF1 versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  14. Optimization of small long-life PWR based on thorium fuel

    SciTech Connect

    Subkhi, Moh Nurul; Suud, Zaki Waris, Abdul; Permana, Sidik

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

  15. Safety analysis of B and W Standard PWR using thorium-based fuels

    SciTech Connect

    Uotinen, V.O.; Carroll, W.P.; Jones, H.M.; Toops, E.C.

    1980-06-01

    A study was performed to assess the safety and licenseability of the Babcock and Wilcox standard 205-fuel assembly PWR when it is fueled with three types of thoria-based fuels denatured (/sup 233/U//sup 238/U-Th)O/sub 2/, denatured (/sup 235//U/sup 238/U-Th)O/sub 2/, and (Th-Pu)O/sub 2/. Selected transients were analyzed using typical PWR safety analysis calculational methods. The results support the conclusion that it is feasible from a safety standpoint to utilize either of the denatured urania-thoria fuels in the standard B and W plant. In addition, it appears that the use of thoria-plutonia fuels would probably also be feasible. These tentative conclusions depend on a data that is more limited than that available for UO/sub 2/ fuels.

  16. Optimization of small long-life PWR based on thorium fuel

    NASA Astrophysics Data System (ADS)

    Subkhi, Moh Nurul; Suud, Zaki; Waris, Abdul; Permana, Sidik

    2015-09-01

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% 233U & 2.8% 231Pa, 6% 233U & 2.8% 231Pa and 7% 233U & 6% 231Pa give low excess reactivity.

  17. Conceptual design study of small long-life PWR based on thorium cycle fuel

    NASA Astrophysics Data System (ADS)

    Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

    2014-09-01

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higer conversion ratio in thermal region compared to uranium cycle produce some significant of 233U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation.

  18. Conceptual design study of small long-life PWR based on thorium cycle fuel

    SciTech Connect

    Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

    2014-09-30

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation.

  19. MC21 analysis of the MIT PWR benchmark: Hot zero power results

    SciTech Connect

    Kelly Iii, D. J.; Aviles, B. N.; Herman, B. R.

    2013-07-01

    MC21 Monte Carlo results have been compared with hot zero power measurements from an operating pressurized water reactor (PWR), as specified in a new full core PWR performance benchmark from the MIT Computational Reactor Physics Group. Included in the comparisons are axially integrated full core detector measurements, axial detector profiles, control rod bank worths, and temperature coefficients. Power depressions from grid spacers are seen clearly in the MC21 results. Application of Coarse Mesh Finite Difference (CMFD) acceleration within MC21 has been accomplished, resulting in a significant reduction of inactive batches necessary to converge the fission source. CMFD acceleration has also been shown to work seamlessly with the Uniform Fission Site (UFS) variance reduction method. (authors)

  20. Materials Reliability Program: Environmental Fatigue Testing of Type 304L Stainless Steel U-Bends in Simulated PWR Primary Water (MRP-137)

    SciTech Connect

    R.Kilian

    2004-12-01

    Laboratory data generated in the past decade indicate a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. In recent comprehensive review of laboratory, component and structural test data performed through the EPRI Materials Reliability Program, flow rate was identified as a critical variable that was generally not considered in laboratory studies but applicable in plant operating environments. Available data for carbon/low-alloy steel piping components suggest that high flow is beneficial regarding the effects of a reactor water environment. Similar information is lacking for stainless steel piping materials. This report documents progress made to date in an extensive testing program underway to evaluate the effects of flow rate on the corrosion fatigue of 304L stainless steel under simulated PWR primary water environmental conditions.

  1. Estimating pressurized water reactor decommissioning costs: A user`s manual for the PWR Cost Estimating Computer Program (CECP) software. Draft report for comment

    SciTech Connect

    Bierschbach, M.C.; Mencinsky, G.J.

    1993-10-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the US Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personnel computer, provides estimates for the cost of decommissioning PWR plant stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  2. Reviews.

    ERIC Educational Resources Information Center

    Journal of Chemical Education, 1988

    1988-01-01

    Reviews three computer software packages for Apple II computers. Includes "Simulation of Hemoglobin Function,""Solution Equilibrium Problems," and "Thin-Layer Chromatography." Contains ratings of ease of use, subject matter content, pedagogic value, and student reaction according to two separate reviewers for each program. (CW)

  3. Reviews

    NASA Astrophysics Data System (ADS)

    2002-09-01

    CD-ROM REVIEWS (449) It's Physics Furry Elephant: Electricity Explained BOOK REVIEWS (450) What Are the Chances? Voodoo Deaths, Office Gossip and Other Adventures in Probability Dictionary of Mechanics: A handbook for teachers and students Intermediate 2 Physics PLACES TO VISIT (452) Spaceguard Centre WEB WATCH (455) Risk

  4. Contain analysis of hydrogen distribution and combustion in PWR dry containments

    SciTech Connect

    Yang, J.W.; Nimnual, S.

    1991-01-01

    Hydrogen transport and combustion in a PWR dry containment are analyzed using the CONTAIN code for a multi-compartment model of the Zion plant. The analysis includes consideration of both degraded core and full core meltdown accidents initiated by a small break LOCA. The importance of intercell flow mixing on distributions of gas composition and temperature in various compartments are evaluated. Thermal stratification and combustion behavior are discussed. 4 refs., 8 figs., 2 tabs.

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

    SciTech Connect

    Cardoni, Jeffrey

    2010-11-01

    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.

  6. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    SciTech Connect

    Naegeli, Robert Earl

    2004-06-01

    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  7. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    NASA Astrophysics Data System (ADS)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  8. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    SciTech Connect

    McGraw, C.; Ilas, G.

    2012-07-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

  9. PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

    SciTech Connect

    Schwirian, R.; Robison, G.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal.

  10. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Mertyurek, Ugur; Radulescu, Georgeta

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  11. Steam Line Break and Station Blackout Transients for Proliferation Resistant Hexagonal Tight Lattice BWR

    SciTech Connect

    Upendra Rohatgi; Jae Jo; Bub Dong Chung; Hiroshi Takahashi; Downar, T.J.

    2002-07-01

    Safety analyses of a proliferation resistant, economically competitive, high conversion, boiling water reactor (HCBWR) fueled with fissile plutonium and fertile thorium oxide fuel elements, and with passive safety systems are presented here. The HCBWR developed here is characterized by a very tight lattice with a relatively small water volume fraction in the core which therefore operates with a fast reactor neutron spectrum, and a considerably improved neutron economy compared to the current generation of Light Water Reactors. A tight lattice BWR core has very narrow flow channels with a hydraulic diameter less than half of the regular BWR core. The tight lattice core presented a special challenge to core cooling, because of reduced water inventory and high friction in the core. The primary safety concern when reducing the moderator to fuel ratio and when using a tightly packed lattice arrangement is to maintain adequate cooling of the core during both normal operation and accident scenarios. In the preliminary HCBWR design, the core has been placed in a vessel with a large chimney section, and the vessel is connected with an Isolation Cooling System (ICS). The vessel is placed in a containment with a Gravity Driven Cooling System (GDCS) and a Passive Containment Cooling System (PCCS) in a configuration similar to General Electric's (GE) Simplified Boiling Water Reactor (SBWR). The safety systems are similar to the SBWR; the ICS and PCCS are scaled with power. An internal recirculation pump was placed in the downcomer to augment the buoyancy head provided by the chimney. The buoyancy provided by the chimney alone could not generate sufficient recirculation in the vessel since the tight lattice configuration resulted in much larger friction in the core than the SBWR. A modified RELAP5 Code was used to simulate and analyze two of the most limiting events for a tight pitch lattice core: the Station Blackout and the Main Steam Line Break events. The constitutive

  12. Reviews.

    ERIC Educational Resources Information Center

    Journal of Chemical Education, 1987

    1987-01-01

    Provides a review of both the Apple and IBM versions of ENZPACK, a software package which is designed to assist in the teaching of enzyme kinetics in courses where this topic is treated in some depth. (TW)

  13. Reviews.

    ERIC Educational Resources Information Center

    Journal of Chemical Education, 1989

    1989-01-01

    Reviews two chemistry software packages: (1) "Organic Reaction Chemistry" (organic chemistry, college level, Apple II); and (2) "Chemical Reactions, Reactions in Aqueous Solution, and Oxidation Reduction Reactions" (general chemistry, college level, IBM). (MVL)

  14. Reviews.

    ERIC Educational Resources Information Center

    Greenleaf, Floyd; And Others

    1986-01-01

    Reviews eight textbooks, readers, and books. Topics include Latin America, colonial America, the Carolinians, women in French textbooks, the Vikings, the Soviet Union, nineteenth-century Black America, and Ernest Rutherford. (TRS)

  15. Reviews.

    ERIC Educational Resources Information Center

    Science Teacher, 1987

    1987-01-01

    Provides reviews of four computer software packages designed for use in science education. Describes courseware dealing with a variety of tips for teaching physics concepts, chemical reactions in an aqueous solution, mitosis and meiosis, and photosynthesis. (TW)

  16. Reviews.

    ERIC Educational Resources Information Center

    Science Teacher, 1988

    1988-01-01

    Reviews four software packages available for IBM PC or Apple II. Includes "Graphical Analysis III"; "Space Max: Space Station Construction Simulation"; "Guesstimation"; and "Genetic Engineering Toolbox." Focuses on each packages' strengths in a high school context. (CW)

  17. Reviews.

    ERIC Educational Resources Information Center

    Radcliffe, George; And Others

    1988-01-01

    Reviews three software packages: 1) a package containing 68 programs covering general topics in chemistry; 2) a package dealing with acid-base titration curves and allows for variables to be changed; 3) a chemistry tutorial and drill package. (MVL)

  18. Generic BWR-4 degraded core in-vessel study. Status report

    SciTech Connect

    Not Available

    1984-11-01

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination.

  19. An analysis of molten-corium-induced failure of drain pipes in BWR Mark 2 containments

    SciTech Connect

    Taleyarkhan, R.P. ); Podowski, M.Z. )

    1991-01-01

    This study has focused on mechanistic simulation and analysis of potential failure modes for inpedestal drywell drain pipes in the Limerick boiling water reactor (BWR) Mark 2 containment. Physical phenomena related to surface tension breakdown, heatup, melting, ablation, crust formation and failure, and core material relocation into drain pipes with simultaneous melting of pipe walls were modeled and analyzed. The results of analysis have been used to assess the possibility of drain pipe failure and the resultant loss of pressure-suppression capability. Estimates have been made for the timing and amount of molten corium released to the wetwell. The study has revealed that significantly different melt progression sequences can result depending upon the failure characteristics of the frozen metallic crust which forms over the drain cover during the initial stages of debris pour. Another important result is that it can take several days for the molten fuel to ablate the frozen metallic debris layer -- if the frozen layer has cooled below 1100 K before fuel attack. 10 refs., 3 figs., 4 tabs.

  20. BWR primary coolant pressure boundary license renewal industry report; revision 1. Final report

    SciTech Connect

    Braden, D.; Stancavage, P.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). This IR provides the technical basis for license renewal for U.S. boiling water reactor (BWR) primary coolant pressure boundaries (PCPB). The report includes requirements on: carbon and stainless steel pipes and fittings; reactor circulation pumps; internal heat exchangers; pressure relief and in line valves; and component supports. These components are in the main steam, recirculation, feedwater, residual heat removal, reactor core isolation cooling, low and high pressure coolant injection, and low and high pressure core spray systems.

  1. Analysis of fission product revaporization in a BWR reactor cooling system during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This report presents a preliminary analysis of fission product revaporization in the Reactor Cooling System (RCS) after the vessel failure. The station blackout transient for BWR Mark I Power Plant is considered. The TRAPMELT3 models of evaporization, chemisorption, and the decay heating of RCS structures and gases are adopted in the analysis. The RCS flow models based on the density-difference between the RCS and containment pedestal region are developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP is developed for the analysis. The REVAP is incorporated with the MARCH, TRAPMELT3 and NAUA codes of the Source Term Code Pack Package (STCP). The NAUA code is used to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors determining the magnitude of revaporization and subsequent release of the volatile fission product. 8 figs., 1 tab.

  2. TRACE/PARCS Core Modeling of a BWR/5 for Accident Analysis of ATWS Events

    SciTech Connect

    Cuadra A.; Baek J.; Cheng, L.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    The TRACE/PARCS computational package [1, 2] isdesigned to be applicable to the analysis of light water reactor operational transients and accidents where the coupling between the neutron kinetics (PARCS) and the thermal-hydraulics and thermal-mechanics (TRACE) is important. TRACE/PARCS has been assessed for itsapplicability to anticipated transients without scram(ATWS) [3]. The challenge, addressed in this study, is to develop a sufficiently rigorous input model that would be acceptable for use in ATWS analysis. Two types of ATWS events were of interest, a turbine trip and a closure of main steam isolation valves (MSIVs). In the first type, initiated by turbine trip, the concern is that the core will become unstable and large power oscillations will occur. In the second type,initiated by MSIV closure,, the concern is the amount of energy being placed into containment and the resulting emergency depressurization. Two separate TRACE/PARCS models of a BWR/5 were developed to analyze these ATWS events at MELLLA+ (maximum extended load line limit plus)operating conditions. One model [4] was used for analysis of ATWS events leading to instability (ATWS-I);the other [5] for ATWS events leading to emergency depressurization (ATWS-ED). Both models included a large portion of the nuclear steam supply system and controls, and a detailed core model, presented henceforth.

  3. Corrosion fatigue behavior of low alloy steels under simulated BWR coolant conditions

    NASA Astrophysics Data System (ADS)

    Huang, J. Y.; Young, M. C.; Jeng, S. L.; Yeh, J. J.; Huang, J. S.; Kuo, R. C.

    2010-10-01

    The corrosion fatigue crack growth behavior of A533 and A508 low alloy steels under simulated boiling water reactor (BWR) coolant conditions was studied. Corrosion fatigue crack growth rates of A533B3 and A508 cl. 3 steels were significantly affected by the steel sulfur content, loading frequency and dissolved oxygen content of water environments. The data points outside the bound of Eason's model could be attributed to the low frequency, higher steel sulfur content and high dissolved oxygen in water environments. The sulfur dissolved in the water environment from the higher-sulfur steels was sufficiently concentrated to acidify the crack tip chemistry even in the hydrogen water chemistry (HWC). Therefore, nitrogenated or HWC water showed little or no beneficiary effect on the high-sulfur steels. For the steel specimens of the same sulfur level, their corrosion fatigue crack growth rates were comparable in different orientations, which could be related to the exposure of fresh sulfides to the water environment. The percentages of sulfides per unit area, by quantitative metallography, were comparable for the steel specimens of both orientations. When the steel sulfur content was decreased to a critical sulfur content 0.005 wt.%, the crack growth rates decreased remarkably.

  4. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  5. Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

    NASA Astrophysics Data System (ADS)

    Huang, J. Y.; Chiang, M. F.; Jeng, S. L.; Huang, J. S.; Kuo, R. C.

    2013-01-01

    The environmentally assisted cracking behavior of dissimilar metal (DM) welds, including Alloy 52-A 508 and Alloy 82-A508, under simulated BWR coolant conditions was studied. Effects of postweld heat treatment and sulfur content of the base metal on the corrosion fatigue and SCC growth rates of DM welds were evaluated. The crack growth rates for the DM weld heat-treated at 621 °C for 24 h were observed to be faster than those for the as-welded. But the DM weld heat-treated at 621 °C for 8 h + 400 °C for 200 h showed better SCC resistance than the as-welded. The longer the heat treatment at 621 °C, the higher the chromium carbides density along the grain boundary was observed. Sulfur could diffuse out of the base metal and segregate along the grain boundaries of the dilution zone, leading to weakening the grain boundary strength and the SCC resistance of the Alloy 52-A508 weld.

  6. BWR spent-fuel measurements with the ION-1/fork detector and a calorimeter

    SciTech Connect

    Rinard, P.M.; Bosler, G.E.

    1986-08-01

    Gamma-ray and neutron measurements were made on about 50 irradiated boiling-water reactor (BWR) fuel assemblies using the Los Alamos National Laboratory ION-1/fork detector. The assemblies were placed in a dry storage cask (DOE's REA-2023) at the General Electric Morris Operation (GE-MO) as part of a program to evaluate the cask performance. Battelle Pacific Northwest Laboratory (PNL) conducted the program. PNL compared axial radiation profiles developed from ION-1/fork measurements with calculated profiles to interpret the temperature distributions within the cask. The gamma-ray profiles correlated with heat-emission rates measured with a calorimeter, which suggests that the ION-1/fork detector is much faster than the more direct calorimeter. In addition, the radiation profiles from the ION-1/fork detector can prevent cask loadings with undesirable heat source distributions. The detector also provides safeguards information by verifying the declared exposures and cooling times. The genuineness of the assemblies is thus confirmed just before the filling and sealing of a cask. The ION-1/fork detector was permanently installed in the GE-MO fuel storage pond for 1 year without any breakdowns or significant maintenance required. Data were gathered for 9 months and analyzed using techniques developed during previous measurement campaigns. A few anomalies were found in generally satisfactory results. The detector's ease of use, reliability, and reproducibility were excellent.

  7. Calculation of MCPR for BWR transients using the BNL plant analyzer

    SciTech Connect

    Horak, W.C.; Diamond, D.J.

    1987-01-01

    A class of transients of interest includes those from full power that involve some changes from the plant's technical specifications. These changes are allowed if it can be demonstrated that the effect on key parameters, such as the minimum critical power ratio (MCPR), is acceptable. Another class of transients is of interest for similar reasons, those that are initiated from the maximum extended operating domain (MEOD) or with partial feedwater heating. In the MEOD, the reactor conditions may be different from previous experience to increase the speed of the power ascension or obtain more power out of the core at end-of-cycle when the reactivity of the fuel is low. The critical power ratio (CPR) is used to determine the thermal limits of boiling water reactors (BWRs). In this study, CPRs for a series of transients run on the Brookhaven Plant Analyzer (BPA) have been calculated. The transients include nominal base case simulations; simulations with variations in relief valve setpoints and the number of failed feedwater heaters; simulations at the 100% power, 75% flow point on the extended load line of the MEOD; and a simulation with partial feedwater heating. The plant represented with the BPA is a BWR/4 rated at 3293 MW with a 6.38-m vessel.

  8. THERMAL EVALUATION OF THE CONCEPTUAL 24 BWR UCF TUBE BASKET DESIGN DISPOSAL CONTAINER

    SciTech Connect

    T.L. Lotz

    1995-12-18

    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: (1) Characterize the conceptual 24 boiling water reactor (BWR) 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 waste package do not preclude UCF waste package 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.

  9. Assessment of the GOTHIC and RELAP5/MOD3 computer codes against BWR-related experiments from the Full Scale Test Facility (FSTF)

    SciTech Connect

    Schor, L.; Yeung, W.S.; Goodwin, E.F.

    1995-12-31

    Many methods, or computer codes, are used for analyzing the thermal-hydraulic response of containments to postulated accident sequences. Among the codes generally available for this purpose are the GOTHIC and RELAP5 computer codes. However, these codes have not been tested well in the area of Boiling Water Reactor (BWR) containment responses. The purpose of this paper is to evaluate the limitations and capabilities of these two codes in analyzing BWR containment response. Test data from the Full Scale Test Facility (FSTF), which simulates accident response of BWR containments, are used. Assessment results of two tests (Tests M2 and M8) are presented. Comparison of the calculated and measured drywell and wetwell response (pressure, temperature) is made and capabilities and limitations of each code in being able to predict the test phenomena are discussed.

  10. Reviews.

    ERIC Educational Resources Information Center

    Science Teacher, 1989

    1989-01-01

    Reviews a software planetarium package called "Sky Travel." Includes two audiovisuals: "Conquest of Space" and "Windows on Science: Earth Science"; and four books: "Small Energy Sources: Choices that Work,""Stonehenge Complete,""Uneasy Careers and Intimate Lives: Women in Science 1789-1979," and "The Rise of Urbanization and the Decline of…

  11. Reviews.

    ERIC Educational Resources Information Center

    Science Teacher, 1988

    1988-01-01

    Presents information and concerns regarding computer courseware, books, and audiovisual materials reviewed by teachers. Covers a variety of topics including dissection of common classroom specimens, medicine, acid rain projects, molecules, the water cycle, erosion, plankton, and evolution. Notes on availability, price, and needed equipment, where…

  12. Reviews.

    ERIC Educational Resources Information Center

    Journal of Chemical Education, 1988

    1988-01-01

    Reviews two computer programs: "Molecular Graphics," which allows molecule manipulation in three-dimensional space (requiring IBM PC with 512K, EGA monitor, and math coprocessor); and "Periodic Law," a database which contains up to 20 items of information on each of the first 103 elements (Apple II or IBM PC). (MVL)

  13. Reviews.

    ERIC Educational Resources Information Center

    Newland, Robert J.; And Others

    1988-01-01

    Reviews four organic chemistry computer programs and three books. Software includes: (1) NMR Simulator 7--for IBM or Macintosh, (2) Nucleic Acid Structure and Synthesis--for IBM, (3) Molecular Design Editor--for Apple II, and (4) Synthetic Adventure--for Apple II and IBM. Book topics include physical chemistry, polymer pioneers, and the basics of…

  14. Reviews.

    ERIC Educational Resources Information Center

    Carpenter, Jeanette; And Others

    1988-01-01

    Reviews two software packages: Graphical Analysis III and Lewis Diagrams. Finds Graphical Analysis III to be a fast and versatile graphing program for high school science classes with access to Apple II microcomputers. Lewis Diagrams is designed to aid in determining Lewis structures of molecules and ions for IBM computers. (MVL)

  15. Reviews.

    ERIC Educational Resources Information Center

    Repak, Arthur J.; And Others

    1988-01-01

    Computer software, audiovisuals, and books are reviewed. Includes topics on interfacing, ionic equilibrium, space, the classification system, Acquired Immune Disease Syndrome, evolution, human body processes, energy, pesticides, teaching school, cells, and geological aspects. Availability, price, and a description of each are provided. (RT)

  16. Reviews.

    ERIC Educational Resources Information Center

    Journal of Chemical Education, 1988

    1988-01-01

    Reviews three computer software packages for chemistry education including "Osmosis and Diffusion" and "E.M.E. Titration Lab" for Apple II and "Simplex-V: An Interactive Computer Program for Experimental Optimization" for IBM PC. Summary ratings include ease of use, content, pedagogic value, student reaction, and cost. (CW)

  17. Reviews.

    ERIC Educational Resources Information Center

    Science Teacher, 1989

    1989-01-01

    Reviews seven software programs: (1) "Science Baseball: Biology" (testing a variety of topics); (2) "Wildways: Understanding Wildlife Conservation"; (3) "Earth Science Computer Test Bank"; (4) "Biology Computer Test Bank"; (5) "Computer Play & Learn Series" (a series of drill and test programs); (6) "ENLIST Micros" (resources on computing for…

  18. Reviews

    NASA Astrophysics Data System (ADS)

    2003-05-01

    DISTANCE-LEARNING COURSES (263) Planetary Science and Astronomy BOOK REVIEWS (263) A New Kind of Science Planetary Science: The Science of Planets Around Stars EQUIPMENT (265) The Science Enhancement Program (SEP) Geiger Counter WEB WATCH (265) Revision sites SOFTWARE (267) Exploration of Physics Volume 1

  19. Instability due to a two recirculation pump trip in a BWR using RAMONA-4B computer code with 3D neutron kinetics

    SciTech Connect

    Cheng, H.S.; Rohatgi, U.S.

    1993-06-01

    An investigation was made of the potential for thermal-hydraulic instabilities coupled to neutronic feedback in a BWR due to a two recirculation pump trip event using the RAMONA-4B computer code with 3D neutron kinetics. It is concluded that a high-power (100%) and low-flow (75%) initial condition would most likely lead to in-phase density wave oscillations after the tripping of both recirculation pumps, and that RAMONA-4B is capable of predicting such thermal-hydraulic instabilities coupled to neutronic feedback in BWR and in SBWR.

  20. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    NASA Astrophysics Data System (ADS)

    Tanaka, Ken-ichi

    2016-06-01

    We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV) of a Boiling Water Reactor (BWR) by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au) and Nickel (Ni) at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  1. A Study on the Conceptual Design of a 1,500 MWe Passive PWR with Annular Fuel

    SciTech Connect

    Kwi Lim Lee; Soon Heung Chang

    2004-07-01

    In this study, the preliminary conceptual design of a 1500 MWe pressurized water reactor (PWR) with annular fuel has been performed. This design is derived from the AP1000 which is a 1000 MWe PWR with two-loop. However, the present design is a 1500 MWe PWR with three-loop, passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. The preliminary design parameters of this reactor have been determined through simple relation to those of AP1000 for reactor, reactor coolant system, and passive safety injection system. Using the MATRA code, we analyze the core designs for two alternatives on fuel assembly types: solid fuel and annular fuel. The performance of reactor cooling systems is evaluated through the accident of the cold leg break in the core makeup tank loop by using MARS2.1 code. This study presents the developmental strategy, preliminary design parameters and safety analysis results. (authors)

  2. Vessel failure time for a low-pressure short-term station blackout in a BWR-4

    SciTech Connect

    Carbajo, J.J. )

    1993-01-01

    A low-pressure, short-term station blackout severe accident sequence has been analyzed using the MELCOR code, version 1.8.1, in a boiling water reactor (BWR)-4. This paper presents a sensitivity study evaluating the effect of several MELCOR input parameters on vessel failure time. Results using the MELCOR/CORBH package and the BWRSAR code are also presented and compared to the MELCOR results. These calculated vessel failure times are discussed, and a judgment is offered as to which is the most realistic.

  3. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    SciTech Connect

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  4. PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

    SciTech Connect

    Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C; Murphy, Brian D; Mueller, Don

    2007-09-01

    The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.

  5. Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

    SciTech Connect

    Fridman, E.; Shwageraus, E.; Galperin, A.

    2006-07-01

    The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of this work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)

  6. Development and validation of advanced CFD models for detailed predictions of void distribution in a BWR bundle

    NASA Astrophysics Data System (ADS)

    Neykov, Boyan

    In recent years, a commonly adopted approach is to use Computational Fluid Dynamics (CFD) codes as computational tools for simulation of different aspects of the nuclear reactor thermal-hydraulic performance where high-resolution and high-fidelity modeling is needed. Within the framework of this PhD work, the CFD code STAR-CD [1] is used for investigations of two phase flow in air-water systems as well as boiling phenomena in simple pipe geometry and in a Boiling Water Reactor (BWR) fuel assembly. Based on the two-fluid Eulerian solver, improvements of the STAR-CD code in the treatment of the drag, lift and wall lubrication forces in a dispersed two phase flow at high vapor (gas) phase fractions are investigated and introduced. These improvements constitute a new two phase modeling framework for STAR-CD, which has been shown to be superior as compared to the default models in STAR-CD. The conservation equations are discretized using the finite-volume method and solved using a solution procedure is based on Pressure Implicit with Splitting of Operators (PISO) algorithm, adapted to the solution of the two-fluid model. The improvements in the drag force modeling include investigation and integration of models with dependence on both void fraction and bubble diameter. The set of the models incorporated into STAR-CD is selected based on an extensive literature review focused on two phase systems with high vapor fractions. The research related to the modeling of wall lubrication force is focused on the validation of the already existing model in STAR-CD. The major contribution of this research is the development and implementation of an improved correlation for the lift coefficient used in the lift force formula. While a variety of correlations for the lift coefficient can be found in the open literature, most of those were derived from experiments conducted at low vapor (gas) phase fractions and are not applicable to the flow conditions existing in the BWRs. Therefore

  7. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  8. Application of LBB to high energy piping systems in operating PWR

    SciTech Connect

    Swamy, S.A.; Bhowmick, D.C.

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  9. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    SciTech Connect

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-10-03

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation.

  10. Overview of New Tools to Perform Safety Analysis: BWR Station Black Out Test Case

    SciTech Connect

    D. Mandelli; C. Smith; T. Riley; J. Nielsen; J. Schroeder; C. Rabiti; A. Alfonsi; Cogliati; R. Kinoshita; V. Pasucci; B. Wang; D. Maljovec

    2014-06-01

    Dynamic Probabilistic Risk Assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). While system simulator codes accurately model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by: 1) sampling values of a set of parameters from the uncertainty space of interest (using the simulation controller codes), and 2) simulating the system behavior for that specific set of parameter values (using the system simulator codes). For complex systems, one of the major challenges in using DPRA methodologies is to analyze the large amount of information (i.e., large number of scenarios ) generated, where clustering techniques are typically employed to allow users to better organize and interpret the data. In this paper, we focus on the analysis of a nuclear simulation dataset that is part of the Risk Informed Safety Margin Characterization (RISMC) Boiling Water Reactor (BWR) station blackout (SBO) case study. We apply a software tool that provides the domain experts with an interactive analysis and visualization environment for understanding the structures of such high-dimensional nuclear simulation datasets. Our tool encodes traditional and topology-based clustering techniques, where the latter partitions the data points into clusters based on their uniform gradient flow behavior. We demonstrate through our case study that both types of clustering techniques complement each other in bringing enhanced structural understanding of the data.

  11. Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

    SciTech Connect

    Gorzel, A.

    2006-07-01

    Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)

  12. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  13. Reviews

    NASA Astrophysics Data System (ADS)

    2005-07-01

    WE RECOMMEND When Physics Became King This book delves into the history of science since the 18th century. The History of the Laser An interesting read that will teach you far more than its title suggests. History of Physics Selected Reprints A fascinating collection of physics papers spanning four decades. Datalogging set-ups Five great products from Leybold Didactic’s CASSY range. Videocom Measure motion and convert it to graphs with this great device. Basic Raybox This simple piece of equipment offers great performance. WORTH A LOOK Virtual Physics Lab John Nunn’s software demystifies science using clear illustrations. HANDLE WITH CARE Microchem Electricity Kit This box of equipment for introducing electricity lacks quality. Raymond the Raybox A disappointing raybox. The basic version reviewed on p389 is better. WEB WATCH A rough guide to e-learning.

  14. Technical Specification action statements requiring shutdown. A risk perspective with application to the RHR/SSW systems of a BWR

    SciTech Connect

    Mankamo, T.; Kim, I.S.; Samanta, P.K.

    1993-11-01

    When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements.

  15. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    SciTech Connect

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. ); Wagner, K.C. )

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  16. Phenomenon analysis of stress corrosion cracking in the vessel head penetrations of French PWR`s

    SciTech Connect

    Pichon, C.; Buisine, D.; Faidy, C.; Gelpi, A.; Vaindirlis, M.

    1995-12-31

    During a hydrotest in 1991, a leak was detected on,a reactor vessel head (RVH) penetration of a French PWR. This leak was due to a phenomenon of Primary Water Stress Corrosion Cracking (PWSCC) affecting these penetrations in Alloy 600. The destructive and non-destructive examinations undertaken during the following months highlighted the generic nature of the degradations. In order to well understand this phenomenon and implement the most suitable maintenance policy, a large scale scientific program was decided and performed jointly by Electricite de France and FRAMATOME. The paper will present all the results obtained in this program concerning the parameters governing the PWSCC. In particular the following fields will be developed: (1) the material, its microstructure in line with the manufacturing and its susceptibility to PWSCC; (2) the stresses and their evaluations by measurements, mock up corrosion tests and Finite Element Analysis (FEA); (3) the effect of surface finish on crack initiation; and (4) the crack growth rate. This phenomenon analysis will be useful for evaluating the risk of PWSCC on other Alloy 600 areas in PWR`s primary system.

  17. Regeneratively Cooled Liquid Oxygen/Methane Technology Development Between NASA MSFC and PWR

    NASA Technical Reports Server (NTRS)

    Robinson, Joel W.; Greene, Christopher B.; Stout, Jeffrey B.

    2012-01-01

    The National Aeronautics & Space Administration (NASA) has identified Liquid Oxygen (LOX)/Liquid Methane (LCH4) as a potential propellant combination for future space vehicles based upon exploration studies. The technology is estimated to have higher performance and lower overall systems mass compared to existing hypergolic propulsion systems. NASA-Marshall Space Flight Center (MSFC) in concert with industry partner Pratt & Whitney Rocketdyne (PWR) utilized a Space Act Agreement to test an oxygen/methane engine system in the Summer of 2010. PWR provided a 5,500 lbf (24,465 N) LOX/LCH4 regenerative cycle engine to demonstrate advanced thrust chamber assembly hardware and to evaluate the performance characteristics of the system. The chamber designs offered alternatives to traditional regenerative engine designs with improvements in cost and/or performance. MSFC provided the test stand, consumables and test personnel. The hot fire testing explored the effective cooling of one of the thrust chamber designs along with determining the combustion efficiency with variations of pressure and mixture ratio. The paper will summarize the status of these efforts.

  18. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    SciTech Connect

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    2013-07-01

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  19. Application of the MELCOR code to design basis PWR large dry containment analysis.

    SciTech Connect

    Phillips, Jesse; Notafrancesco, Allen; Tills, Jack Lee

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  20. Development of a new lattice physics code robin for PWR application

    SciTech Connect

    Zhang, S.; Chen, G.

    2013-07-01

    This paper presents a description of methodologies and preliminary verification results of a new lattice physics code ROBIN, being developed for PWR application at Shanghai NuStar Nuclear Power Technology Co., Ltd. The methods used in ROBIN to fulfill various tasks of lattice physics analysis are an integration of historical methods and new methods that came into being very recently. Not only these methods like equivalence theory for resonance treatment and method of characteristics for neutron transport calculation are adopted, as they are applied in many of today's production-level LWR lattice codes, but also very useful new methods like the enhanced neutron current method for Dancoff correction in large and complicated geometry and the log linear rate constant power depletion method for Gd-bearing fuel are implemented in the code. A small sample of verification results are provided to illustrate the type of accuracy achievable using ROBIN. It is demonstrated that ROBIN is capable of satisfying most of the needs for PWR lattice analysis and has the potential to become a production quality code in the future. (authors)

  1. Three Dimensional Analysis of 3-Loop PWR RCCA Ejection Accident for High Burnup

    SciTech Connect

    Marciulescu, Cristian; Sung, Yixing; Beard, Charles L.

    2006-07-01

    The Rod Control Cluster Assembly (RCCA) ejection accident is a Condition IV design basis reactivity insertion event for Pressurized Water Reactors (PWR). The event is historically analyzed using a one-dimensional (1D) neutron kinetic code to meet the current licensing criteria for fuel rod burnup to 62,000 MWD/MTU. The Westinghouse USNRC-approved three-dimensional (3D) analysis methodology is based on the neutron kinetics version of the ANC code (SPNOVA) coupled with Westinghouse's version of the EPRI core thermal-hydraulic code VIPRE-01. The 3D methodology provides a more realistic yet conservative analysis approach to meet anticipated reduction in the licensing fuel enthalpy rise limit for high burnup fuel. A rod ejection analysis using the 3D methodology was recently performed for a Westinghouse 3-loop PWR at an up-rated core power of 3151 MWt with reload cores that allow large flexibility in assembly shuffling and a fuel hot rod burnup to 75,000 MWD/MTU. The analysis considered high enrichment fuel assemblies at the control rod locations as well as bounding rodded depletions in the end of life, zero power and full power conditions. The analysis results demonstrated that the peak fuel enthalpy rise is less than 100 cal/g for the transient initiated at the hot zero power condition. The maximum fuel enthalpy is less than 200 cal/g for the transient initiated from the full power condition. (authors)

  2. RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

    SciTech Connect

    David Andrs; Ray Berry; Derek Gaston; Richard Martineau; John Peterson; Hongbin Zhang; Haihua Zhao; Ling Zou

    2012-05-01

    The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to

  3. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect

    T. Schmitt

    2005-08-17

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  4. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect

    V. DeLa Brosse

    2003-03-27

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  5. Analysis of BWR OPRM plant data and detection algorithms with DSSPP

    SciTech Connect

    Yang, J.; Vedovi, J.; Chung, A. K.; Zino, J. F.

    2012-07-01

    All U.S. BWRs are required to have licensed stability solutions that satisfy General Design Criteria (GDC) 10 and 12 of 10 CFR 50 Appendix A. Implemented solutions are either detect and suppress or preventive in nature. Detection and suppression of power oscillations is accomplished by specialized hardware and software such as the Oscillation Power Range Monitor (OPRM) utilized in Option III and Detect and Suppress Solution - Confirmation Density (DSS-CD) stability Long-Term Solutions (LTSs). The detection algorithms are designed to recognize a Thermal-Hydraulic Instability (THI) event and initiate control rod insertion before the power oscillations increase much higher above the noise level that may threaten the fuel integrity. Option III is the most widely used long-term stability solution in the US and has more than 200 reactor years of operational history. DSS-CD represents an evolutionary step from the stability LTS Option III and its licensed domain envelopes the Maximum Extended Load Line Limit Analysis Plus (MELLLA +) domain. In order to enhance the capability to investigate the sensitivity of key parameters of stability detection algorithms, GEH has developed a new engineering analysis code, namely DSSPP (Detect and Suppress Solution Post Processor), which is introduced in this paper. The DSSPP analysis tool represents a major advancement in the method for diagnosing the design of stability detection algorithms that enables designers to perform parametric studies of the key parameters relevant for THI events and to fine tune these system parameters such that a potential spurious scram might be avoided. Demonstrations of DSSPPs application are also presented in this paper utilizing actual plant THI data. A BWR/6 plant had a plant transient that included unplanned recirculation pump transfer from fast to slow speed resulting in about 100% to {approx}40% rated power decrease and about 99% to {approx}30% rated core flow decrease. As the feedwater temperature

  6. STEAM LINE BREAK AND STATION BLACKOUT TRANSIENTS FOR PROLIFERATION RESISTANT HEXAGONAL TIGHT LATTICE BWR.

    SciTech Connect

    ROHATGI,U.S.; JO,J.; CHUNG,B.D.; TAKAHASHI,H.

    2002-06-09

    Safety analyses of a proliferation resistant, economically competitive, high conversion, boiling water reactor (HCBWR) fueled with fissile plutonium and fertile thorium oxide fuel elements, and with passive safety systems are presented here. The HCBWR developed here is characterized by a very tight lattice with a relatively small water volume fraction in the core which therefore operates with a fast reactor neutron spectrum, and a considerably improved neutron economy compared to the current generation of Light Water Reactors. The tight lattice core has a very narrow flow channels with a hydraulic diameter less than half of the regular BWR core and, thus, presents a special challenge to core cooling, because of reduced water inventory and high friction in the core. The primary safety concern when reducing the moderator to fuel ratio and when using a tightly packed lattice arrangement is to maintain adequate cooling of the core during both normal operation and accident scenarios. In the preliminary HCBWR design, the core has been placed in a vessel with a large chimney section, and the vessel is connected with Isolation Condenser System (ICs). The vessel is placed in containment with Gravity Driven Cooling System (GDCS) and Passive Containment Cooling System (PCCS) in a configuration similar to General Electric's Simplified Boiling Water Reactor (SBWR). The safety systems are similar to SBWR; ICs and PCCS are scaled with power. An internal recirculation pump was placed in the downcomer to augment the buoyancy head provided by the chimney, since the buoyancy provided by the chimney alone could not generate sufficient recirculation in the vessel as the tight lattice configuration resulted in much larger friction in the core than the SBWR. The constitutive relationships for RELAP5 were assessed for narrow channels, and as a result the heat transfer package was modified. The modified RELAP5 was used to simulate and analyze two of the most limiting events for a tight

  7. Qualification of helium measurement system for detection of fuel failures in a BWR

    NASA Astrophysics Data System (ADS)

    Larsson, I.; Sihver, L.; Loner, H.; Grundin, A.; Helmersson, J.-O.; Ledergerber, G.

    2014-05-01

    There are several methods for surveillance of fuel integrity during the operation of a boiling water reactor (BWR). The detection of fuel failures is usually performed by analysis of grab samples of off-gas and coolant activities, where a measured increased level of ionizing radiation serves as an indication of new failure or degradation of an already existing one. At some nuclear power plants the detection of fuel failures is performed by on-line nuclide specific measurements of the released fission gases in the off-gas system. However, it can be difficult to distinguish primary fuel failures from degradation of already existing failures. In this paper, a helium measuring system installed in connection to a nuclide specific measuring system to support detection of fuel failures and separate primary fuel failures from secondary ones is presented. Helium measurements provide valuable additional information to measurements of the gamma emitting fission gases for detection of primary fuel failures, since helium is used as a fill gas in the fuel rods during fabrication. The ability to detect fuel failures using helium measurements was studied by injection of helium into the feed water systems at the Forsmark nuclear power plant (NPP) in Sweden and at the nuclear power plant Leibstadt (KKL) in Switzerland. In addition, the influence of an off-gas delay line on the helium measurements was examined at KKL by injecting helium into the off-gas system. By using different injection rates, several types of fuel failures with different helium release rates were simulated. From these measurements, it was confirmed that the helium released by a failed fuel can be detected. It was also shown that the helium measurements for the detection of fuel failures should be performed at a sampling point located before any delay system. Hence, these studies showed that helium measurements can be useful to support detection of fuel failures. However, not all fuel failures which occurred at

  8. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    SciTech Connect

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1985-04-01

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

  9. Methods and findings of a systems interaction study of a Westinghouse PWR

    SciTech Connect

    Youngblood, R.; Hanan, N.; Fitzpatrick, R.; Xue, D.; Bozoki, G.; Fresco, A.; Papazoglou, I.; Mitra, S.; Macdonald, G.; Chelliah, E.

    1985-01-01

    This paper describes the methods and findings of a systems interaction study of a Westinghouse PWR. BNL conducted the study as a methods application that was performed to support the resolution of Unresolved Safety Issue A-17 on Systems Interactions. The method calls for a fault tree model of the plant to be developed in stages, corresponding to successively increasing levels of scope and detail. A functional model is developed first, resolved only to sufficient detail to reflect support system dependences; this guides the subsequent searches for spatial and induced-human interactions. This process has led to the identification of an active single failure causing loss of low pressure injection following a large or medium LOCA.

  10. Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

    SciTech Connect

    Simion, G.P.; VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B.; Bulmahn, K.D.

    1993-06-01

    This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement.

  11. Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

    NASA Astrophysics Data System (ADS)

    Burdo, James S.

    This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the

  12. Three Dimensional Radiation Transport Analyses in Pwr with Tort and Mcnp

    NASA Astrophysics Data System (ADS)

    Fukuya, Koji; Nakata, Hayato; Kimura, Itsuro; Kitagawa, Hideo; Ohmura, Masaki; Ito, Taku; Shin, Kazuo

    2003-06-01

    Three dimensional (3D) neutron and gamma calculations for structural materials inside the reactor vessel in a commercial PWR were performed using the 3D transport code TORT and the Monte Carlo code MCNP to assess the accuracy of calculations using these codes and libraries. Comparisons with two dimensional DORT calculations with various libraries and surveillance dosimetry measurements indicated that TORT and MCNP calculations give similar agreements with surveillance measurements to DORT calculations. Influences of the cross section data, ENDF/B-IV, ENDF/B-VI and JENDL3.2 on attenuation of the fast flux and dpa rate in the reactor vessel, relative contributions of gamma-rays and thermal neutrons to dpa were discussed.

  13. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    SciTech Connect

    Herer, C.

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  14. Grid-to-rod flow-induced impact study for PWR fuel in reactor

    DOE PAGESBeta

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; Wang, Jy-An John

    2016-06-10

    The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

  15. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  16. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR

    SciTech Connect

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed.

  17. VISA: a computer code for predicting the probability of reactor pressure-vessel failure. [PWR

    SciTech Connect

    Stevens, D.L.; Simonen, F.A.; Strosnider, J. Jr.; Klecker, R.W.; Engel, D.W.; Johnson, K.I.

    1983-09-01

    The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NRC staff evaluation of pressurized thermal shock. VISA uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics are used to model crack initiation and propagation. parameters for initial crack size, copper content, initial RT/sub NDT/, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents the version of VISA used in the NRC staff report (Policy Issue from J.W. Dircks to NRC Commissioners, Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982, SECY-82-465) and includes a user's guide for the code.

  18. Failure probability of PWR reactor coolant loop piping. [Double-ended guillotine break

    SciTech Connect

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria.

  19. Fog inerting effects on hydrogen combustion in a PWR ice condenser contaminant

    SciTech Connect

    Luangdilok, W.; Bennett, R.B.

    1995-05-01

    A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the upward lean flammability limits of the H{sub 2}-air-steam mixture in the ice condenser upper plenum region of a pressurized water reactor (PWR) ice condenser contaminant during postulated large loss of coolant accident (LOCA) conditions indicate that combustion may be suppressed beyond the downward flammability limit (8 percent H{sub 2} by volume). 18 refs., 3 tabs.

  20. Feasibility of recycling thorium in a fusion-fission hybrid/PWR symbiotic system

    SciTech Connect

    Josephs, J. M.

    1980-12-31

    A study was made of the economic impact of high levels of radioactivity in the thorium fuel cycle. The sources of this radioactivity and means of calculating the radioactive levels at various stages in the fuel cycle are discussed and estimates of expected levels are given. The feasibility of various methods of recycling thorium is discussed. These methods include direct recycle, recycle after storage for 14 years to allow radioactivity to decrease, shortening irradiation times to limit radioactivity build up, and the use of the window in time immediately after reprocessing where radioactivity levels are diminished. An economic comparison is made for the first two methods together with the throwaway option where thorium is not recycled using a mass energy flow model developed for a CTHR (Commercial Tokamak Hybrid Reactor), a fusion-fission hybrid reactor which serves as fuel producer for several PWR reactors.

  1. Analysis of MERCI decay heat measurement for PWR UO{sub 2} fuel rod

    SciTech Connect

    Jaboulay, J.C.; Bourganel, S.

    2012-01-15

    Decay heat measurements, called the MERCI experiment, were conducted at Commissariat a l'Energie Atomique (CEA)/Saclay to characterize accurately residual power at short cooling time and verify its prediction by decay code and nuclear data. The MOSAIC calorimeter, developed and patented by CEA/Grenoble (DTN/SE2T), enables measurement of the decay heat released by a pressurized water reactor (PWR) fuel rod sample between 200 and 4 W within a precision of 1%. The MERCI experiment included three phases. At first, a UO{sub 2} fuel rod sample was irradiated in the CEA/Saclay experimental reactor OSIRIS. The burnup achieved at the end of irradiation was similar to 3.5 GWd/tonne. The second phase was the transfer of the fuel rod sample from its irradiation location to a hot cell, to be inserted inside the MOSAIC calorimeter. It took 26 min to carry out the transfer. Finally, decay heat released by the PWR sample was measured from 27 min to 42 days after shutdown. Post irradiation examinations were performed to measure concentrations of some heavy nuclei (U, Pu) and fission products (Cs, Nd). The decay heat was predicted using a calculation scheme based on the PEPIN2 depletion code, the TRIPOLI-4 Monte Carlo code, and the JEFF3.1.1 nuclear data file. The MERCI experiment analysis shows that the discrepancy between the calculated and the experimental decay heat values is included between -10% at 27 min and +6% at 12 h, 30 min otter shutdown. From 4 up to 42 days of cooling time, the difference between calculation and measurement is about ± 1%, i.e., experimental uncertainty. The MERCI experiment represents a significant contribution for code validation; the time range above 10{sup 5} s has not been validated previously. (authors)

  2. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    SciTech Connect

    DeHart, M.D.

    1993-01-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also 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 spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, 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. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

  3. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  4. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    SciTech Connect

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  5. Uncertainties in source term estimates for a station blackout accident in a BWR with Mark I containment

    SciTech Connect

    Lee, M.; Cazzoli, E.; Liu, Y.; Davis, R.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.; Khatib-Rahbar, M.

    1988-01-01

    In this paper, attention is limited to a single accident progression sequence, namly a station blackout accident in a BWR with a Mark I containment building. Identified as an important accident in the draft version of NUREG-1150 a station blackout involves loss of both off-site power and dc power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation cooling systems. This paper illustrates the calculated uncertainties (Probability Density Functions) associated with the radiological releases into the environment for the nine fission product groups at 10 hours following the initiation of core-concrete interactions. Also shown are the results ofthe STCP base case simulation. 5 refs., 1 fig., 1 tab.

  6. Analysis of high pressure boil-off situation during MSIV closure ATWS in a typical BWR/4

    SciTech Connect

    Neymotin, L.Y.; Slovik, G.C.; Saha, P.

    1986-01-01

    The objective of this paper is to provide a best-estimate analysis of the MSIV Closure ATWS in the Browns Ferry Unit 1 BWR with Mark 1 containment. The calculations have been performed using the RAMONA-3B code which has a three-dimensional neutron kinetics model coupled with one-dimensional (multi-channel core representation), four-equation, nonhomogeneous, nonequilibrium thermal hydraulics. The code also allows for one-dimensional neutronic core representation. The 1-D capability of the code has been employed in this calculation since a thorough sensitivity study showed that for a full ATWS, a one-dimensional (axial) neutron kinetics adequately describes the core behavior. (Note that the core steady-state symmetry in this case was preserved throughout the transient so that radial effects could be neglected.) The calculation described in the paper was started from a steady-state fuel condition corresponding to the end of Cycle 5 of the Browns Ferry reactor.

  7. Neutronics Design and Fuel Cycle Analysis of a High Conversion BWR with Pu-Th Fuel

    SciTech Connect

    Xu, Yunlin; Downar, T.J.; Takahashi, H.; Rohatgi, U.S.

    2002-07-01

    As part of the U.S. Department of Energy's (DOE) Nuclear Energy Research Initiative (NERI), a 'Generation IV' high conversion Boiling Water Reactor design is being investigated at Purdue University and Brookhaven National Laboratory. One of the primary innovative design features of the core proposed here is the use of Thorium as fertile material. In addition to the advantageous nonproliferation and waste characteristics of thorium fuel cycles, the use of thorium is particularly important in a tight pitch, high conversion lattice in order to insure a negative void coefficient throughout the operating life of the reactor. The principal design objective of a high conversion light water reactor is to substantially increase the conversion ratio (fissile atoms produced per fissile atoms consumed) of the reactor without compromising the safety performance of the plant. Since existing LWRs have a relatively low conversion ratio they require relatively frequent refueling which limits the economic efficiency of the plant. Also, the high volume of spent fuel can pose a burden for waste storage and the accumulation of plutonium in the uranium fuel cycle can become a materials proliferation issue. The development of Fast Breeder Reactors (FBR) as an alternative technology to alleviate some of these concerns has been delayed for various reasons. An intermediate solution has been to examine tight pitch light water reactors which can provide significant improvements in the fuel cycle performance of the existing LWRs by taking advantage of the increased conversion ratios from the harder neutron spectrum in the tight pitch lattice, as well as the by taking advantage of the waste and nonproliferation benefits of the thorium fuel cycle. Several High Conversion BWR designs have been proposed by researchers in Japan and elsewhere during the past several years. One of the more promising HCR designs is the Reduced Moderation Water Reactor (RMWR) proposed by JAERI [1]. Their design was

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

    SciTech Connect

    J.W. Davis

    1996-08-29

    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.

  9. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    NASA Astrophysics Data System (ADS)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  10. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    SciTech Connect

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  11. Possible Methods to Estimate Core Location in a Beyond-Design-Basis Accident at a GE BWR with a Mark I Containment Stucture

    SciTech Connect

    Walston, S; Rowland, M; Campbell, K

    2011-07-27

    It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting in a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.

  12. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis, Model description. Volume 1

    SciTech Connect

    Borkowski, J.A.; Wade, N.L.; Giles, M.M.; Rouhani, S.Z.; Shumway, R.W.; Singer, G.L.; Taylor, D.D.; Weaver, W.L.

    1992-08-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODl version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MODI version produces results consistent with previous versions. Assessment calculations using the two TRAC-BF1 versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  13. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis: User`s guide. Volume 2

    SciTech Connect

    Rettig, W.H.; Wade, N.L.

    1992-06-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODI version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MOD1 version produces results consistent with previous versions. Assessment calculations using the two TRAC-BFI versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  14. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    SciTech Connect

    Peterson, T.A.

    1997-06-01

    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  15. PwrSoC (integration of micro-magnetic inductors/transformers with active semiconductors) for more than Moore technologies

    NASA Astrophysics Data System (ADS)

    Mathuna, Cian Ó.; Wang, Ningning; Kulkarni, Santosh; Roy, Saibal

    2013-07-01

    This paper introduces the concept of power supply on chip (PwrSoC) which will enable the development of next-generation, functionally integrated, power management platforms with applications in dc-dc conversion, gate drives, isolated power transmission and ultimately, high granularity, on-chip, power management for mixed-signal, SOC chips. PwrSoC will integrate power passives with the power management IC, in a 3D stacked or monolithic form factor, thereby delivering the performance of a highefficiency dc-dc converter within the footprint of a low-efficiency linear regulator. A central element of the PwrSoC concept is the fabrication of power micro-magnetics on silicon to deliver micro-inductors and micro-transformers. The paper details the magnetics on silicon process which combines thin film magnetic core technology with electroplated copper conductors. Measured data for micro-inductors show inductance operation up to 20 MHz, footprints down to 0.5 mm2, efficiencies up to 93% and dc current carrying capability up to 600 mA. Measurements on micro-transformers show voltage gain of approximately - 1 dB at between 10 MHz and 30 MHz. Contribution to the Topical Issue “International Semiconductor Conference Dresden-Grenoble - ISCDG 2012”, Edited by Gérard Ghibaudo, Francis Balestra and Simon Deleonibus.

  16. Evaluation of stress corrosion cracking of irradiated 304L stainless steel in PWR environment using heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

    2016-08-01

    IASCC has been a major concern regarding the structural and functional integrity of core internals of PWR's, especially baffle-to-former bolts. Despite numerous studies over the past few decades, additional evaluation of the parameters influencing IASCC is still needed for an accurate understanding and modeling of this phenomenon. In this study, Fe irradiation at 450 °C was used to study the cracking susceptibility of 304 L austenitic stainless steel. After 10 MeV Fe irradiation to 5 dpa, irradiation-induced damage in the microstructure was characterized and quantified along with nano-hardness measurements. After 4% plastic strain in a PWR environment, quantitative information on the degree of strain localization, as determined by slip-line spacing, was obtained using SEM. Fe-irradiated material strained to 4% in a PWR environment exhibited crack initiation sites that were similar to those that occur in neutron- and proton-irradiated materials, which suggests that Fe irradiation may be a representative means for studying IASCC susceptibility. Fe-irradiated material subjected to 4% plastic strain in an inert argon environment did not exhibit any cracking, which suggests that localized deformation is not in itself sufficient for initiating cracking for the irradiation conditions used in this study.

  17. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    SciTech Connect

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

  18. Validation of the scale system for PWR spent fuel isotopic composition analyses

    SciTech Connect

    Hermann, O.W.; Bowman, S.M.; Parks, C.V.; Brady, M.C.

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  19. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    SciTech Connect

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-07-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  20. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    SciTech Connect

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E.

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  1. Feasibility of recycling thorium in a fusion-fission hybrid/PWR symbiotic system

    SciTech Connect

    Josephs, J.M.

    1980-12-31

    A study was made of the economic impact of high levels of radioactivity in the thorium fuel cycle. The sources of this radioactivity and means of calculating the radioactive levels at various stages in the fuel cycle are discussed and estimates of expected levels are given. The feasibility of various methods of recycling thorium is discussed. These methods include direct recycle, recycle after storage for 14 years to allow radioactivity to decrease, shortening irradiation times to limit radioactivity build up, and the use of the window in time immediately after reprocessing where radioactivity levels are diminished. An economic comparison is made for the first two methods together with the throwaway option where thorium is not recycled using a mass energy flow model developed for a CTHR (Commercial Tokamak Hybrid Reactor), a fusion fission hybrid reactor which serves as fuel producer for several PWR reactors. The storage option is found to be most favorable; however, even this option represents a significant economic impact due to radioactivity of 0.074 mills/kW-h which amounts to $4 x 10/sup 9/ over a 30 year period assuming a 200 gigawatt supply of electrical power.

  2. Integrated Radiation Transport and Thermo-Mechanics Simulation of a PWR Assembly

    SciTech Connect

    Clarno, Kevin T; Hamilton, Steven P; Philip, Bobby; Sampath, Rahul S; Allu, Srikanth; Berrill, Mark A; Barai, Pallab; Banfield, James E

    2012-01-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step towards incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source terms, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses, of traditional (single-pin) nuclear fuel performance simulation. AMPFuel was used to model an entire 17 x 17 Pressurized Water Reactor (PWR) fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins, the 25 guide tubes, top and bottom structural regions, and the upper and lower (neutron) reflector regions. The final full-assembly calculation was executed on Jaguar (Cray XT5) at the Oak Ridge Leadership Computing Facility using 40,000 cores in under 10 hours to model over 162 billion degrees of freedom for 10 loading steps.

  3. Modern Fuel Cladding in Demanding Operation - ZIRLO in Full Life High Lithium PWR Coolant

    SciTech Connect

    Kargol, Kenneth; Stevens, Jim; Bosma, John; Iyer, Jayashri; Wikmark, Gunnar

    2007-07-01

    There is an increasing demand to optimize the PWR water chemistry in order to minimize activity build-up in the plants and to avoid CIPS and other fuel related issues. Operation with a constant pH between 7.2 and 7.4 is generally considered an important part in achieving the optimized water chemistry. The extended long cycles currently used in most of the U.S. PWRs implies that the lithium concentration at BOC will be outside the general operating experience with such a coolant chemistry regime. With the purpose to extend the experience of high lithium coolant operation, such water chemistry has been used in a few PWRs, i.e. CPSES Unit 2 and Diablo Canyon Units 1 and 2, all with ZIRLO{sup TM} cladding. Operation with a lithium concentration up to 4.2 ppm does not show any impact of the elevated lithium, while operation with up to 6 ppm possibly produce some limited corrosion acceleration in the region of sub-nucleate boiling but has no detrimental impact under the conditions limited by current operating experience. (authors)

  4. Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

    SciTech Connect

    Feltus, M.A.; Pozsgai, C. )

    1992-01-01

    Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative nature of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.

  5. Brief account of the effect of overcooling accidents on the integrity of PWR pressure vessels

    SciTech Connect

    Cheverton, R.D.

    1982-01-01

    The occurrence in recent years of several (PWR) accident initiating events that could lead to severe thermal shock to the reactor pressure vessel, and the growing awareness that copper and nickel in the vessel material significantly enhance radiation damage in the vessel, have resulted in a reevaluation of pressure-vessel integrity during postulated overcooling accidents. Analyses indicate that the accidents of concern are those involving both thermal shock and pressure loadings, and that an accident similar to that at Rancho Seco in 1978 could, under some circumstances and at a time late in the normal life of the vessel, result in propagation of preexistent flaws in the vessel wall to the extent that they might completely penetrate the wall. More severe accidents have been postulated that would result in even shorter permissible lifetimes. However, the state-of-the-art fracture-mechanics analysis may contain excessive conservatism, and this possibility is being investigated. Furthermore, there are several remedial measures, such as fuel shuffling, to reduce the damage rate, and vessel annealing, to restore favorable material properties, that may be practical and used if necessary. 5 figures.

  6. Modeling and design of a reload PWR core for a 48-month fuel cycle

    SciTech Connect

    McMahon, M.V.; Driscoll, M.J.; Todreas, N.E.

    1997-05-01

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

  7. Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.

    1994-11-15

    Version 00 The MARIA System calculates cross sections for PWR fuel assembly calculations. It generates the cross sections library for the diffusion calculations with burnup and feedback effects (CARMEN System, NEA 0649 and RSIC CCC-487) and the k(infinite) and M**2 parameters for the nodal calculations (SIMULA, NEA 0768). MARIA includes three modules. PRELIM generates the input data for the fuel assembly calculation module, for all fuel assembly types in the core and at any conditionmore » of power rate and temperature. WIMS-TRACA is a modified version of the fuel assembly calculation program WIMS-D/4 (NEA 0329 and RSIC CCC-576), which generates the collapsed cross sections versus burn up needed by the CARMEN code (reference cell, boron, xenon, samarium, and light water). POSWIM calculates the transport corrections to the diffusion constant of the absorber materials generated by WIMS-TRACA, to be used directly in the diffusion code when rods or burnable absorber rods are present.« less

  8. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  9. Whole-core comet solutions to a 3-dimensional PWR benchmark problem with gadolinium

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2012-07-01

    A pressurized water reactor (PWR) benchmark problem with gadolinium was used to determine the accuracy and computational efficiency of the coarse mesh radiation transport method COMET. The benchmark problem contains 193 square fuel assemblies. The COMET solution (eigenvalue, assembly averaged and fuel pin averaged fission density distributions) was compared with those obtained from the corresponding Monte Carlo reference solution using the same 2-group material cross section library. The comparison showed that both the core eigenvalue and fission density distribution averaged over each assembly and fuel pin predicated by COMET agree very well with the corresponding MCNP reference solution if the incident flux response expansion used in COMET is truncated at 2nd order in the two spatial and the two angular variables. The benchmark calculations indicate that COMET has Monte Carlo accuracy. In, particular, the eigenvalue difference between the codes ranged from 17 pcm to 35 pcm, being within 2 standard deviations of the calculational uncertainty. The mean flux weighted relative differences in the assembly and fuel pin fission densities were 0.47% and 0.65%, respectively. It was also found that COMET's full (whole) core computational speed is 30,000 times faster than MCNP in which only 1/8 of the core is modeled. It is estimated that COMET would have been about over 6 orders of magnitude faster than MCNP if the full core were also modeled in MCNP. (authors)

  10. UO 2/Zry-4 chemical interaction layers for intact and leak PWR fuel rods

    NASA Astrophysics Data System (ADS)

    Kim, Kyu-Tae

    2010-09-01

    In this study, the UO 2 pellet-Zry-4 cladding interfaces of intact and leak PWR fuel rods were examined with the help of an optical microscope and a scanning electron microscope to investigate typical chemical interaction layers formed at the pellet-cladding interface during the normal reactor operations. The two intact and two leak fuel rods with the burnup of between 35,000 and 53,000 MWD/MTU were selected to evaluate the effects of gap-gas compositions and fuel burnup on the chemical interaction layer formation. Based on the optical and scanning electron micrographs, it is found that the intact fuel rod generates apparently one interaction layer of (U,Zr)O 2-x at the interface, whereas the leak fuel rod generates apparently two interaction layers of ZrO 2-x and (U,Zr)O 2-x. These interaction layers for the intact and leak fuel rods were predicted by several diffusion paths drawn on a U-Zr-O ternary phase diagram. The variations of chemical element compositions around the interface of one intact rod were generated by an electron probe micro-analyzer to confirm the interaction layers at the pellet-cladding interface. The interaction layer growth rates of the ZrO 2-x and (U,Zr)O 2-x phases were estimated, using the layer thicknesses and the reaction times.

  11. Aging mechanisms in the Westinghouse PWR (Pressurized Water Reactor) Control Rod Drive system

    SciTech Connect

    Gunther, W.; Sullivan, K.

    1991-01-01

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs.

  12. ABWR start-up test analysis using BWR core simulator with three-dimensional direct response matrix method

    SciTech Connect

    Mitsuyasu, T.; Ishii, K.; Hino, T.; Aoyama, M.

    2012-07-01

    The ABWR start-up test analysis has been done with the BWR core simulator using the three--dimensional direct response matrix (3D-DRM) method. The Monte Carlo code VMONT made the sub-response matrices for the 3D-DRM method. Each boundary surface was subdivided by 4 x 4 for transverse segments, by 4 for angular segments and by 4 for axial zones in a node. For the calculation speedup, the 3D-DRM code used the divided sub-response matrices data set. The code used the MPI and OpenMP for the parallelized method. The median value is set as the average critical eigenvalues. The changes from the maximum value to the minimum value are 0.34 %{Delta}k with the spectral history method and 0.40 %{Delta}k without it, and the respective standard deviations were 0.12 % and 0.14 %. Using the spectral history method decreased the variation by 0.06 %{Delta}k. The root mean square differences of the axial power distribution were about 6 % between the analysis results and the plant data. Using the currents which converged in the previous exposure step reduced the number of iterations when the CR pattern changed only slightly. The averaged calculation time for each exposure step was about 5 hours on 12 PC Linux cluster servers with Core 2 Quad 3 GHz. (authors)

  13. KRAM, A lattice physics code for modeling the detailed depletion of gadolinia isotopes in BWR lattice designs

    SciTech Connect

    Knott, D.; Baratta, A. )

    1990-01-01

    Lattice physics codes are used to deplete the burnable isotopes present in each lattice design, calculate the buildup of fission products, and generate the few-group cross-section data needed by the various nodal simulator codes. Normally, the detailed depletion of gadolinia isotopes is performed outside the lattice physics code in a one-dimensional environment using an onion-skin model, such as the method used in MICBURN. Results from the onion-skin depletion, in the form of effective microscopic absorption cross sections for the gadolinia, are then used by the lattice physics code during the lattice-depletion analysis. The reactivity of the lattice at any point in the cycle depends to a great extent on the amount of gadolinia present. In an attempt to improve the modeling of gadolinia depletion from fresh boiling water reactor (BWR) fuel designs, the electric Power Research Institute (EPRI) lattice-physics code CPM-2 has been modified extensively. In this paper, the modified code KRAM is described, and results from various lattice-depletion analyses are discussed in comparison with results from standard CPM-2 and CASMO-2 analyses.

  14. Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

    DOE PAGESBeta

    Maljovec, D.; Liu, S.; Wang, B.; Mandelli, D.; Bremer, P. -T.; Pascucci, V.; Smith, C.

    2015-07-14

    Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated,more » where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.« less

  15. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    SciTech Connect

    Robb, Kevin R

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  16. Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

    SciTech Connect

    Maljovec, D.; Liu, S.; Wang, B.; Mandelli, D.; Bremer, P. -T.; Pascucci, V.; Smith, C.

    2015-07-14

    Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated, where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.

  17. Development of Probabilistic Risk Assessment Model for BWR Shutdown Modes 4 and 5 Integrated in SPAR Model

    SciTech Connect

    S. T. Khericha; S. Sancakter; J. Mitman; J. Wood

    2010-06-01

    Nuclear plant operating experience and several studies show that the risk from shutdown operation during modes 4, 5, and 6 can be significant This paper describes development of the standard template risk evaluation models for shutdown modes 4, and 5 for commercial boiling water nuclear power plants (BWR). The shutdown probabilistic risk assessment model uses full power Nuclear Regulatory Commission’s (NRC’s) Standardized Plant Analysis Risk (SPAR) model as the starting point for development. The shutdown PRA models are integrated with their respective internal events at-power SPAR model. This is accomplished by combining the modified system fault trees from SPAR full power model with shutdown event tree logic. For human reliability analysis (HRA), the SPAR HRA (SPAR-H) method is used which requires the analysts to complete relatively straight forward worksheet, including the performance shaping factors (PSFs). The results are then used to estimate HEP of interest. The preliminary results indicate the risk is dominated by the operator’s ability to diagnose the events and provide long term cooling.

  18. Effects of lateral separation of oxidic and metallic core debris on the BWR MK I containment drywell floor

    SciTech Connect

    Hyman, C.R.; Weber, C.F.; Hodge, S.A.

    1986-01-01

    In evaluating core debris/concrete interactions for a BWR MK I containment design, it is common practice to assume that at reactor vessel breach, the core debris is homogeneous and of low viscosity, so that it flows through the pedestal doorway and spreads in a radially uniform fashion throughout the drywell floor. In a recent study performed by the NRC-sponsored Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, calculations indicate that at reactor vessel bottom head failure, the debris temperature is such that the debris metals (Zr, Fe, Ni, Cr) are completely molten while the oxides (UO/sub 2/ ZrO/sub 2/, FeO) are completely frozen. Thus, the frozen oxides are expected to remain within the reactor pedestal while the molten metals radially separate from the frozen oxides, flow through the reactor pedestal doorway, and spread over the annular region of the drywell floor between the pedestal and the containment shell. This paper assesses the impact on calculated containment response and the production and release of fission product-laden aerosols for two different cases of debris distribution: uniform distribution and the laterally separated case of 95% oxides-5% metals inside the pedestal and 5% oxides-95% metals outside the pedestal. The computer codes used are CORCON-MOD2, MARCON 2.1B and VANESA.

  19. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

    SciTech Connect

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-07-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  20. Core structure heat-up and material relocation in a BWR short-term station blackout accident

    SciTech Connect

    Schmidt, R.C.; Dosanjh, S.S.

    1990-01-01

    This paper presents an analytical and numerical analysis which evaluates the core-structure heat-up and subsequent relocation of molten core materials during a NWR short-term station blackout accident with ADS. A simplified one-dimensional approach coupled with bounding arguments is first presented to establish an estimate of the temperature differences within a BWR assembly at the point when structural material first begins to melt. This analysis leads to the conclusions that the control blade will be the first structure to melt and that at this point in time, overall temperature differences across the canister-blade region will not be more than 200 K. Next, a three-dimensional heat-transfer model of the canister-blade region within the core is presented that uses a diffusion approximation for the radiation heat transfer. This is compared to the one-dimensional analysis to establish its compatibility. Finally, the extension of the three-dimensional model to include melt relocation using a porous media type approximation is described. The results of this analysis suggest that under these conditions significant amounts of material will relocate to the core plate region and refreeze, potentially forming a significant blockage. The results also indicate that a large amount of lateral spreading of the melted blade and canister material into the fuel rod regions will occur during the melt progression process. 22 refs., 18 figs., 1 tab.

  1. Use of scaled BWR lower plenum boron mixing tests to qualify the boron transport model used in TRACG

    SciTech Connect

    Cook, M. M.; Straka, M.; Chu, Y. C.; Heck, C. L.; Andersen, J. G. M.; Jacobs, R. H.

    2012-07-01

    In 2001 GEH applied best estimate methods combined with a statistical methodology to determine upper bound limits for key licensing parameters for anticipated operation occurrence (AOO) transient and anticipated transients without scram (ATWS) overpressure analyses for operating Boiling Water Reactors (BWRs). The methodology was subsequently extended for ESBWR AOO, ATWS, loss of coolant, and stability analyses. GEH is extending the methodology to long-term ATWS analyses for the operating BWRs. A long-term ATWS scenario uses injection of borated water to achieve reactor shutdown. Predicting the mixing and transport of boron is important for calculating the impact on the key licensing parameters. For the many operating BWRs where the denser boron solution is injected into the lower plenum, stratification may occur, delaying boron transport to the core region. CFD modeling can be used to model the stratification and mixing of the boron solution, but such calculations are extremely computer intensive and not cost effective; therefore, a more-empirical approach supported by a theoretical scaling of the dominant phenomena and backed by test data and benchmark calculations is used. The paper presents the TRACG lower plenum boron transport model qualification effort. The scaling basis used to implement the TRACG boron transport model for BWR applications is discussed. (authors)

  2. An overview of BWR Mark-I containment venting risk implications

    SciTech Connect

    Wagner, K.C.; Dallman, R.J.; Galyean, W.J.

    1988-11-01

    Venting of boiling water reactors with Mark-I containments has been suggested as a way to prevent catastrophic failure and/or mitigate the consequences resulting from a severe accident. Based on phenomenological, human factors, and risk considerations, the potential benefits and downsides of venting Mark-I containments were analyzed. Several generic venting systems and two proposed utility systems were reviewed. Based on generic considerations, the offsite consequences during risk dominant accidents were qualitatively assessed for four different vent systems. A quantitative risk study of an early venting strategy was performed, based on the existing Peach Bottom hardware and the draft NUREG-1150 results for Peach Bottom. Appendices are also included which contain reviews of the Pilgrim and Vermont Yankee venting submitals, a response to the seven questions from the NRC about the Pilgrim venting strategy, and a review of the venting strategy directed by Revision 4 of the Boiling Water Reactor Emergency Procedures Guidelines. 16 refs., 7 figs., 7 tabs.

  3. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    SciTech Connect

    Miro, R.; Maggini, F.; Barrachina, T.; Verdu, G.; Gomez, A.; Ortego, A.; Murillo, J. C.

    2006-07-01

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

  4. PWR core and spent fuel pool analysis using scale and nestle

    SciTech Connect

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  5. Parametric Study of Control Rod Exposure for PWR Burnup Credit Criticality Safety Analyses

    SciTech Connect

    Sanders, C.E.

    2001-07-20

    The Interim Staff Guidance on burnup credit (ISG-8) for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office, recommends the use of analyses that provide an ''adequate representation of the physics'' and notes particular concern with the ''need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted.'' In the absence of readily available information on the extent of control rod (CR) usage in U.S. PWRs and the subsequent reactivity effect of CR exposure on discharged SNF, NRC staff have indicated a need for greater understanding in these areas. In response, this paper presents results of a parametric study of the effect of CR exposure on the reactivity of discharged SNF for various CR designs (including Axial Power Shaping Rods), fuel enrichments, and exposure conditions (i.e., burnup and axial insertion). The study is performed in two parts. In the first part, two-dimensional calculations are performed, effectively assuming full axial CR insertion. These calculations are intended to bound the effect of CR exposure and facilitate comparisons of the various CR designs. In the second part, three-dimensional calculations are performed to determine the effect of various axial insertion conditions and gain a better understanding of reality. The results from the study demonstrate that the reactivity effect increases with increasing CR exposure (e.g., burnup) and decreasing initial fuel enrichment (for a fixed burnup). Additionally, the results show that even for significant burnup exposures, minor axial CR insertions (e.g., < 20 cm) result in an insignificant effect on the k{sub eff} of a spent fuel cask.

  6. International experience with a multidisciplinary table top exercise for response to a PWR accident

    SciTech Connect

    Lakey, J.R.A.

    1996-06-01

    Table Top Exercises are used for the training of emergency response personnel from a wide range of disciplines whose duties range from strategic to tactical, from managerial to operational. The exercise reported in this paper simulates the first two or three hours of an imaginary accident on a generic PWR site (named Seaside or Lakeside depending on its location). It is designed to exercise the early response of staff of the utility, government, local authority and the media and some players represent the public. The relatively few scenarios used for this exercise are based on actual events scaled to give off-site consequences which demand early assessment and therefore stress the communication procedures. The exercise is applicable in different cultures and has been used in over 20 short courses held in the USA, UK, Sweden, Prague, and Hong Kong. There are two styles of support for players: a linear program which ensures that all players follow the desired path through the event and an open program which is triggered by umpires (who play the reactor crew from a script) and by requests from other players. In both cases the exercise ends with a Press Conference. Players have an initial briefing and are assigned to roles; those who must speak at interviews and at the Press Conference arc given separate briefing by an expert in Public Affairs. The exercise runs with up to six groups and the communication rate reaches about 30 to 40 messages per hour for each group. The exercise can be applied to test management and communication systems and to study human response to emergencies because the merits of individual players are highlighted in the relatively stressful conditions of the initial stage of an accident. For some players the exercise is the first time that they have been required to carry out their task in front of other people.

  7. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

    SciTech Connect

    Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa; Xu, Yiban; Cao, Liping

    2016-01-01

    The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

  8. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    NASA Astrophysics Data System (ADS)

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

  9. On-line PWR RHR pump performance testing following motor and impeller replacement

    SciTech Connect

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  10. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    NASA Astrophysics Data System (ADS)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (E<0.625 eV) are required as these affect the CR integrity through stresses/strains induced by coupled clad embrittlement / absorber swelling phenomena. The concept of the PSI scheme under development is to provide from validated core analysis models, the volumetric neutron source to a full core MCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  11. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    NASA Astrophysics Data System (ADS)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  12. Aging assessment of the boiling-water reactor (BWR) standby liquid control system. Phase 1

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  13. Aging assessment of the boiling-water reactor (BWR) standby liquid control system

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  14. Description and assessment of RAMONA-3B Mod. 0 Cycle 4: a computer code with three-dimensional neutron kinetics for BWR system transients

    SciTech Connect

    Wulff, W; Cheng, H S; Diamond, D J; Khatib-Rahbar, M

    1984-01-01

    This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future.

  15. Mechanistic modeling of evaporating thin liquid film instability on a BWR fuel rod with parallel and cross vapor flow

    NASA Astrophysics Data System (ADS)

    Hu, Chih-Chieh

    This work has been aimed at developing a mechanistic, transient, 3-D numerical model to predict the behavior of an evaporating thin liquid film on a non-uniformly heated cylindrical rod with simultaneous parallel and cross flow of vapor. Interest in this problem has been motivated by the fact that the liquid film on a full-length boiling water reactor fuel rod may experience significant axial and azimuthal heat flux gradients and cross flow due to variations in the thermal-hydraulic conditions in surrounding subchannels caused by proximity to inserted control blade tip and/or the top of part-length fuel rods. Such heat flux gradients coupled with localized cross flow may cause the liquid film on the fuel rod surface to rupture, thereby forming a dry hot spot. These localized dryout phenomena can not be accurately predicted by traditional subchannel analysis methods in conjunction with empirical dryout correlations. To this end, a numerical model based on the Level Contour Reconstruction Method was developed. The Standard k-ε turbulence model is included. A cylindrical coordinate system has been used to enhance the resolution of the Level Contour Reconstruction Model. Satisfactory agreement has been achieved between the model predictions and experimental data. A model of this type is necessary to supplement current state-of-the-art BWR core thermal-hydraulic design methods based on subchannel analysis techniques coupled with empirical dry out correlations. In essence, such a model would provide the core designer with a "magnifying glass" by which the behavior of the liquid film at specific locations within the core (specific axial node on specific location within a specific bundle in the subchannel analysis model) can be closely examined. A tool of this type would allow the designer to examine the effectiveness of possible design changes and/or modified control strategies to prevent conditions leading to localized film instability and possible fuel failure.

  16. Investigation of the Effect of Fixed Absorbers on the Reactivity of PWR Spent Nuclear Fuel for Burnup Credit

    SciTech Connect

    Wagner, John C.; Sanders, Charlotta E.

    2002-08-15

    The effect of fixed absorbers on the reactivity of pressurized water reactor (PWR) spent nuclear fuel (SNF) in support of burnup-credit criticality safety analyses is examined. A fuel assembly burned in conjunction with fixed absorbers may have a higher reactivity for a given burnup than an assembly that has not used fixed absorbers. As a result, guidance on burnup credit, issued by the U.S. Nuclear Regulatory Commission's Spent Fuel Project Office, recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommendation eliminates a large portion of the currently discharged SNF from loading in burnup credit casks and thus severely limits the practical usefulness of burnup credit. Therefore, data are needed to support the extension of burnup credit to additional SNF. This research investigates the effect of various fixed absorbers, including integral burnable absorbers, burnable poison rods, control rods, and axial power shaping rods, on the reactivity of PWR SNF. Trends in reactivity with relevant parameters (e.g., initial fuel enrichment, burnup and absorber type, exposure, and design) are established, and anticipated reactivity effects are quantified. Where appropriate, recommendations are offered for addressing the reactivity effects of the fixed absorbers in burnup-credit safety analyses.

  17. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    SciTech Connect

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

  18. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    SciTech Connect

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria

  19. Analysis of a Defected Dissimilar Metal Weld in a PWR Power Plant

    SciTech Connect

    Efsing, P.; Lagerstrom, J.

    2002-07-01

    During the refueling outage 2000, inspections of the RC-loops of one of the Ringhals PWR-units, Ringhals 4, indicated surface breaking defects in the axial direction of the piping in a dissimilar weld between the Low alloy steel nozzle and the stainless safe end in the hot leg. In addition some indications were found that there were embedded defects in the weld material. These defects were judged as being insignificant to the structural integrity. The welds were inspected in 1993 with the result that no significant indications were found. The weld it self is a double U weld, where the thickness of the material is ideally 79,5 mm. Its is constructed by Inconel 182 weld material. At the nozzle a buttering was applied, also by Inconel 182. The In-service inspection, ISI, of the object indicated four axial defects, 9-16 mm deep. During fabrication, the areas where the defects are found were repaired at least three times, onto a maximum depth of 32 mm. To evaluate the defects, 6 boat samples from the four axial defects were cut from the perimeter and shipped to the hot-cell laboratory for further examination. This examination revealed that the two deep defects had been under sized by the ISI outside the requirement set by the inspection tolerances, while the two shallow defects were over sized, but within the tolerances of the detection system. When studying the safety case it became evident that there were several missing elements in the way this problems is handled with respect to the Swedish safety evaluation code. Among these the most notable at the beginning was the absence of reliable fracture mechanical data such as crack growth laws and fracture toughness at elevated temperature. Both these questions were handled by the project. The fracture mechanical evaluation has focused on a fit for service principal. Thus defects both in the unaffected zones and the disturbed zones, boat sample cutouts, of the weld have been analyzed. With reference to the Swedish safety

  20. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    NASA Astrophysics Data System (ADS)

    Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen; Zhou, Bangxin; Shoji, Tetsuo

    2016-04-01

    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T-L and L-T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T-L orientation with a higher crack growth rate than that in the specimen L-T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L-T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant.

  1. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J. ); Rey, J.M. )

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of unexpected'' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a new and improved'' state of the art has emerged recently.

  2. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J.; Rey, J.M.

    1992-05-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of ``unexpected`` instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a ``new and improved`` state of the art has emerged recently.

  3. Performance of iron-chromium-aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    NASA Astrophysics Data System (ADS)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun; Heuser, Brent J.; Mandapaka, Kiran K.; Was, Gary S.

    2016-03-01

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

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

    SciTech Connect

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

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

  5. Mechanistic determination of fission product releases for a Mark III BWR plant

    SciTech Connect

    Ludewig, H.; Yu, W.S.; Jaung, R.; Pratt, W.T.

    1986-01-01

    During the review of the GESSAR-II PRA by the Nuclear Regulatory Commission (NRC) and their contractors at Brookhaven National Laboratory (BNL) it was necessary to reanalyze potential fission product releases to the environment for several core meltdown accident sequences. The reanalysis was performed at BNL in two stages. The first stage was carried out prior to detailed mechanistic models were available at BNL and consisted of a sensitivity analysis using the MARCH and CORRAL computer codes. The effects of uncertainties in primary system retention, suppression pool scrubbing and core/concrete interactions on fission product release were handled by varying inputs to the MARCH/CORRAL codes. In this paper we outline the second stage of fission product release calculations, which was based on a system of codes developed under sponsorship of the Accident Source Term Program Office (ASTPO), USNRC. A comparison will be made between the range of source terms calculated by the first approach and the point estimate calculations provided by the more mechanistic codes. 8 refs., 2 tabs.

  6. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    NASA Astrophysics Data System (ADS)

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  7. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect

    Bi, G.; Liu, C.; Si, S.

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no

  8. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    SciTech Connect

    Hartini, Entin Andiwijayakusuma, Dinan

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  9. Effects of PbO on the oxide films of incoloy 800HT in simulated primary circuit of PWR

    NASA Astrophysics Data System (ADS)

    Tan, Yu; Yang, Junhan; Wang, Wanwan; Shi, Rongxue; Liang, Kexin; Zhang, Shenghan

    2016-05-01

    Effects of trace PbO on oxide films of Incoloy 800HT were investigated in simulated primary circuit water chemistry of PWR, also with proper Co addition. The trace PbO addition in high temperature water blocked the protective spinel oxides formation of the oxide films of Incoloy 800HT. XPS results indicated that the lead, added as PbO into the high temperature water, shows not only +2 valance but also +4 and 0 valances in the oxide film of 800HT co-operated with Fe, Cr and Ni to form oxides films. Potentiodynamic polarization results indicated that as PbO concentration increased, the current densities of the less protective oxide films of Incoloy 800HT decreased in a buffer solution tested at room temperature. The capacitance results indicated that the donor densities of oxidation film of Incoloy 800HT decreased as trace PbO addition into the high temperature water.

  10. Effect of temperature and dissolved hydrogen on oxide films formed on Ni and Alloy 182 in simulated PWR water

    NASA Astrophysics Data System (ADS)

    Mendonça, R.; Bosch, R.-W.; Van Renterghem, W.; Vankeerberghen, M.; de Araújo Figueiredo, C.

    2016-08-01

    Alloy 182 is a nickel-based weld metal, which is susceptible to stress corrosion cracking in PWR primary water. It shows a peak in SCC susceptibility at a certain temperature and hydrogen concentration. This peak is related to the electrochemical condition where the Ni to NiO transition takes place. One hypothesis is that the oxide layer at this condition is not properly developed and so the material is not optimally protected against SCC. Therefore the oxide layer formed on Alloy 182 is investigated as a function of the dissolved hydrogen concentration and temperature around this Ni/NiO transition. Exposure tests were performed with Alloy 182 and Ni coupons in a PWR environment at temperatures between 300 °C and 345 °C and dissolved hydrogen concentration between 5 and 35 cc (STP)H2/kg. Post-test analysis of the formed oxide layers were carried out by SEM, EDS and XPS. The exposure tests with Ni coupons showed that the Ni/NiO transition curve is at a higher temperature than the curve based on thermodynamic calculations. The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures, but that the morphology changed from spinel crystals to needle like oxides when the Ni/NiO transition curve was approached. Oxide layers were present below the Ni/NiO transition curve i.e. when the Ni coupon was still free of oxides. In addition an evolved slip dissolution model was proposed that could explain the observed experimental results and the peak in SCC susceptibility for Ni-based alloys around the Ni/NiO transition.

  11. Nuclear instrumentation, process instrumentation and control, and engineered safety features. Volume nine

    SciTech Connect

    Not Available

    1986-01-01

    Volume nine covers nuclear instrumentation (detection of nuclear radiation, gas-filled detectors, measuring neutron population, BWR/PWR nuclear instrumentation), process instrumentation and control (what is process instrumentation, pressure detectors and transducers, temperature detectors and transducers, level detectors and transducers, flow detectors and transducers, mechanical position detectors and transducers, what are the major processes controlled, BWR and PWR process instrumentation and control), engineered safety features (why are engineered safety features provided, the design basis accident, engineered safety feature operation, PWR engineered safety feature systems, BWR engineered safety feature systems).

  12. Effects of hydrogen water chemistry on corrosion fatigue behavior of cold-worked 304L stainless steel in simulated BWR coolant environments

    NASA Astrophysics Data System (ADS)

    Chiang, M. F.; Young, M. C.; Huang, J. Y.

    2011-04-01

    Corrosion fatigue behavior of stainless steel 304L (SS304L) in a simulated BWR coolant with hydrogen injection was investigated. Hydrogen water chemistry slightly mitigated the corrosion fatigue degradation of the as-received SS304L specimens, but, on the contrary, it slightly increased the corrosion fatigue crack growth rates (CFCGRs) of the cold-worked specimens. All the CFCGR-tested specimens showed similar fracture features, except for the amounts of deposited corrosion debris. The results indicated that decreasing the oxygen concentration of water environment is not an effective measure to suppress the fatigue crack growth rate of cold-worked SS304L. The CFCGRs of the SS304L were determined by an interaction between corrosion, oxide-induced crack closure and cold work in corrosive environments. At a specific level of reduction, cold work could enhance the corrosion fatigue resistance of SS304 both in the air-saturated and HWC coolant environments.

  13. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    SciTech Connect

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-10-15

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect.

  14. Combined numerical and experimental investigations of local hydrodynamics and coolant flow mass transfer in Kvadrat-type fuel assemblies of PWR reactors with mixing grids

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Samoilov, O. B.; Khrobostov, A. E.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Sorokin, V. D.

    2014-08-01

    Results of research works on studying local hydrodynamics and mass transfer for coolant flow in the characteristic zones of PWR reactor fuel assemblies in case of using belts of mixing spacer grids are presented. The investigations were carried out on an aerodynamic rig using the admixture diffusion method (the tracer-gas method). Certain specific features pertinent to coolant flow in the fuel rod bundles of Kvadrat-type fuel assemblies were revealed during the experiments. The obtained study results were included in the database for verifying computation fluid dynamics computer codes and detailed cell-wise calculations of reactor cores with Kvadrat-type fuel assemblies. The obtained results can also be used for more exact determination of local coolant flow hydrodynamic and mass transfer characteristics in assessing thermal reliability of PWR reactor cores.

  15. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

    SciTech Connect

    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter; Bertholdt, Horst-Otto; Adams, Andreas; Impertro, Michael; Roesch, Josef

    2013-07-01

    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  16. Evaluation of storing Shippingport Core II spent blanket fuel assemblies in the T Plant PWR Core II fuel pool without active cooling

    SciTech Connect

    Gilbert, E.R.; Lanning, D.D.; Dana, C.M.; Hedengren, D.C.

    1994-10-01

    PWR Core II fuel pool chiller-off test was conducted because it appeared possible that acceptable pool-water temperatures could be maintained without operating the chillers, thus saving hundreds of thousands of dollars in maintenance and replacement costs. Test results showed that the water-cooling capability is no longer needed to maintain pool temperature below 38{degrees}C (100{degrees}F).

  17. BWR steel containment corrosion

    SciTech Connect

    Tan, C.P.; Bagchi, G.

    1996-04-01

    The report describes regulatory actions taken after corrosion was discovered in the drywell at the Oyster Creek Plant and in the torus at the Nine Mile Point 1 Plant. The report describes the causes of corrosion, requirements for monitoring corrosion, and measures to mitigate the corrosive environment for the two plants. The report describes the issuances of generic letters and information notices either to collect information to determine whether the problem is generic or to alert the licensees of similar plants about the existence of such a problem. Implementation of measures to enhance the containment performance under severe accident conditions is discussed. A study by Brookhaven National Laboratory (BNL) of the performance of a degraded containment under severe accident conditions is summarized. The details of the BNL study are in the appendix to the report.

  18. Summary report on optimized designs for shipping casks containing 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect

    Bucholz, J.A.

    1983-04-01

    The purpose of this study was to develop new conceptual designs for large Pb, Fe, and U-shielded spent fuel casks which have been optimized for the shipment of 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel assemblies. Design specifications for about 100 cases of potential interest are presented along with a brief 20-page synopsis of the associated analyses. Optimized shielding requirements are presented for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. With respect to criticality, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. Steady-state and transient heat transfer analyses for casks under nominal and accident conditions were performed using the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. Based on criticality, shielding, and heat transfer considerations, it appears that optimized cask designs could be developed to carry 15 to 18 five-year-old PWR fuel assemblies or as many as 18 to 21 ten-year-old PWR fuel assemblies. 4 figures, 4 tables.

  19. Scoping design analyses for optimized shipping casks containing 1-, 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect

    Bucholz, J.A.

    1983-01-01

    This report details many of the interrelated considerations involved in optimizing large Pb, Fe, or U-metal spent fuel shipping casks containing 1, 2, 3, 5, 7, or 10-year-old PWR fuel assemblies. Scoping analyses based on criticality, shielding, and heat transfer considerations indicate that some casks may be able to hold as many as 18 to 21 ten-year-old PWR fuel assemblies. In the criticality section, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. In another section, details of many n/..gamma.. shielding optimization studies are presented, including the optimal n/..gamma.. design points and the actual shielding requirements for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. Multigroup source terms based on ORIGEN2 calculations at these and other decay times are also included. Lastly, the numerical methods and experimental correlations used in the steady-state and transient heat transfer analyses are fully documented, as are pertinent aspects of the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. (While only casks for square, intact PWR fuel assemblies were considered in this study, the SCOPE code may also be used to design and analyze casks containing canistered spent fuel or other waste material. An abbreviated input data guide is included as an appendix).

  20. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    NASA Astrophysics Data System (ADS)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  1. Comparison of PWR - Burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results

    SciTech Connect

    Oberle, P.; Broeders, C. H. M.; Dagan, R.

    2006-07-01

    The increasing tendency towards fuel lifetime extension in thermal nuclear reactors motivated validation work for available evaluation tools for nuclear fuel burnup calculations. In this study two deterministic codes with different transport solvers and one Monte Carlo method are investigated. The code system KAPROS/KARBUS uses the classical deterministic First Collision Probability method utilizing a cylinderized Wigner-Seitz cell. In the SCALES.0/TRITON/NEWT code the Extended Step Characteristic method is applied. In a first step the two deterministic codes are compared with experimental results from the KWO-Isotope Correlation Experiment up to 30 MWD/kg HM burnup, published in 1981. Two pin cell calculations are analyzed by comparison of calculated and experimental results for important heavy isotope vectors. The results are very satisfactory. Subsequently, further validation at higher burnup (< 80 MWD/kg HM) is provided by comparison of the two deterministic codes and the Monte Carlo based burnup code MONTEBURNS for PWR UO{sub 2} fuel assembly calculations. Possible reasons for differences in the results are analyzed and discussed. Especially the influence of cross section data and processing is presented. (authors)

  2. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    SciTech Connect

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  3. Presentation of the MERC work-flow for the computation of a 2D radial reflector in a PWR

    SciTech Connect

    Clerc, T.; Hebert, A.; Leroyer, H.; Argaud, J. P.; Poncot, A.; Bouriquet, B.

    2013-07-01

    This paper presents a work-flow for computing an equivalent 2D radial reflector in a pressurized water reactor (PWR) core, in adequacy with a reference power distribution, computed with the method of characteristics (MOC) of the lattice code APOLLO2. The Multi-modelling Equivalent Reflector Computation (MERC) work-flow is a coherent association of the lattice code APOLLO2 and the core code COCAGNE, structured around the ADAO (Assimilation de Donnees et Aide a l'Optimisation) module of the SALOME platform, based on the data assimilation theory. This study leads to the computation of equivalent few-groups reflectors, that can be spatially heterogeneous, which have been compared to those obtained with the OPTEX similar methodology developed with the core code DONJON, as a first validation step. Subsequently, the MERC work-flow is used to compute the most accurate reflector in consistency with all the R and D choices made at Electricite de France (EDF) for the core modelling, in terms of number of energy groups and simplified transport solvers. We observe important reductions of the power discrepancies distribution over the core when using equivalent reflectors obtained with the MERC work-flow. (authors)

  4. Linking Grain Boundary Microstructure to Stress Corrosion Cracking of Cold Rolled Alloy 690 in PWR Primary Water

    SciTech Connect

    Bruemmer, Stephen M.; Olszta, Matthew J.; Toloczko, Mychailo B.; Thomas, Larry E.

    2012-10-01

    Grain boundary microstructures and microchemistries are examined in cold-rolled alloy 690 tubing and plate materials and comparisons are made to intergranular stress corrosion cracking (IGSCC) behavior in PWR primary water. Chromium carbide precipitation is found to be a key aspect for materials in both the mill annealed and thermally treated conditions. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG carbide distribution. For the same degree of cold rolling, alloys with few IG precipitates exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal that cracked carbides appeared to blunt propagation of IGSCC cracks in many cases. Preliminary results suggest that the localized grain boundary strains and stresses produced during cold rolling promote IGSCC susceptibility and not the cracked carbides and voids.

  5. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    SciTech Connect

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  6. OBSERVATIONS AND IMPLICATIONS OF INTERGRANULAR STRESS CORROSION CRACK GROWTH OF ALLOY 152 WELD METALS IN SIMULATED PWR PRIMARY WATER

    SciTech Connect

    Toloczko, Mychailo B.; Olszta, Matthew J.; Overman, Nicole R.; Bruemmer, Stephen M.

    2013-08-15

    Significant intergranular (IG) crack growth during stress corrosion cracking (SCC) tests has been documented during tests in simulated PWR primary water on two alloy 152 specimens cut from a weldment produced by ANL. The cracking morphology was observed to change from transgranular (TG) to mixed mode (up to ~60% IG) during gentle cycling and cycle + hold loading conditions. Measured crack growth rates under these conditions often suggested a moderate degree of environmental enhancement consistent with faster growth on grain boundaries. However, overall SCC propagation rates at constant stress intensity (K) or constant load were very low in all cases. Initial SCC rates up to 6x10-9 mm/s were occasionally measured, but constant K/load growth rates dropped below ~1x10-9 mm/s with time even when significant IG engagement existed. Direct comparisons were made among loading conditions, measured crack growth response and cracking morphology during each test to assess IGSCC susceptibility of the alloy 152 specimens. These results were analyzed with respect to our previous SCC crack growth rate measurements on alloy 152/52 welds.

  7. Development of modified MDA (M-MDA), PWR fuel cladding tube for high duty operation in future

    SciTech Connect

    Watanabe, Seiichi; Kido, Toshiya; Arakawa, Yasushi

    2007-07-01

    A new cladding material of M-MDA has been developed in order to prepare for a strong growing demand for advanced fuel which can maintain its integrity even under high duties due to more efficient operation such as higher burnup, higher LHR, and longer operation cycle which will contribute the suppression of environmental burdens like CO{sub 2} emission. The main aim of M-MDA is to have excellent corrosion resistance while the other properties are inherited from MDA, which has been adopted to the step 2 fuel, instead of Zry-4, of Japanese PWR plant whose upper limit of assembly discharged burnup is 55 MWd/kgU. And we could confirm that the main aim of M-MDA was achieved by means of out-of-pile tests. In order to confirm improvement of corrosion resistance of M-MDA in the actual operation, irradiation test of M-MDA in the commercial reactor of Vandellos II is ongoing. The latest results of on-site examination after every end of cycle showed that oxide thickness of M-MDA-SR was much smaller than that of MDA at rod discharged burnup of approximately 60 MWd/kgU. The final irradiation cycle was completed on April 2007 and then we will obtain corrosion data of M-MDA over 70 MWd/kgU. M-MDA is a candidate alloy for advanced fuel under higher duty usage. (authors)

  8. Improvement of the thermal margins in the Swedish Ringhals-3 PWR by introducing new fuel assemblies with thorium

    SciTech Connect

    Lau, C. W.; Demaziere, C.; Nylen, H.; Sandberg, U.

    2012-07-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)

  9. Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

    SciTech Connect

    Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R.

    2006-07-01

    High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)

  10. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    NASA Astrophysics Data System (ADS)

    LaFleur, Adrienne M.; Menlove, Howard O.

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies.

  11. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  12. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  13. Identification and Resolution of Safety Issues for the Advanced Integral Type PWR

    SciTech Connect

    Kim, Woong Sik; Jo, Jong Chull; Yune, Young Gill; Kim, Hho Jung

    2004-07-01

    This paper presents the interim results of a study on the identification and resolution of safety issues for the AIPWR licensing. The safety issues discussed in this paper include (1) policy issues for which decision-makings are needed for the procedural requirements of licensing system in the regulatory policy point of view, (2) technical issues for which either development of new requirements or amendment of some existing requirements is needed, or (3) other technical issues for which safety verifications are required. The study covers (a) the assessment of applicability of the issues identified from the previous studies to the case of the AIPWR, (b) identification of safety issues through analysis of the international experiences in the design and licensing of advanced reactors, and technical review of the AIPWR design, and (c) development of the resolutions of safety issues, and application of the resolutions to the amendment of regulatory requirements and the licensing review of the AIPWR. As the results of this study, a total of twenty eight safety issues was identified: fourteen issues from the previous studies, including the establishment of design safety goals; four issues from the foreign practices and experiences, including the risk-informed licensing; and ten issues by the AIPWR design review, including reliability of passive safety systems. Ten issues of them have been already resolved and the succeeding study is under way to resolve the remaining ones. (authors)

  14. Proceedings: 2003 EPRI Workshop on Condensate Polishing

    SciTech Connect

    2004-02-01

    Successful condensate polishing operations maintain control of ionic and particulate impurity transport to the pressurized water reactor (PWR) steam generator and the boiling water reactor (BWR) reactor and recirculation system, thus allowing the units to operate more reliably. This report contains the work presented at EPRI's 2003 Workshop on Condensate Polishing, where 30 papers were presented on current issues, research, and utility experiences involving polishing issues at both PWR and BWR units.

  15. Phenomena and Parameters Important to Burnup Credit

    SciTech Connect

    Parks, C.V.

    2001-01-10

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.

  16. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    SciTech Connect

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions.

  17. Statistical Safety Evaluation of BWR Turbine Trip Scenario Using Coupled Neutron Kinetics and Thermal Hydraulics Analysis Code SKETCH-INS/TRACE5.0

    NASA Astrophysics Data System (ADS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.

  18. Compilation of corrosion data on CAN-DECON. Volume 1. General, galvanic, crevice, and pitting corrosion data from CANDU and BWR tests. Final report

    SciTech Connect

    Michalko, J.P.; Bonnici, P.J.; Smee, J.L.

    1985-10-01

    Nuclear power station ALARA radiation exposure criteria require, in many cases, decontamination of specific equipment or systems before maintenance, inspection, or work in an adjacent high radiation area. Chemical decontamination, which can be performed away from the high radiation fields, can often best satisfy these ALARA exposure criteria. CAN-DECON, a dilute chemical decontamination process was developed to meet the needs of the Canadian CANDU reactors. It was found to be effective in dissolving BWR oxide films that contain the entrapped radioactive species contributing to high radiation fields. During the development phase of the process and during subsequent field application, CAN-DECON has undergone extensive testing to determine the extent of oxide film dissolution and the degree of corrosion of materials used in construction of reactor components. This has been accomplished on many of the various materials of construction found in the components of the systems decontaminated. Materials tested include carbon steels with range of carbon content 0.1 to 0.4 wt %, 300 series, 400 series, and specialty stainless steels, low alloy steels, and gasket and seal materials. CAN-DECON caused little or no significant corrosion or deterioration on any of the materials tested when applied under conditions appropriate to that class of material. 2 figs., 63 tabs.

  19. Crack initiation testing of thimble tube material under PWR conditions to determine a stress threshold for IASCC

    NASA Astrophysics Data System (ADS)

    Bosch, R. W.; Vankeerberghen, M.; Gérard, R.; Somville, F.

    2015-06-01

    IASCC (Irradiation Assisted Stress Corrosion Cracking) crack initiation tests have been carried out on thimble tube material retrieved from a Belgian PWR. The crack initiation tests were carried out by constant load testing of thimble tube specimens at different stress levels. The time-to-failure was determined as a function of the applied stress to find a stress threshold under which no stress corrosion cracking will take place. The thimble tube was made of 316L cold-worked stainless steel and the dose profile along the thimble tube ranges from 45 to 80 dpa. This allows adding crack initiation data for dose values that have not been significantly reported, i.e. in the range of 45-55 dpa and at 80 dpa. The results can be used to determine whether the stress under which no IASCC occurs saturates for a dose larger than 30 dpa or whether a small further threshold decrease with dose can be observed. Over a period of four years, more than 40 specimens have been tested with doses ranging from 45 to 80 dpa at stress levels between 40% and 70% of the irradiated yield stress. Fracture occurred at all stress levels (but not all specimens) although the time-to-failure increased with decreasing stress. The results show that intergranular cracking was the main fracture mode in all failed O-rings. Three of six 80 dpa O-rings subjected to 40% and 45% of the yield stress did not fail after six months of testing. Based on these results and a comparison with literature data, an apparent stress limit for IASCC could be estimated at 40% of the irradiated yield stress.

  20. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  1. Advanced Fabrication Technique and Thermal Performance Prediction of U-Mo/Zr-alloy Dispersion Fuel Pin for High Burnup PWR

    NASA Astrophysics Data System (ADS)

    Suwardi

    2010-06-01

    In recent years, a novel class of zirconium alloys having the melting temperature of 990-1160 K has been developed. Based on novel zirconium matrix alloys, high uranium content fuel pin with U-9Mo has been developed according to capillary impregnation technique. The pin shows it is thermal conductivity ranging from 18 to 22 w/m/K that is comparably higher than UO2 pellet pin. The paper presents the met-met fabrication and thermal performance analysis of the fuel in typical PWR. The fabrication consists of mixing UO2 powder or granules and a novel Zr-alloy powder having low melting point, filling the mixture in a cladding tube that one of its end has been plugged, heating the pin to above melting temperature of Zr-alloy for an hour, natural cooling and heat treating at 300 K for 1/2 hr. The thermal analysis takes into account the pore and temperature distribution and high burn up effect to pellet conductivity. The thermal diffusivity ratio of novel to conventional fuel has been used as correction factor for the novel fuel conductivity. The results show a significant lowering pellet temperature along the radius until 1000 K at the hottest position. The analysis underestimates since the gap conductivity has been treated as decreased by 2% fission gas released that is not real since the use of lower temperature, and also decreasing thermal conductivity by porosity formation will much lower. The analysis shows that the novel fuel has very good thermal properties which able to pass the barrier of 65 MWD/kg-U, the limit to day commercial fuel. The burn-up extension means fewer fresh fuel is needed to produce electricity, preserve natural uranium resource, easier fuel handling operational per energy produced

  2. Rod consolidation of RG and E's (Rochester Gas and Electric Corporation) spent PWR (pressurized water reactor) fuel

    SciTech Connect

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister.

  3. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    SciTech Connect

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  4. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    SciTech Connect

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  5. Los Alamos PWR decay-heat-removal studies. Summary results and conclusions

    SciTech Connect

    Boyack, B E; Henninger, R J; Horley, E; Lime, J F; Nassersharif, B; Smith, R

    1986-03-01

    The adequacy of shutdown-decay-heat removal in pressurized water reactors (PWRs) is currently under investigation by the Nuclear Regulatory Commission. One part of this effort is the review of feed-and-bleed procedures that could be used if the normal cooling mode through the steam generators were unavailable. Feed-and-bleed cooling is effected by manually activating the high-pressure injection (HPI) system and opening the power-operated relief valves (PORVs) to release the core decay energy. The feasibility of the feed-and-bleed concept as a diverse mode of heat removal has been evaluated at the Los Alamos National Laboratory. The TRAC-PF1 code has been used to predict the expected performance of the Oconee-1 and Calvert Cliffs-1 reactors of Bobcock and Wilcox and Combustion Engineering, respectively, and the Zion-1 and H.B. Robinson-2 plants of Westinghouse. Feed and bleed was successfully applied in each of the four plants studied, provided it was initiated no later than the time of loss of secondary heat sink. Feed and bleed was successfully applied in two of the plants, Oconee-1 and Zion-1, provided it was initiated no later than the time of primary system saturation. Feed and bleed in Calvert Cliffs-1 when initiated at the time of primary system saturation did result in core dryout; however, the core heatup was eventually terminated by coolant injection. Feed-and-bleed initiation at primary system saturation was not studied for H.B. Robinson-2. Insights developed during the analyses of specific plant transients have been identified and documented. 33 refs., 107 figs., 26 tabs.

  6. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    SciTech Connect

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  7. Study of Two-Phase Heat Transfer in Nano-fluids for Nuclear Applications

    SciTech Connect

    Kim, S.J.; Truong, B.; Buongiorno, J.; Hu, L.W.; Bang, I.C.

    2006-07-01

    Nano-fluids are engineered colloidal suspensions of nano-particles in a base fluid. We are investigating the two-phase heat transfer behavior of water-based nano-fluids, to evaluate their potential use in nuclear applications, including the PWR primary coolant and PWR and BWR safety systems. A simple pool boiling wire experiment shows that a significant increase in Critical Heat Flux (CHF) can be achieved at modest nano-particle concentrations. For example, the CHF increases by 50% in nano-fluids with alumina nano-particles at 0.001%v concentration. The CHF enhancement appears to correlate with the presence of a layer of nano-particles that builds up on the heated surface during nucleate boiling. A review of the prevalent Departure from Nucleate Boiling (DNB) theories suggests that an alteration of the nucleation site density (brought about by the nano-particle layer) could plausibly explain the CHF enhancement. (authors)

  8. Performance Spec. for Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shipping Port Spent Fuel Canisters

    SciTech Connect

    JOHNSON, D.M.

    2000-03-14

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders.

  9. COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.

    1986-12-01

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.

  10. New innovative electrocoagulation (EC) treatment technology for BWR colloidal iron utilizing the seeding and filtration electronically (SAFET{sup TM}) system

    SciTech Connect

    Denton, Mark S.; Bostick, William D.

    2007-07-01

    The presence of iron (iron oxide from carbon steel piping) buildup in Boiling Water Reactor (BWR) circuits and wastewaters is decades old. In, perhaps the last decade, the advent of precoatless filters for condensate blow down has compounded this problem due to the lack of a solid substrate (e.g., Powdex resin pre-coat) to help drop the iron out of solution. The presence and buildup of this iron in condensate phase separators (CPS) further confounds the problem when the tank is decanted back to the plant. Iron carryover here is unavoidable without further treatment steps. The form of iron in these tanks, which partially settles and is pumped to a de-waterable high integrity container (HIC), is particularly difficult and time consuming to de-water (low shear strength, high water content). The addition upstream from the condensate phase separator (CPS) of chemicals, such as polymers, to carry out the iron, only produces an iron form even more difficult to filter and de-water (even less shear strength, higher water content, and a gel/slime consistency). Typical, untreated colloidal material contains both sub-micron particles up to, let's say 100 micron. It is believed that the sub-micron particles penetrate filters, or sheet filters, thus plugging the pores for what should have been the successful filtration of the larger micron particles. Like BWR iron wastewaters, fuel pools/storage basins (especially in the decon. phase) often contain colloids which make clarity and the resulting visibility nearly impossible. Likewise, miscellaneous, often high conductivity, waste streams at various plants contain such colloids, iron, salts (sometimes seawater intrusion and referred to as Salt Water Collection Tanks), dirt/clay, surfactants, waxes, chelants, etc. Such waste streams are not ideally suited for standard dead-end (cartridges) or cross-flow filtration (UF/RO) followed even by demineralizers. Filter and bed plugging are almost assured. The key to solving these dilemmas

  11. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    SciTech Connect

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment.

  12. Review of APR+ Level 2 PSA. Revision 2

    SciTech Connect

    Lehner, John R.; Mubayi, Vinod; Pratt, W. Trevor; Kim, Do Sam; Cho, Yong Jin; Cho, Sang Jin; Kim, In Goo

    2012-02-17

    Brookhaven National Laboratory (BNL) assisted the Korea Institute of Nuclear Safety (KINS) in reviewing the Level 2 Probabilistic Safety Assessment (PSA) of the APR+ Advanced Pressurized Water Reactor (PWR) prepared by the Korea Hydro & Nuclear Power Co., Ltd (KHNP) and KEPCO Engineering & Construction Co., Inc. (KEPCO-E&C). The work described in this report involves a review of the APR+ Level 2 PSA submittal [Ref. 1]. The PSA and, therefore, the review is limited to consideration of accidents initiated by internal events. As part of the review process, the review team also developed three sets of Requests for Additional Information (RAIs). These RAIs were provided to KHNP and KEPCO-E&C for their evaluation and response. This final detailed report documents the review findings for each technical element of the PSA and includes consideration of all of the RAIs made by the reviewers as well as the associated responses. This final report was preceded by an interim report [Ref. 2] that focused on identifying important issues regarding the PSA. In addition, a final meeting on the project was held at BNL on November 21-22, 2011, where BNL and KINS reviewers discussed their preliminary review findings with KHNP and KEPCO-E&C staffs. Additional information obtained during this final meeting was also used to inform the review findings of this final report. The review focused not only on the robustness of the APR+ design to withstand severe accidents, but also on the capability and acceptability of the Level 2 PSA in terms of level of detail and completeness. The Korean nuclear regulatory authorities will decide whether the PSA is acceptable and the BNL review team is providing its comments for KINS consideration. Section 2.0 provides the basis for the BNL review. Section 3.0 presents the review of each technical element of the PSA. Conclusions and a summary are presented in Section 4.0. Section 5.0 contains the references.

  13. Effects of long-term thermal aging on the stress corrosion cracking behavior of cast austenitic stainless steels in simulated PWR primary water

    NASA Astrophysics Data System (ADS)

    Li, Shilei; Wang, Yanli; Wang, Hui; Xin, Changsheng; Wang, Xitao

    2016-02-01

    The stress corrosion cracking (SCC) behavior of cast austenitic stainless steels of unaged and thermally aged at 400 °C for as long as 20,000 h were studied by using a slow strain rate testing (SSRT) system. Spinodal decomposition in ferrite during thermal aging leads to hardening in ferrite and embrittlement of the SSRT specimen. Plastic deformation and thermal aging degree have a great influence on the oxidation rate of the studied material in simulated PWR primary water environments. In the SCC regions of the aged SSRT specimen, the surface cracks, formed by the brittle fracture of ferrite phases, are the possible locations for SCC. In the non-SCC regions, brittle fracture of ferrite phases also occurs because of the effect of thermal aging embrittlement.

  14. LOCA hydroloads calculations with multidimensional nonlinear fluid/structure interaction. Volume 1: STEALTH 1D single-phase fluid studies. Final report. [PWR

    SciTech Connect

    Santee, G.E. Jr.; Mortensen, G.A.; Caraher, D.L.

    1980-04-01

    This report, which is the first in a series of reports for RP-1065, describes the first step in the stepwise approach for developing the methodology to assess the hydroloads on a large PWR during the subcooled portions of a hypothetical LOCA. The first step in the methodology considers enhancements and special modifications to the 1D STEALTH computer code in order that acoustic phenomena in piping and vessel networks may be simulated. The resulting code is termed 1D STEALTH-HYDRO. The 1D STEALTH-HYDRO enhancements consist of three control volume models to simulate area changes, orifices, and tees in piping networks. The theory of the control volume models is described.

  15. LOCA hydroloads calculations with multidimensional nonlinear fluid/structure interaction. Volume 3. Fluid/structure interaction studies using 3-D STEALTH/WHAMSE. Final report. [PWR

    SciTech Connect

    Santee, G.E. Jr.; Chang, F.H.; Mortensen, G.A.; Brockett, G.F.; Gross, M.B.; Belytschko, T.B.

    1982-11-01

    This report, the third in a series of reports for RP-1065, describes the final step in the stepwise approach for developing the three-dimensional, nonlinear, fluid-structure interaction methodology to assess the hydroloads on a large PWR during the subcooled portions of a hypothetical LOCA. The final step in the methodology implements enhancements and special modifications to the STEALTH 3D computer program and the WHAMSE 3D computer program. After describing the enhancements, the individual and the coupled computer programs are assessed by comparing calculational results with either analytical solutions or with experimental data. The coupled 3D STEALTH/WHAMSE computer program is then applied to the simulation of HDR Test V31.1 to further assess the program and to investigate the role that fluid-structure interaction plays in the hydrodynamic loading of reactor internals during subcooled blowdown.

  16. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

  17. Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly

    NASA Astrophysics Data System (ADS)

    da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2014-04-01

    Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

  18. Spent fuel dry storage technology development: thermal evaluation of isolated drywells containing spent fuel (1 kW PWR spent fuel assembly)

    SciTech Connect

    Unterzuber, R; Wright, J B

    1980-09-01

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site. Two PWR spent fuel assemblies having a decay heat level of approximately 1.1 kW were encapsulated inside the E-MAD Hot Bay and placed in instrumented near-surface drywell storage cells. Temperatures from the two isolated drywells and the adjacent soil have been recorded throughout the 19 month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (254{sup 0}F and 203{sup 0}F, respectively) during August 1979. Thereafter, all temperatures decreased and showed a cycling pattern which responded to seasonal atmospheric temperature changes. A computer model was utilized to predict the thermal response of the drywell. Computer predictions of the drywell temperatures and the temperatures of the surrounding soil are presented and show good agreement with the test data.

  19. PWR FLECHT SEASET 21-rod bundle flow blockage task data and analysis report. NRC/EPRI/Westinghouse Report No. 11. Appendices K-P

    SciTech Connect

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  20. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  1. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  2. Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

    SciTech Connect

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.

    2008-05-05

    In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.

  3. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  4. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    NASA Astrophysics Data System (ADS)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  5. Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

    NASA Astrophysics Data System (ADS)

    Salko, Robert K.

    COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for

  6. Probabilistic pipe fracture evaluations for leak-rate-detection applications

    SciTech Connect

    Rahman, S.; Ghadiali, N.; Paul, D.; Wilkowski, G.

    1995-04-01

    Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crack size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.

  7. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    SciTech Connect

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurin; Natesan, Ken

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  8. Constraints on silicates formation in the Si-Al-Fe system: Application to hard deposits in steam generators of PWR nuclear reactors

    NASA Astrophysics Data System (ADS)

    Berger, Gilles; Million-Picallion, Lisa; Lefevre, Grégory; Delaunay, Sophie

    2015-04-01

    Introduction: The hydrothermal crystallization of silicates phases in the Si-Al-Fe system may lead to industrial constraints that can be encountered in the nuclear industry in at least two contexts: the geological repository for nuclear wastes and the formation of hard sludges in the steam generator of the PWR nuclear plants. In the first situation, the chemical reactions between the Fe-canister and the surrounding clays have been extensively studied in laboratory [1-7] and pilot experiments [8]. These studies demonstrated that the high reactivity of metallic iron leads to the formation of Fe-silicates, berthierine like, in a wide range of temperature. By contrast, the formation of deposits in the steam generators of PWR plants, called hard sludges, is a newer and less studied issue which can affect the reactor performance. Experiments: We present here a preliminary set of experiments reproducing the formation of hard sludges under conditions representative of the steam generator of PWR power plant: 275°C, diluted solutions maintained at low potential by hydrazine addition and at alkaline pH by low concentrations of amines and ammoniac. Magnetite, a corrosion by-product of the secondary circuit, is the source of iron while aqueous Si and Al, the major impurities in this system, are supplied either as trace elements in the circulating solution or by addition of amorphous silica and alumina when considering confined zones. The fluid chemistry is monitored by sampling aliquots of the solution. Eh and pH are continuously measured by hydrothermal Cormet© electrodes implanted in a titanium hydrothermal reactor. The transformation, or not, of the solid fraction was examined post-mortem. These experiments evidenced the role of Al colloids as precursor of cements composed of kaolinite and boehmite, and the passivation of amorphous silica (becoming unreactive) likely by sorption of aqueous iron. But no Fe-bearing was formed by contrast to many published studies on the Fe

  9. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    SciTech Connect

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

  10. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  11. Dryout of BWR fuel elements

    SciTech Connect

    Reisch, Frigyes

    2006-07-01

    To increase the power output of the presently operating power reactors is a worldwide trend. One limiting factor from the safety and commercial point of views is the maximum allowable thermal load of the fuel. The findings of the presented loop experiments are that the margin to the burnout of the fuel elements can be defined by a single parameter the void. (authors)

  12. Reviewing the Reviewers

    ERIC Educational Resources Information Center

    Winkler, Karen J.

    2009-01-01

    Most professors have mixed feelings about participating on peer-review panels. It's an honor. It helps the discipline. It's a waste of time. It's biased. Michele Lamont wanted to know whether it works: specifically, whether, and how, professors identify excellence. So the multi-titled Harvard University scholar--professor of European studies,…

  13. Mitsubishi technical review. Volume 20, Number 2. Series Number 57, 1983

    SciTech Connect

    Not Available

    1983-06-01

    Contents: Features of Current PWR Power Plant (General Review); Development of a Plant Operation Guidance System; Total Radioactive Solid Waste Processing System; Recent Technology Trend in PWR Turbine Plant; Remote Operation and Automation of the Maintenance Works under the Radiation Field (illustration); Development of Arc-Welding Robot for Nozzle of Boiler Drum; Field Study on Mechanical Strength of Sea Ice at East Coast of Hokkaido; A Study of Elementary Control Techniques for Car Air Conditioner; Radioactive Solid Waste Storage System; Completion of Coal-fired Marine Boiler; Mitsubishi Hydraulic Loading Shovel MS580; Mitsubishi Dry Type Mechanical Screen (Rack Type); Mitsubishi L300 '83 Year Model; 1983 Year Model Mitsubishi CANTER 1.5-3.5 Tonner Trucks; Large Sized Touring Coach, MS7 Series; Oceanographic Research Vessel 'TANSEI MARU';Mitsubishi Dredging Assist System for Cutter Suction Dredger 'MIDAS-C';Mitsubishi Testing Plant of Biaxially Oriented Film Production; Refreshing Breeze in All Seasons with Mitsubishi Daiya Ceiling Recessed Packaged Type Heat Pump Air Conditioner.

  14. Simulation of PWR plant by a new version of TRAC-PF1 code including a three-dimensional neutronic model and a transport boron model

    SciTech Connect

    Alloggio, G.; Brega, E.; Basile, D.; Guandalini, R.; Pollachini, L.

    1996-08-01

    This paper is aimed at presenting a solution method for time-dependent problems coupling thermal-hydraulic behavior and neutronic changes during selected transients in Pressurized Water Reactors. A two-group three-dimensional reactor kinetics model, based on the Analytical Nodal Method with a detailed feedback model, has been implemented in TRAC-PF1 code replacing the original point-kinetics approximation. A geometry conversion was done to match, in the core discretization, the cylindrical geometry of the TRAC-PF1 with the Cartesian geometry of the three-dimensional neutronic model. In this version of TRAC-PFI, named TRAC-EN for IBM computers, a Boron transport model has been implemented. The transient Boron concentration is computed from a Boron mass balance after the coolant mass, energy and momentum balances have been completed. In order to evaluate the new code capabilities, a model of the two-loops 600 MW Westinghouse reactor was implemented. Some specific PWR transients that exhibit interesting nuclear and thermal-hydraulic responses, e.g., control rod ejection and pressurization transients, are presented. To check the Boron transport model, a local Boron dilution transient was analyzed. The results obtained by using the space and time dependent neutronic model can not be predicted by point kinetics approximation. Furthermore, some events that apparently concern only the core, are also involving the primary circuit the responses of which can not be neglected because they affect the neutronic behavior.

  15. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    SciTech Connect

    Lafleur, Adrienne M; Charlton, William S; Menlove, Howard O; Swinhoe, Martyn T

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  16. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    SciTech Connect

    Unterzuber, R.; Wright, J.B.

    1980-09-01

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  17. A Study on Structured Simulation Framework for Design and Evaluation of Human-Machine Interface System -Application for On-line Risk Monitoring for PWR Nuclear Power Plant-

    SciTech Connect

    Zhan, J.; Yang, M.; Li, S.C.; Peng, M.J.; Yan, S.Y.; Zhang, Z.J.

    2006-07-01

    The operators in the main control room of Nuclear Power Plant (NPP) need to monitor plant condition through operation panels and understand the system problems by their experiences and skills. It is a very hard work because even a single fault will cause a large number of plant parameters abnormal and operators are required to perform trouble-shooting actions in a short time interval. It will bring potential risks if operators misunderstand the system problems or make a commission error to manipulate an irrelevant switch with their current operation. This study aims at developing an on-line risk monitoring technique based on Multilevel Flow Models (MFM) for monitoring and predicting potential risks in current plant condition by calculating plant reliability. The proposed technique can be also used for navigating operators by estimating the influence of their operations on plant condition before they take an action that will be necessary in plant operation, and therefore, can reduce human errors. This paper describes the risk monitoring technique and illustrates its application by a Steam Generator Tube Rupture (SGTR) accident in a 2-loop Pressurized Water Reactor (PWR) Marine Nuclear Power Plant (MNPP). (authors)

  18. On the Application of CFD Modeling for the Prediction of the Degree of Mixing in a PWR During a Boron Dilution Transient

    SciTech Connect

    Lycklama, Jan-Aiso; Hoehne, Thomas

    2006-07-01

    In a Pressurized Water Reactor, negative reactivity is present in the core by means of Boric acid as a soluble neutron absorber in the coolant water. During a so-called Boron Dilution Transient (BDT), a de-borated slug of coolant water is transported from the cold leg into the reactor vessel, and the borated coolant water is diluted by mixing with this un-borated water. The resulting decrease in the boron concentration leads to an insertion of positive reactivity in the core, which may lead to a reactivity excursion. The associated power peak may damage the fuel rods. The mixing of borated and un-borated water in downcomer and lower plenum is an important process, because it mitigates the degree of reactivity insertion. In the present study the application of Computational Fluid Dynamics (CFD) for the prediction of this mixing of un-borated with borated water in the RPV has been assessed. The analyses have been compared with the measurement data from the Rossendorf coolant mixing model (ROCOM) experiment. The ROCOM test facility represents the primary cooling system of a KONVOI type of PWR (1300 MW{sub el}). In spite of the complicated spatial, temporal, and geometrical aspects of the flow in the RPV, the agreement between the calculated and the experimental data is good. The CFD model tends to slightly under predict the degree of mixing in the RPV resulting in a slight under-prediction of the boron concentration at the core. (authors)

  19. Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

    NASA Astrophysics Data System (ADS)

    Riglet-Martial, Ch.; Sercombe, J.; Lamontagne, J.; Noirot, J.; Roure, I.; Blay, T.; Desgranges, L.

    2016-11-01

    The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that irradiated fuels (of nominal stoichiometry close to 2.000) under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U4O9 type in its cold section, of lower temperature. The parameters governing the oxidation states of UO2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

  20. Reviewer acknowledgement

    PubMed Central

    2014-01-01

    Contributing reviewers A peer-reviewed journal would not survive without the generous time and insightful comments of the reviewers whose efforts often go unrecognized. AIDS Research and Therapy is very grateful for the support of highly qualified peer reviewers and would like to show its appreciation by thanking the following for their assistance with review of manuscripts for the journal in 2013. PMID:24495401