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Sample records for review pwr bwr

  1. 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

  2. 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.

  3. Containment integrity of SEP plants under combined loads. [PWR; BWR

    SciTech Connect

    Lo, T.; Nelson, T.A.; Chen, P.Y.; Persinko, D.; Grimes, C.

    1984-06-01

    Because the containment structure is the last barrier against the release of radioactivity, an assessment was undertaken to identify the design weaknesses and estimate the margins of safety for the SEP containments under the postulated, combined loading conditions of a safe shutdown earthquake (SSE) and a design basis accident (DBA). The design basis accident is either a loss-of-coolant accident (LOCA) or a main steam line break (MSLB). The containment designs analyzed consisted of three inverted light-bulb shaped drywells used in boiling water reactor (BWR) systems, and three steel-lined concrete containments and a spherical steel shell used in pressurized water reactor (PWR) systems. These designs cover a majority of the containment types used in domestic operating plants. The results indicate that five of the seven designs are adequate even under current design standards. For the remaining two designs, the possible design weaknesses identified were buckling of the spherical steel shell and over-stress in both the radial and tangential directions in one of the concrete containments near its base.

  4. 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.

  5. 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.

  6. Nodal analysis for reactor kinetics and stability. [PWR; BWR

    SciTech Connect

    Park, J.K.; Becker, M.; Park, G.C.

    1983-07-01

    General space kinetics models have been developed for more accurate stability analysis utilizing nodal analysis, a commonly used technique for analyzing power distributions in large power reactors. Kinetics parameters for use in these kinetics models have been properly derived by utilizing self-consistent nodal data and power distributions. The procedure employed in the nodal code SIMULATE has been utilized for power distribution, since that methodology is general and includes various commonly used nodal methods as special cases. Cross sections are correlated as functions of void fraction and exposure. A computer program investigating thermo-hydrodynamic stability, NUFREQ has been modified to accommodate general spatial kinetics models with an improved thermal-hydraulics model. Stability analyses have been performed for density wave oscillations for a representative operating BWR system. Spatial coupling effects on the stability margins were found to be significant.

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

    Energy Science and Technology Software Center (ESTSC)

    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

  9. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR & BWR accident sequences

    SciTech Connect

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-08-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences.

  10. 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.

  11. 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.

  12. 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.

  13. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    SciTech Connect

    McCardell, R.K.; MacDonald, P.E.

    1980-01-01

    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  14. Effect of high burn-up and MOX fuel on reprocessing, vitrification and disposal of PWR and BWR spent fuels based on accurate burn-up calculation

    SciTech Connect

    Yoshikawa, T.; Iwasaki, T.; Wada, K.; Suyama, K.

    2006-07-01

    To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)

  15. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    SciTech Connect

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  16. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    SciTech Connect

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-07-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

  17. 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

  18. 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.

  19. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    SciTech Connect

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  20. Workshop on data-acquisition and -display systems: directions after TMI. [PWR; BWR

    SciTech Connect

    Not Available

    1980-11-01

    The accident at Three Mile Island Unit-2 raised questions as to the adequacy of data acquisition and display systems in commercial nuclear power plants. A series of recommendations have developed from the various groups that have analyzed the accident in order to improve the oprator's overview of the plant safety conditions and to facilitate information transfer to technical support centers in emergency situations. This report is the result of an NSAC-sponsored workshop, where the various recommendations and emerging regulatory requirements were reviewed in an attempt to provide an integrated basis for their implementation.

  1. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    SciTech Connect

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. 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. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.

  2. Near-term improvements for nuclear power plant control room annunciator systems. [PWR; BWR

    SciTech Connect

    Rankin, W.L.; Duvernoy, E.G.; Ames, K.R.; Morgenstern, M.H.; Eckenrode, R.J.

    1983-04-01

    This report sets forth a basic design philosophy with its associated functional criteria and design principles for present-day, hard-wired annunciator systems in the control rooms of nuclear power plants. It also presents a variety of annunciator design features that are either necessary for or useful to the implementation of the design philosophy. The information contained in this report is synthesized from an extensive literature review, from inspection and analysis of control room annunciator systems in the nuclear industry and in related industries, and from discussions with a variety of individuals who are knowledgeable about annunciator systems, nuclear plant control rooms, or both. This information should help licensees and license applicants in improving their hard-wired, control room annunciator systems as outlined by NUREG-0700.

  3. Determination of post-DNB and post-BT fuel design limits. [PWR; BWR

    SciTech Connect

    Croucher, D.W.; Loyd, R.J.

    1980-01-01

    Categories of light water reactor transients and the departure from nucleate boiling (DNB) and boiling transition (BT) fuel design limits in light water reactors are reviewed. These fuel design limits for reactor licensing may be overly conservative because experiments have shown that fuel rods do not fail and may not experience damage as a result of momentary operation in film boiling or dryout conditions. Damage to the fuel rod is strongly dependent on the peak cladding temperature and the length of time at that temperature durng the transient. Testing of two potential licensing fuel design limits is suggested: (a) fuel rod functional capabilities are retained and fuel system dimensions remain within operational telerances; and (b) cladding deformation is permitted, but no significant oxidation is allowed. Damage mechanisms which may affect post-DNB or post-BT operation of fuel rods are permanent rod bowing and pellet-cladding interaction. The data necessary to support a fuel design limit and a means of obtaining these data are outlined.

  4. Qualification of active mechanical equipment for nuclear plants. Final report. [PWR; BWR

    SciTech Connect

    Mollerus, F.J.; Allen, R.D.; Gilcrest, J.D.; Rowland, M.C.

    1985-03-01

    This report describes a methodology which can be used for qualifying active mechanical equipment for safety-related service in nuclear power plants in a cost effective manner. The emphasis in this report is to develop guidance on technically sound methods that may be used to qualify mechanical equipment without resorting to piece-by-piece testing. This report identifies the types of mechanical equipment which may need to be qualified and discusses the considerations which are important in the determination of qualification loading conditions. In particular, the traditional practice of seismic qualification of all safety related equipment is reviewed to provide a well-founded technical basis for decoupling environmental loads from seismic (and dynamic) loads. Detailed examination of design, function and potential failure modes of valves, pumps, diesel engines, drive turbines, fans and snubbers provides the technical foundation for the appropriate qualification methods to be used for environmental, seismic and dynamic qualification. These appropriate qualification methods include: design, operating and earthquake experience; analysis and testing; evaluation of the effects of radiation on non-metallic parts, consideration of thermal and wear aging; surveillance and maintenance; and proper documentation.

  5. 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.

  6. Experimental investigation of the enthalpy and mass flow distribution in 16-rod clusters with BWR-PWR geometries and conditions

    NASA Astrophysics Data System (ADS)

    Herkenrath, H.; Hufschmidt, W.; Jung, U.; Weckermann, F.

    Enthalpy and mass flow distribution at the outlet of two test sections with uniform heating in axial and radial direction under steady state conditions was measured by simultaneous sampling of five of six characteristic subchannels in the bundle, using the isokinetic technique and analyzing the outlet quantities by a calorimetric method. Results show low steam quality for the corner subchannel under BWR conditions, due to a thick liquid film on the unheated channel wall. Experimental data confirm the usefullness of the subchannel sampling technique for understanding thermohydraulic phenomena under two-phase flow conditions in multirod bundles. Subchannel resistance coefficients for both types of spacers under one-phase flow conditions were calculated by a substructure method, showing a high local value of the corner subchannel. Total resistance of the spacer was evaluated using local drag coefficients. It agrees well with measured values under adiabatic conditions.

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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).

  15. A review of irradiation assisted stress corrosion cracking

    NASA Astrophysics Data System (ADS)

    Scott, P.

    1994-08-01

    The aim of this review is to assess from the available data whether irradiation in PWR primary water can adversely affect the properties of stainless steels due to irradiation assisted stress corrosion cracking (IASCC). The following aspects are examined: (i) Irradiation damage of the material, (ii) The influence of water radiolysis. Since the irradiation damage processes are similar for both PWR and BWR systems, differences observed in the intergranular cracking properties of core components of both systems must be attributable to differences in the synergistic interactions with the coolant chemistry. These aspects are analysed in detail to determine to what extent BWR experience can be used to predict IASCC in PWR core components. Several related potential failure mechanisms are also reviewed such as radiation hardening, radiation creep and helium or hydrogen embrittlement. The probable role of some or all of these failure mechanisms in core component failures observed to date, and in experiments ostensibly designed to observe IASCC, is critically examined.

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. X-ray measurements of water fog density. [PWR; BWR

    SciTech Connect

    Camp, A.L.

    1982-11-01

    Water-fog densities were measured in a laboratory experiment using x-ray diagnostics. Fog densities were measured, varying the flow rate, nozzle type, nozzle configuration, nozzle height above the x-ray beam, and water surface tension. Suspended water volume fractions between 0.0008 and 0.0074 percent were measured. The fog density increases approximately as the square root of the flow rate; the other parameters had little effect on the density.

  3. Nuclear power plant fire protection: philosophy and analysis. [PWR; BWR

    SciTech Connect

    Berry, D. L.

    1980-05-01

    This report combines a fire severity analysis technique with a fault tree methodology for assessing the importance to nuclear power plant safety of certain combinations of components and systems. Characteristics unique to fire, such as propagation induced by the failure of barriers, have been incorporated into the methodology. By applying the resulting fire analysis technique to actual conditions found in a representative nuclear power plant, it is found that some safety and nonsafety areas are both highly vulnerable to fire spread and impotant to overall safety, while other areas prove to be of marginal importance. Suggestions are made for further experimental and analytical work to supplement the fire analysis method.

  4. Fuel axial relocation in ballooning fuel rods. [PWR; BWR

    SciTech Connect

    Siefken, L.J.

    1983-01-01

    Fuel movement, in the longitudinal direction in ballooning fuel rods, shifts the position of heat generation and may cause an increase in cladding temperature in the ballooning region. This paper summarizes the axial fuel relocation data obtained in fuel rod tests conducted in the United States and the Federal Republic of Germany, describes a model for calculating fuel axial relocation, and gives a quantitative analysis of the impact of fuel relocation on cladding temperature. The amount of fuel relocation in 18 ballooned fuel rods was determined from neutron radiographs, niobium gamma decay counts, and photomicrographs. The fuel rods had burnups in the range of 0 to 35,000 MWd/t and cladding hoop strains varying from 0 to 72%.

  5. 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.

  6. Overview of the PBF test results. [PWR; BWR

    SciTech Connect

    Zeile, H.J.

    1980-01-01

    The Thermal Fuels Behavior Program (TFBP) of EG and G Idaho conducts fuel behavior research in the Power Burst Facility (PBF) at INEL and at the Halden Reactor in Norway. The fuels behavior research in the PBF is directed toward providing a detailed understanding of the response of light water reactor (LWR) nuclear fuel assemblies to off-normal and hypothesized accident conditions. Single fuel rods and clusters of highly instrumented fuel rods are installed within a central test space of the PBF core for testing. The core can be operated in various modes to provide test conditions typical of accidents and off-normal conditions that may be experienced in a pressurized water reactor or a boiling water reactor.

  7. Detonation of hydrogen-air mixtures. [PWR; BWR

    SciTech Connect

    Lee, J.H.S.; Knystautas, R.; Benedick, W.B.

    1983-01-01

    The detonation of a hydrogen-air cloud subsequent to an accidental release of hydrogen into ambient surroundings cannot be totally ruled out in view of the relative sensitivity of the hydrogen-air system. The present paper investigates the key parameters involved in hydrogen-air detonations and attempts to establish quantitative correlations between those that have important practical implications. Thus, for example, the characteristic length scale lambda describing the cellular structure of a detonation front is measured for a broad range of hydrogen-air mixtures and is quantitatively correlated with the key dynamic detonation properties such as detonability, transmission and initiation.

  8. 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.

  9. 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.

  10. Evaluation of internal flooding in a BWR

    SciTech Connect

    Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

    1985-01-01

    Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event.

  11. 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.

  12. High-fidelity multiphysics simulation of BWR assembly with coupled TORT-TD/CTF

    SciTech Connect

    Magedanz, J.; Perin, Y.; Avramova, M.; Pautz, A.; Puente-Espel, F.; Seubert, A.; Sureda, A.; Velkov, K.; Zwermann, W.

    2012-07-01

    This paper describes the application of the coupled codes TORT-TD and CTF to the pin-by-pin modeling of a BWR fuel assembly with thermal-hydraulic feedback. TORT-TD, developed at GRS, is a time-dependent three dimensional discrete ordinates code. CTF is the PSU's improved version of the subchannel code COBRA-TF, which uses a two-fluid, three-field model to represent two-phase flow with entrained droplets, and is commonly applied to evaluate LWR safety margins. The coupled codes system TORT-TD/CTF, already applied to several PWR cases involving MOX, was adapted from PWR to BWR applications. The purpose of the research described in this paper is to verify the coupling for modeling two-phase flow at the pin cell level. Using an ATRIUM-10 assembly, the system's steady-state capabilities were tested on two cases: one without control blade insertion and another with partially inserted blades. The influence of the neutron absorber on local axial and radial parameters is presented. The description of an inlet flow reduction transient is an example for the time-dependent capability of the coupled system. (authors)

  13. BWR nondestructive evaluation source book

    SciTech Connect

    Neeley, V.I. ); Walker, S.M. ); Behravesh, M.M. )

    1990-06-01

    In response to increasing incidences of IGSCC in BWRs in the last decade, EPRI and the BWR Owners Group jointly developed and successfully implemented numerous projects relating to the use of NDE techniques to detect and size IGSCC. This report provides a brief summary of these various projects along with their historical perpectives in a single, condensed source. The various IGSCC problems of BWRs and the projects that were conducted to address them are more thoroughly described in many other EPRI reports referenced herein. 97 refs., 50 figs.

  14. (Boiling water reactor (BWR) CORA experiments)

    SciTech Connect

    Ott, L.J.

    1990-10-16

    To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.

  15. BWR core melt progression phenomena: Experimental analyses

    SciTech Connect

    Ott, L.J.

    1992-06-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component.

  16. BWR core melt progression phenomena: Experimental analyses

    SciTech Connect

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component.

  17. 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

  18. 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.

  19. 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).

  20. 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

  1. Probability of failure in BWR (Boiling Water Reactor) reactor coolant piping: Volume 2, Pipe failure induced by crack growth and failure of intermediate supports

    SciTech Connect

    Lo, T.; Bumpus, S.E.; Chinn, D.J.; Mensing, R.W.; Holman, G.S.

    1989-03-01

    The US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to conduct a study to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in the major coolant piping systems of nuclear power plants is large enough to warrant the current stringent design requirements of designing against the postulated effects of a DEGB. The study includes both the PWR (Pressurized Water Reactor) and the BWR (Boiling Water Reactor) plants in the United States. Following the study of PWR plants, a study of BWR reactor coolant piping was performed. The Brunswick Steam Electric Plant at Southport, North Carolina was selected as the pilot plant for the BWR evaluation. The probability of pipe failure in three major coolant pipings was assessed: the recirculation loops, the primary steam lines, and the main feedwater lines. In the case of recirculation loops, both the existing and a proposed replacement system were studied. A probabilistic fracture mechanics approach was used in this study to estimate the crack growth and to assess the crack stability in the piping systems throughout the lifetime of the plant. The effects of the failure of intermediate pipe supports were also examined. The results of the assessment indicated that the probability of occurrence of DEGB due to crack growth and instability is small if the problem of intergranular stress corrosion cracking (IGSCC) is resolved by the use of the replacement system. 33 refs., 41 figs., 32 tabs.

  2. 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.

  3. 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).

  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. 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.

  6. 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.

  7. Computations of the fluid annulus surrounding a pump shaft in synchronous vibration. Final report. [PWR; BWR

    SciTech Connect

    Cohen, L.M.; Gross, M.B.; Yeung, W.

    1983-07-01

    The purpose of this study is to analyze the fluid forces that arise from the fluid annulus surrounding a pump shaft in synchronous vibration. The computational method for determining the fluid forces is based on the two-dimensional, incompressible implicit version of the STEALTH 2D code. Using this computational tool, this study analyzes pump-shaft synchronous vibrations both in the low-Reynolds-number laminar-flow regime and the high-Reynolds-number turbulent-flow regime. Moreover, the study was able to investigate both infinitesimal (perturbation) shaft displacements as well as finite (substantial) shaft displacements. The new numerical results predict forces from the fluid annulus which are substantially greater than the forces computed by extrapolating a perturbation analysis to finite displacements. The new forces are substantially greater because the two-dimensional calculations are much stiffer than the perturbation analysis, which neglects the radial-velocity component of the flow. The numerical results for the fluid forces are summarized by simple analytic formulas.

  8. Transient fuel-pin temperature calculations using describing functions. Final report. [PWR; BWR

    SciTech Connect

    Guidotti, T.E.; Peddicord, K.L.; Nielsen, L.H.

    1982-02-01

    Temperature dependent properties make the transient thermal analysis of reactor fuel pins nonlinear. Traditionally, finite difference or finite element methods have been used to solve this system. An analytically based method is presented in which the transient temperature profiles are calculated using describing functions to characterize the system nonlinearities. In the past the describing functions approach has been used exclusively in the solution of ordinary differential equation systems. This study appears to be the first application of describing functions in the solution of partial differential equations.

  9. 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.

  10. Fuel Performance Improvement Program. Semiannual progress report, October 1979-March 1980. [PWR; BWR

    SciTech Connect

    Not Available

    1980-04-01

    Progress on the Fuel Performance Improvement Program's fuel design tests and demonstration irradiations for October 1979 through March 1980 is reported. Included are the results of out-of-reactor experiments with Zircaloy cladding using a device that simulates the interaction between fuel and cladding. Also included are reports on the irradiation of the advanced LWR fuel designs in the Halden Boiling Water Reactor and in Consumers Power Company's Big Rock Point Reactor. The establishment of the technical bases and licensing requirements for the advanced fuel concepts are also described.

  11. 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.

  12. Fission product source term research at Oak Ridge National Laboratory. [PWR; BWR

    SciTech Connect

    Malinauskas, A.P.

    1985-01-01

    The purpose of this work is to describe some of the research being performed at ORNL in support of the effort to describe, as realistically as possible, fission product source terms for nuclear reactor accidents. In order to make this presentation manageable, only those studies directly concerned with fission product behavior, as opposed to thermal hydraulics, accident sequence progression, etc., will be discussed.

  13. Aerosol release and transport program. Quarterly progress report, October-December 1981. [LMFBR; PWR; BWR

    SciTech Connect

    Adams, R. E.; Tobias, M. L.

    1982-05-01

    This report summarizes progress for the Aerosol Release and Transport Program sponsored by the Nuclear Regulatory Commission's Office of Nuclear Regulatory Research, Division of Accident Evaluation, for the period October-December 1981. Topics discussed include (1) under-sodium tests in the Fuel Aerosol Simulant Test (FAST) Facility, (2) U/sub 3/O/sub 8/ and Fe/sub 2/O/sub 3/ in steam (light-water reactor accident) aerosol experiments in the Nuclear Safety Pilot Plant, (3) generation and characterization of cadmium and CdO aerosols in the basic aerosol experimental program, (4) core-melt tests of Zircaloy-clad fuel capsules, (5) initial results of a piston-model bubble oscillation code allowing liquid bypass, and (6) calculations with the UVABUBL code to compare with underwater and under-sodium period measurements in FAST experiments.

  14. Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

    DOEpatents

    Tokarz, R.D.

    1981-10-27

    This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

  15. Portable monitor for the determination of reactor coolant surface radionuclide activities. [PWR; BWR

    SciTech Connect

    Harvill, J.P.; Corbett, J.; Bergmann, C.A.

    1982-07-01

    The determination of primary surface deposited radionuclide activities has, in the past, been severely limited by physical structures and equipment size. In order to simplify the acquisition of this in-plant data, a portable gamma spectroscopy system has been assembled. This system consists of a planar high purity germanium (HpGe) detector mounted in a Multi-use Portable Spectrometer (MPS) cryostat purchased from Princeton Gamma-Tech and an ND-Six portable multichannel analyzer (MCA) purchased from Nuclear Data, Inc. Weighing less than 30 pounds, the entire system is completely battery powered and allows storage of accumulated spectra on mini-cassettes. Shielding and collimation is provided by a tungsten-based alloy obtained from Teledyne Power Alloys. The analytical methodology necessary for determining pipewall surface activities and manway insert activities has also been developed.

  16. 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.

  17. 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.

  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. BNL severe-accident sequence experiments and analysis program. [PWR; BWR

    SciTech Connect

    Greene, G.A.; Ginsberg, T.; Tutu, N.K.

    1983-01-01

    In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described.

  20. Generating human reliability estimates using expert judgment. Volume 2. Appendices. [PWR; BWR

    SciTech Connect

    Comer, M.K.; Seaver, D.A.; Stillwell, W.G.; Gaddy, C.D.

    1984-11-01

    The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessments (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 2 provides detailed procedures for using the techniques, detailed descriptions of the analyses performed to evaluate the techniques, and HEP estimates generated as part of this project. The results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. Judgments were shown to be consistent and to provide HEP estimates with a good degree of convergent validity. Of the two techniques tested, direct numerical estimation appears to be preferable in terms of ease of application and quality of results.

  1. 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.

  2. Spinning-turbine-fragment impacts on casing models. Final report. [PWR; BWR

    SciTech Connect

    Wilbeck, J.S.

    1983-07-01

    Ten 1/11-scale model turbine-missile-impact tests were conducted at the Naval Air Propulsion Center under the supervision of Southwest Research Institute. These tests were conducted in support of the EPRI program to assess turbine-missile effects in nuclear plant design. The objective of the tests was to determine the effects of missile spin, blade crush, and target edge conditions on the impact of turbine disk fragments on the steel casing. The burst of a modified gas-turbine rotor in a high-speed spin chamber provided three missiles with the proper rotational and translational velocities of actual steam-turbine fragments. Tests of unbladed, spinning missiles were compared with previous tests of unbladed, nonspinning missiles. The total residual energy of the spinning missiles was the same as that of the nonspinning missiles launched in a piercing orientation. Tests with bladed missiles showed that for equal burst speeds, the residual energy of bladed missiles is less than that of unbladed missiles.

  3. Evaluation of flaw characteristics and their influence on inservice inspection reliability. [PWR; BWR

    SciTech Connect

    Becker, F.L.

    1980-01-01

    This report describes the results of the first year's effort of a five year program which is being conducted by Battelle, Pacific Northwest Laboratories, on behalf of the US Nuclear Regulatory Commission. This initial effort was directed toward identification and quantification of inspection uncertainties, which are likely to occur during inservice inspection of LWR primary piping systems, and their influence on inspection reliability. These experiments were conducted on 304 stainless steel samples, however, the results are equally applicable to other materials. Later portions of the program will extend these measurements and evaluations to other materials and conditions.

  4. Evaluation of nonequilibrium effects in bundle dispersed-flow film boiling. [PWR; BWR

    SciTech Connect

    Morris, D.G.; Mullins, C.B.; Yoder, G.L.

    1983-01-01

    The effects of thermodynamic nonequilibrium in dispersed flow film boiling heat transfer are examined. Steady-state and transient rod-bundle data are used to evaluate several empirical heat-transfer models commonly employed to predict post-CHF behavior. The models that account for thermodynamic nonequilibrium perform adequately, while those that ignore nonequilibrium effects incur errors in wall superheat as high as 190/sup 0/K. Nonequilibrium effects can also be treated by explicitly modeling the phenomena. The thermal-hydraulic code COBRA-TF employs this approach. Using bundle data, the models in the code are evaluated. Analysis suggests that the interfacial heat transfer is overpredicted.

  5. Heat-transfer characteristics of dry porous particulate beds with internal heat generation. [PWR; BWR

    SciTech Connect

    Kelly, J. E.; Hitchcock, J. T.; Schwarz, M. L.

    1982-08-01

    A combined experimental and analytical program has been undertaken to study the thermal characteristics of dry debris beds especially at high temperatures where radiative heat transfer contributes. Experiments have been conducted in-pile using intrinsic fission heating of UO/sub 2/ to simulate decay heat power levels. Both pure UO/sub 2/ and mixed UO/sub 2//steel beds in a helium atmosphere have been studied. Temperatures as high as 3100 K have been attained. Theremocouples and ultrasonic thermometers have been used to measure the bed temperatures. Also recent post-test metallographic examinations have revealed useful information on the materials interactions which can be correlated to changes in the thermal behavior of the bed. In addition to the experiments, a number of porous medium thermal conductivity models have been evaluated. From these, two models representing upper and lower bounds for the conductivity, in conjunction with a two-dimensional heat transfer analysis have been compared to the data.

  6. 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.

  7. Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors. [PWR; BWR

    SciTech Connect

    Chambers, R.; Laats, E.T.

    1981-01-01

    A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power history, fuel and cladding behavior models, and materials properties correlations. Three of the nine EMs provide either input or model specifications, or set the reference temperature for stored energy calculations. The remaining six EMs were intended to add thermal conservatism through model changes. To determine the relative influence of these six EMs upon fuel behavior calculations for commercial power reactors, a sensitivity study was conducted. That study is the subject of this paper.

  8. 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.

  9. Design report for JAERI Slab Core hot leg spool piece. [PWR; BWR

    SciTech Connect

    Rahl, T.E.; Tatar, G.A.

    1980-08-01

    This document represents the Design Report for the JAERI Slab Core Hot Leg Spool Piece. This report, in accordance with Section VIII of the ASME Boiler and Pressure Vessel Code, has been prepared in sufficient detail to demonstrate that the stress and fatigue limits of the Code are satisfied when the spool piece is subjected to the Design and Service Conditions specified in the Design Specification.

  10. Experimental study of debris-bed coolability under pool-boiling conditions. [PWR; BWR; LMFBR

    SciTech Connect

    Catton, I.; Dhir, V.K.; Somerton, C.W.

    1983-05-01

    An experimental investigation has been conducted into the dryout of a bed of inductively heated particles cooled by an overlying liquid pool. Particles of diameters 4763 ..mu..m, 3175 ..mu..m, 1588 ..mu..m, and 589-787 ..mu..m have been used. Acetone and water have been used as the coolant with bed heights varying from 5 to 40 cm. Results are presented in terms of the dryout heat as a function of bed height. It has been found that the ratio of the overlying liquid pool height to the particulate bed height can influence the dryout heat flux. Comparison with other experimetal studies was good and a comparison with proposed theoretical models was also made.

  11. Vessel V-8 repair and preparation of low upper-shelf weldment. [PWR; BWR

    SciTech Connect

    Domian, H.A.

    1982-06-01

    This report describes the weld repair of the Heavy Section Steel Technology (HSST) Program Intermediate Test Vessel V-8 and the making of a submerged-arc low upper-shelf test weld as a long seam in the vessel. The development of the low upper-shelf weld metal and its mechanical property characterization, including fracture toughness, is reported.

  12. 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.

  13. 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.

  14. Core-concrete molten pool dynamics and interfacial heat transfer. [PWR; BWR

    SciTech Connect

    Benjamin, A.S.

    1980-01-01

    Theoretical models are derived for the heat transfer from molten oxide pools to an underlying concrete surface and from molten steel pools to a general concrete containment. To accomplish this, two separate effects models are first developed, one emphasizing the vigorous agitation of the molten pool by gases evolving from the concrete and the other considering the insulating effect of a slag layer produced by concrete melting. The resulting algebraic expressions, combined into a general core-concrete heat transfer representation, are shown to provide very good agreement with experiments involving molten steel pours into concrete crucibles.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Process inherent ultimate safety/boiling-water reactor PIUS/BWR

    SciTech Connect

    Forsberg, C.W.

    1985-01-01

    This document is a series of viewgraphs on: design basis of PIUS/BWR, definition of PIUS/BWR, mechanisms of safe shutdown and afterheat cooling, advantages of PIUS/BWR, and research and development requirements. (DLC)

  1. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    SciTech Connect

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

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation.

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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.

  7. 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.

  8. 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.

  9. PWR systems transient analysis: a reactor-safety perspective

    SciTech Connect

    Kennedy, M.F.; Abramson, P.B.; McDonald, T.A.

    1982-01-01

    In the simulation of transient events in large PWR reactor systems for reactor safety studies, the plant model is quite detailed and must include most of the plant components and control systems to adequately analyze the range of transients. The results discussed were calculated with the RELAP4/MOD6 code and reveal the need for the analysis to carefully review and understand the results to assure that they are not being adversely affected by the improper solution techniques or changes in models during the calculation.

  10. BWR Core Heat Transfer Code System.

    Energy Science and Technology Software Center (ESTSC)

    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

  11. BWR reactor vessel bottom head failure modes

    SciTech Connect

    Hodge, S.A.

    1989-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. The effect of the BWR procedural and structural differences upon the progression of a severe accident sequence during the period preceding movement of core debris into the reactor vessel lower plenum has been discussed previously. It is the purpose of this paper to briefly address the events occurring after debris relocation past the core plate and to describe the subsequent expected modes of bottom head pressure boundary failure. As an example, the calculated timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom Atomic Power Station is also presented. 14 refs., 4 figs., 1 tab.

  12. 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)

  13. 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.

  14. Impact of B{sub 4}C on the stratification of molten steel and BWR-type core material under IVR conditions

    SciTech Connect

    Fischer, M.; Sulatsky, A. A.; Krushinov, E. V.

    2012-07-01

    As part of the validation of the In-Vessel Melt Retention (IVR) strategy for its KERENA BWR, AREVA NP has performed a quantitative assessment of the potential impact of thermochemical phenomena. This was motivated by the fact that several of these phenomena, namely the formation of a dense metallic phase, have the potential to lead to a strong increase in local heat fluxes, with the risk of early IVR failure. In this context, experiments performed in the MASCA project with PWR-type corium melts were repeated using a typical BWR core melt: characterized by a lower U/Zr-ratio and higher contents of Zr, steel, and boron carbide (B{sub 4}C). Applying an improved 'cold crucible' induction heating technique, two series of test were performed, first without B{sub 4}C and then with a B{sub 4}C content of about 1.4 wt%. These two values bound the uncertainty with respect to the incorporation of B{sub 4}C into the molten pool. The target of the tests was to localize the point of equal densities between the dense metallic phase (Fe, U, Zr and O) and the residual oxidic phase in thermo-chemical equilibrium. An interesting result of these experiments was that, different from earlier MASCA tests with PWR-type corium that showed an only insignificant impact of B{sub 4}C on metal density, the new experiments reveal a strong corresponding effect, which can over-compensates the density increase caused by U-migration into the metallic melt. This deviating result is attributed, first, to the higher Zr-fraction in the BWR-type core melt, and second, to the higher content of B{sub 4}C in the BWR-type melt in comparison to the MASCA tests (B{sub 4}C-content <0.5 wt%). Based on the obtained results it is predicted that - under certain conditions -B{sub 4}C can completely prevent the formation of a dense metallic phase, independent of the amount of molten steel in the melt. The paper gives an overview of the performed experiments and their main results and provides theoretical models to

  15. 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

  16. 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

  17. 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

  18. 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

  19. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    SciTech Connect

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  20. 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.

  1. (Severe accident technology of BWR (Boiling Water Reactor) reactors)

    SciTech Connect

    Ott, L.J.

    1989-10-23

    The traveler attended the 1989 CORA Workshop at KfK, FRG. Participation included the presentation included the presentation of three papers on work performed by the Boiling Water Reactor Severe Accident Technology (BWRSAT) program at Oak Ridge National Laboratory (ORNL) in Boiling Water Reactor (BWR) severe accident analyses. The Statement of Work (June 1989) for the BWRSAT Program provides for code analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that BWRSAT personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to arrange for acquisition of detailed CORA facility drawings, experimental data, and related engineering. 17 refs.

  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. Interpretation of the results of the CORA-33 dry core BWR test

    SciTech Connect

    Ott, L.J.; Hagen, S.

    1993-11-01

    All BWR degraded core experiments performed prior to CORA-33 were conducted under ``wet`` core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ``dry`` core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ``dry`` core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ``dry`` BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ``dry`` core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed.

  4. Impact of Plant Noise on BWR Stability Analyses

    SciTech Connect

    Andersen, J.G.M.; Pearson, G.; Chung, A.K.; Heck, C.L.; Vedovi, J.

    2007-07-01

    A small amount of thermal-hydraulic noise is present in a boiling water reactor (BWR). The noise originates primarily from minor fluctuations in fluid flow and pressure distribution in the recirculation system of the BWR and manifests itself as a small fluctuation on the order of 1-2% for the average power range monitors (APRM) during normal operation. A larger noise level is observed for single-loop operation than for two-loop operation. This noise has an impact on the stability performance of the BWR. This is particularly the case when the noise contains a significant component at the resonant frequency for BWR instabilities, which is typically on the order of 0.5 Hz. For a pump trip event that can lead to instability, the noise will impact the growth rate of the reactor instability. The initial magnitude of the oscillations will be larger as the decay ratio increases above unity. For operation at low flow, such as for minimum pump speed or single loop operation where the decay ratio is larger, the impact of noise could lead to small oscillations at the resonant frequency for the APRM signals. The impact of noise on BWR instabilities is analyzed with the TRACG code. TRACG consists of a multi dimensional two-fluid thermal hydraulics model and the three-dimensional kinetics model consistent with the GE 3D core simulator, PANACEA. TRACG models the reactor primary system and has been extensively qualified against test data and BWR plant data. Thermal hydraulic instability test data, as well as data from BWR instability events and tests, have been used extensively in this qualification. This paper demonstrates the impact of noise on BWR stability response for events leading to instability, such as pump trip events, as well as operation at low core flow due to single loop operation. The impact is illustrated through sensitivity studies with the TRACG code and by comparison to plant data. The impact of reactor noise on the performance of the instability detection system

  5. 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.

  6. 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.

  7. SCCIG: a phenomenological model for iodine stress corrosion cracking of Zircaloy. Volume 1. Final report. [PWR; BWR

    SciTech Connect

    Miller, A.K.; Challenger, K.D.; Tasooji, A.; Ranjan, G.V.; Brooks, G.N.; Eisenberger, M.; Thomas, J.M.

    1981-04-01

    This report presents a new model for iodine stress corrosion cracking (SCC) of Zircaloy, developed as an ingredient of EPRI's SPEAR light-water reactor fuel performance code. The model is based on experimental data generated under related EPRI programs and elsewhere, as well as certain theoretical considerations. It is a unified treatment of SCC crack initiation, crack propagation, and non-SCC fracture, treating such sub-process as intergranular failure, transgranular (cleavage plus fluting) failure, ductile rupture, changes in the local stress and strain distribution due to crack formation and growth, iodine penetration along the SCC crack, and the mutual couplings among these processes.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Seismic fragility testing of naturally-aged, safety-related, class 1E battery cells. [PWR; BWR

    SciTech Connect

    Bonzon, L.L.; Hente, D.B.; Kukreti, B.M.; Schendel, J.S.; Black, D.A.; Paulsen, G.D.; Tulk, J.D.; Janis, W.J.; Aucoin, B.D.

    1984-01-01

    The concern over seismic susceptibility of naturally-aged lead-acid batteries used for safety-related emergency power in nuclear power stations was brought about by battery problems that periodically had been reported in Licensee Event Reports (LERs). The Turkey Point Station had reported cracked and buckled plates in several cells in October 1974 (LER 75-5). The Fitzpatrick Station had reported cracked battery cell cases in October 1977 (LER 77-55) and again in September 1979 (LER 79-59). The Browns Ferry Station had reported a cracked cell leaking a small quantity of electrolyte in July 1981 (LER 81-42). The Indian Point Station had reported cracked and leaking cells in both February (LER 82-7) and April 1982 (LER 82-16); both of these LERs indicated the cracked cells were due to expansion (i.e., growth) of the positive plates.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Mechanistic prediction of fission-product release under normal and accident conditions: key uncertainties that need better resolution. [PWR; BWR

    SciTech Connect

    Rest, J.

    1983-09-01

    A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. 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

  7. 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).

  8. 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).

  9. 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.

  10. 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.

  11. 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

  12. 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)

  13. 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.

  14. A novel MELCOR modeling approach for BWR applications

    SciTech Connect

    Sanders, R.L.

    1997-03-01

    Even though improvements in modeling the phenomena that occur in the lower plenum have been made, the MELCOR user must decide on the physical events that are to be modeled and create the corresponding input. In addition to the standard modeling techniques, the flexibility of MELCOR allows for nontraditional modeling as well. For typical BWR applications, the user establishes concentric radial rings for the COR Package. In theory, each ring should be composed of a channel (within the channel box walls) and a bypass (the volume region outside the channel box walls) with each region containing both fluid and structures. However, in MELCOR 1.8.3 and earlier versions, the COR Package allows particulate debris to relocate only inside the channel box walls, using a volume equivalent to the channel and bypass fluid volumes. The consequences of this modeling approach with respect to the combined channel and bypass volumes produces two anomalies in the BWR calculation, particularly for boildown sequences. The first is that the CVH Package will calculate inaccurate channel and bypass water levels for the core region, which can significantly affect predicted metal-water reaction rates. The other is that lack of separation of the channel and bypass volumes will cause an artificially large flow area through the core plate, which can lead to inaccurately determined core plate failure times and unrealistic pressure transients in the lower plenum. In an effort to mitigate the consequences of the current COR package treatment of BWR channel and bypass volumes, an atypical way of modeling the core region by doubling the number of rings so that each ring represents only a channel or bypass region has been developed and has been shown to be beneficial.

  15. Zebra: An advanced PWR lattice code

    SciTech Connect

    Cao, L.; Wu, H.; Zheng, Y.

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  16. Thermodynamic properties of real gases and BWR equation of state

    NASA Astrophysics Data System (ADS)

    Vestfálová, Magda

    2015-05-01

    The fundamental base for the calculation of the thermodynamic properties of materials is thermal equation of state and dependence of some of the basic specific heat capacities on temperature. The dependence of the specific thermal capacity on the second independent variable (for example on the volume) it is already possible to deduce from the thermal equation of state. The aim of this paper is to assess the compliance values of specific heat capacity which was calculated using the BWR thermal equation of the state and experimentally obtained known values of specific heat capacity for the substance, whose characteristics are available in a wide range of state space.

  17. Thermal equation of state BWR for water and water steam

    NASA Astrophysics Data System (ADS)

    Vestfálová, Magda; Petříková, Markéta

    2016-03-01

    The fundamental bases for the calculation of the thermodynamic properties of substances are the thermal equation of state and the dependence of some of the basic specific heat capacities on temperature. Any other thermodynamic properties of a substance can already be deduced from these relations. The aim of this paper is to design the simple thermal equation of state for water steam based on original BWR thermal equation of state, which describes the behavior of H2O in the gaseous phase in region of usual pressures and temperatures with sufficient accuracy.

  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. Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments

    SciTech Connect

    Ott, L.J.

    1994-01-01

    The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments [Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a

  20. Aging assessment of BWR control rod drive systems

    SciTech Connect

    Greene, R.H.

    1991-01-01

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs.

  1. 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.

  2. BWR lower plenum debris bed models for MELCOR

    SciTech Connect

    Hodge, S.A.; Ott, L.J.

    1991-01-01

    Work is underway at Oak Ridge National Laboratory (ORNL) to incorporate certain models of the Boiling Water Reactor Severe Accident Response (BWRSAR) code into a local version of MELCOR. Specifically, the BWR lower plenum debris bed and bottom head response models taken from BWRSAR are being tested within the local MELCOR code structure. Upon successful completion of testing, recommendations for formal adoption of these models will be made to the Nuclear Regulatory Commission (NRC) and to the MELCOR code development staff at Sandia National Laboratories (SNL). The SNL code development staff retain exclusive responsibility for maintaining the configuration control for the official version of MELCOR. The BWR lower plenum debris bed and bottom head response models permit the calculation of heatup, melting, and relocation of the debris after dryout. They predict the response of the lower plenum internal structures and the bottom head as well as the composition and timing of material release from the vessel. They have been previously applied in severe accident analyses for the Containment Performance Improvement (CPI) Program and the Mark I shell survivability study (NUREG/CR-5423), and in recent assessments of candidate accident management strategies. This paper provides a brief description of the purpose and operation of these models. 11 refs., 15 figs., 5 tabs.

  3. Experimental evaluation of cobalt behavior on BWR fuel rod surface

    SciTech Connect

    Karasawa, H.; Asakura, Y.; Sakagami, M.; Uchida, S. )

    1988-06-01

    Cobalt behavior on the boiling water reactor (BWR) fuel rod surface was experimentally evaluated at 285 C and with various pH values. Adsorption of cobalt ions on hematite particles proceeded via the exchange reaction of cobalt ion with the surface hydroxyl of the hematite. The equilibrium constant for the adsorption at 285 C was found to be -- 570 times as large as that at 20 C. The adsorbate formed cobalt ferrite at the rate of 3.4 x 10/sup -2/ g-Co/g-Co adsorbed/h. The dissolution rates of cobalt ferrite and cobalt oxide particles were found to depend on (H/sup -/)/sup 1.1/ and (H/sup -/)/sup 1.2/, respectively, where (H/sup -/) means the H/sup -/ concentration. Cobalt ions were released from these oxides when O/sup 2-/ ions in them combined with two aqueous protons to form water at the oxide-water interface. Cobalt behavior on the fuel rod surface under BWR conditions was discussed using the experimental results.

  4. Comparative analysis of isotopic composition of spent fuel from Takahama-3 PWR PIE database using TRIPOLI-PEPIN code

    SciTech Connect

    Lee, Y. K.

    2006-07-01

    Evaluation of isotopic composition of spent nuclear fuel is essential for reactor physics and fuel cycle back-end applications. A TRIPOLI-PEPIN coupled depletion code, TR4PEP, has been developed to meet these requirements. It combines the continuous-energy Monte Carlo transport code, TRIPOLI4.3 [1] and the point depletion code, PEPIN-2 [2], to perform the burnup dependent material data calculation. The depletion calculation flow of TR4PEP code has been presented on a previous study. Its application on PWR UO{sub 2} and MOX spent fuel has been validated against several international numerical benchmarks. Compared to industry standard deterministic cell codes and other Monte Carlo based depletion codes, TR4PEP deep-burn depletion calculations have shown satisfactory results. [3] In addition to the numerical benchmarks, the analysis of available post irradiation examination (PIE) results by TR4PEP is also important The PIE results at fuel assembly level are accessible only from spent fuel reprocessing plant and these data are not easy to use for code validation due to the dissolution of several assemblies in the same time. The PIE results at fuel pellet level depend not only on the method for the isotopic measurements but also on the irradiation environment and history. A free access PIE database on isotopic composition of spent nuclear fuel is obtainable from OECD/NEA. [4] Both PWR and BWR PIE data at fuel pellet level are taken into account in this database but the only 17 x 17 type PWR fuel available in this database is from Takahama-3 PIE results. To validate TR4PEP with Takahama-3 PIE results, two irradiated UO{sub 2} samples, SF95-4 from fuel assembly NT3G23 and SF97-5 from NT3G24, are considered in this study. Both samples have an initial {sup 235}U enrichment of 4.11 wt% and their burnup are respectively 36.69 and 47.03 GWd/t. Comparative analysis of isotopic composition from SF95-4 and SF97-5 including 19 actinides from {sup 234}U to {sup 247}Cm and 18

  5. 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.

  6. 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)

  7. Influence Of Low Boron Core Design On PWR Transient Behavior

    SciTech Connect

    Aleksandrov Papukchiev, Angel; Yubo Liu; Schaefer, Anselm

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, the concentration of boron in primary coolant is limited by the requirement of having a negative moderator density coefficient. As high boron concentrations have significant impact on reactivity feedback properties, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) content has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) by approx. 50% compared to current German PWR technology. For the assessment of the potential safety advantages, a Loss-of-Feedwater Anticipated Transient Without Scram (ATWS LOFW) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The most significant difference in the transient performance of both designs is the total primary fluid mass released through the pressurizer (PRZ) valves. It is reduced by a factor of four for the low boron reactor, indicating its improved density reactivity feedback. (authors)

  8. 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.

  9. 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.

  10. Applicability of BWR SFD experiments and codes for advanced core component designs

    SciTech Connect

    Ott, L.J.

    1997-12-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories (SNL) in 1986, no experimental database existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper presents the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core on of SFD code.

  11. 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.

  12. 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.

  13. Sensitivity of risk parameters to human errors for a PWR

    SciTech Connect

    Samanta, P.; Hall, R. E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study.

  14. 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.

  15. 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.

  16. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect

    Wang, S.-J.; Chiang, K.-S.; Chiang, S.-C.

    2004-05-15

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  17. Method of characteristics - Based sensitivity calculations for international PWR benchmark

    SciTech Connect

    Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

    2013-07-01

    Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

  18. Comparison of Removed Fuel Compositions of CANDLE, PWR, and FBR

    SciTech Connect

    Nagata, Akito; Sekimoto, Hiroshi

    2007-07-01

    A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replaced fresh fuels. About 40% of natural or depleted uranium undergoes fission. In this paper, spent fuels of PWR, FBR and CANDLE reactor are compared. Fresh fuels of PWR, FBR and CANDLE reactor are 4.1% enriched uranium (UO{sub 2}), MOX with 18.5% plutonium enrichment and natural uranium nitride (natural-UN), respectively. In once-through fuel cycle point of view, low disposal amount for high energy is better and CANDLE reactor can decrease this amount more than other reactors, especially it is only one-tenth of PWR fuel. Also, it can decrease MA and this amount is 0.4 times of PWR. Total FP amount of each reactor is nearly same. However, LLFP amount of CANDLE reactor is the largest. (authors)

  19. 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.

  20. 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

  1. 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.

  2. 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)

  3. 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).

  4. A New Methodology for Early Anomaly Detection of BWR Instabilities

    SciTech Connect

    Ivanov, K. N.

    2005-11-27

    The objective of the performed research is to develop an early anomaly detection methodology so as to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The research utilizes a model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, which is used as a generator of time series data for anomaly detection at an early stage. The model captures critical nonlinear features of coupled thermal-hydraulic and nuclear reactor dynamics and (slow time-scale) evolution of the anomalies as non-stationary parameters. The time series data derived from this nonlinear non-stationary model serves as the source of information for generating the symbolic dynamics for characterization of model parameter changes that quantitatively represent small anomalies. The major focus of the presented research activity was on developing and qualifying algorithms of pattern recognition for power instability based on anomaly detection from time series data, which later can be used to formulate real-time decision and control algorithms for suppression of power oscillations for a variety of anticipated operating conditions. The research being performed in the framework of this project is essential to make significant improvement in the capability of thermal instability analyses for enhancing safety, availability, and operational flexibility of currently operating and next generation BWRs.

  5. BWR pipe crack and weld clad overlay studies

    SciTech Connect

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

    1985-10-01

    This paper presents results on (a) the influence of simulated BWR environments and temperature on the intergranular-stress-corrosion cracking (IGSCC) susceptibility of sensitized stainless steels (SS), (b) the stress-corrosion susceptibility of alternative piping materials, (c) analysis of field components to assess the effectiveness of in-service inspection techniques and the in-reactor performance of weld overlay repairs, and (d) finite-element analyses and experimental measurement of residual stresses in weldments with weld overlays. Fracture-mechanics crack-growth data are presented to confirm correlations between the critical corrosion potentials required to inhibit IGSCC and the level of impurities in the environment. Slow-strain-rate tests show that very low levels of impurities (25 ppb of sulfate) can produce suseptibility to transgranular-stress-corrosion cracking (TGSCC) in Type 316NG SS and that nitrogen levels is SS above 0.1 wt. % appear to increase susceptibility to TGSCC. Preliminary results on a German Type 347NG SS suggest that it is at least as resistant to TGSCC in impurity environments as Type 316NG SS. Measurements on overlay weldments removed from the Hatch-2 reactor confirm that compressive residual stresses are produced on the inner surface of the weldments by the overlay.

  6. 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

  7. 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)

  8. 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.

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

    Energy Science and Technology Software Center (ESTSC)

    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

  10. 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.

  11. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    SciTech Connect

    Edwards, Robert; Huang, Zhengyu

    2001-06-17

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown.

  12. 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)

  13. 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.

  14. Accident sequence analysis for a BWR (Boiling Water Reactor) during low power and shutdown operations

    SciTech Connect

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Dispersed-flow film boiling in rod-bundle geometry: steady-state heat-transfer data and correlation comparisons. [PWR; BWR

    SciTech Connect

    Yoder, G. L.; Morris, D. G.; Mullins, C. B.; Ott, L. J.; Reed, D. A.

    1982-03-01

    Assessment of six film boiling correlations and one single-phase vapor correlation has been made using data from 22 steady state upflow rod bundle tests (series 3.07.9). Bundle fluid conditions were calculated using energy and mass conservation considerations. Results of the steady state film boiling tests support the conclusions reached in the analysis of prior transient tests 3.03.6AR, 3.06.6B, and 3.08.6C. Comparisons between experimentally determined and correlation-predicted heat transfer coefficients, are presented.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    NASA Astrophysics Data System (ADS)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  11. 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.

  12. Electrochemical Potential (ECP) of Clean Heated Fuel Cladding Material and Structural SS under BWR Operating Conditions

    SciTech Connect

    Pop, Mike G.; Bell, Merl; Kilian, Renate; Dorsch, Thomas; Christian, Mueller

    2007-07-01

    To preliminarily monitor the relative effect of advanced water chemistry measures on SS structural material and fuel cladding in BWR environments a number of experiments were performed using laboratory equipment (recirculation loop, autoclave with heated electrodes, reference electrodes, etc.). The simulation of the plant condition was done without impurities or crud deposit contribution (clean surfaces). Subsequent testing, performed during 2007 and not yet cleared for release, is considering the effect of combined complex BWR chemistries and crud deposition. The heated Zircaloy fuel cladding tubing was prepared to simulate heat transfer by internal heating at levels existing in BWR (70 W/cm{sup 2}). For comparison purposes additional type SS347 electrode and unheated zirconium was used. A platinum electrode was used to measure the redox potential of the electrolyte. A high temperature Ag/AgCl electrode was used as a reference electrode. The assembly was installed in a recirculation 1 liter autoclave. Present report presents corrosion potential measurements performed under the following BWR water chemistry conditions (at 288 deg. C fluid exit temperature, 86 bar with surface temperature of Zirconium hot finger at 296 deg. C) - normal (inert) water conditions, - hydrogen injection in three steps from 0.68 ppm to 1.6 ppm, - oxygen injection in three steps from 2.4 ppm to 10 ppm - -methanol 2 ppm and oxygen 2 ppm in a close loop (without methanol refreshing) (authors)

  13. 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.

  14. Enhancing BWR Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect

    Gray S. Chang

    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 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. 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

  15. 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

  16. Reviews

    NASA Astrophysics Data System (ADS)

    2001-07-01

    A-LEVEL RESOURCES REVIEWS SPECIAL AS and A2 books and resources: deciding what to buy? SUMMARY Exam boards, specifications and support materials OCR (Oxford Cambridge and RSA Examinations) CORRECTION BOOK REVIEW Good Practice in Science Teaching WEB WATCH Astronomy and cosmology DVD REVIEW The Video Encyclopedia of Physics Demonstrations SOFTWARE REVIEW Graph Paper Printer

  17. Reviews

    NASA Astrophysics Data System (ADS)

    2003-07-01

    CD REVIEWS (346) Spectrum 7 Physics - Waves SOFTWARE REVIEW (347) Sound Packages BOOK REVIEW (350) Measured Tones, 2nd edition WEB WATCH (351) What’s the frequency, Kenneth? BOOK REVIEW (354) We know what you did last summer ... now do something better this summer

  18. Reviews

    NASA Astrophysics Data System (ADS)

    2001-03-01

    CD-ROM REVIEWS SPECIAL: Multimedia CD-ROMs WEB WATCH: Medical imaging BOOK REVIEW: Understanding Science Lessons CD-ROM REVIEWS SPECIAL Multimedia CD-ROMs: what do they offer to enhance physics teaching? PEAR: Physics Exercises for Assessment and Revision GCSE Physics 1998 33 72 Contact: Europress WEB WATCH Medical imaging BOOK REVIEW Understanding Science Lessons

  19. 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

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

    Energy Science and Technology Software Center (ESTSC)

    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

  1. 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.

  2. 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.

  3. Electropolishing process development for PWR steam generator channel heads

    SciTech Connect

    Asay, R.H.; Graves, P.; Guastaferro, C.T.; Spalaris, C.N. )

    1991-04-01

    A broad range of process parameters was established to smoothen the surface of 309 L weld clad overlay, prototypic of surfaces common is channel heads of replacement PWR (pressurized water reactor) steam generators. Mechanical and electropolishing steps were studied to explore process boundaries, which result in acceptable degree of surface smoothness, without compromising metallurgical properties. Recommended processes and acceptance criteria established in this work, can be applied to electropolish steam generator channel heads. Smooth surfaces are less likely to retain radioactive species, and potentially develop lower radiation fields when these components are placed into service. 7 refs., 11 figs., 12 tabs.

  4. 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.

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    SciTech Connect

    J.S. Tang

    2001-05-03

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

    SciTech Connect

    Ivanov, A.; Sanchez, V.; Hoogenboom, J. E.

    2012-07-01

    As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

  18. BWR control blade/channel box interaction models for SCDAP/RELAP5

    SciTech Connect

    Griffin, F.P.

    1993-12-01

    The core of a boiling water reactor (BWR) consists of an array of fuel assemblies with cross-shaped control blades located between these assemblies. Each fuel assembly consists of a fuel rod bundle surrounded by a Zircaloy channel box. Each control blade consists of small stainless steel absorber tubes filled with B{sub 4}C powder surrounded by a stainless steel blade sheath. Under severe accident conditions, material interactions between the B{sub 4}C, stainless steel, and Zircaloy would have a significant impact on the melting and subsequent relocation of the control blade and channel box structures. This paper describes a new BWR control blade/channel box model for the SCDAP/RELAP5 severe accident analysis code that includes the effects of these material interactions. The phenomena represented by this model and the modeling techniques are derived from ORNL analyses of the BWR severe fuel damage experiments. Two examples of the operation of this new model within SCDAP/RELAP5 are provided.

  19. 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.

  20. 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)

  1. 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)

  2. 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

  3. Reviews

    NASA Astrophysics Data System (ADS)

    2004-01-01

    BOOK REVIEWS (99) Complete A-Z Physics Handbook Science Magic in the Kitchen The Science of Cooking Science Experiments You Can Eat WEB WATCH (101) These journal themes are pasta joke Microwave oven Web links CD REVIEW (104) Electricity and Magnetism, KS3 Big Science Comics

  4. 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

  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. Pump and valve fastener serviceability in PWR nuclear facilities

    SciTech Connect

    Moisidis, N.T.; Ratiu, M.D.

    1996-02-01

    The results of several studies conducted on corrosion of carbon and low-alloy steels in borated water have shown that impingement of borated steam on ferritic steels or contact with a moist paste of boric acid can lead to high corrosion rates due to high local concentrations of boric acid on the surface. The corrosion process of the flange fasteners of pumps and valves is considered a material compatibility and equipment maintenance problem. Therefore, the nuclear utilities of pressurized water reactor (PWR) power plants can prevent this damage by implementing appropriate fastener steel replacement and extended inspections to detect and correct the cause of leakage. A 3-phase corrosion protection program is presented for implementation based on system operability, outage-related accessibility, and cost of fastener replacement versus maintenance frequency increase. A selection criterion for fastener material is indicated based on service limitation: preloading and metal temperature.

  7. Ultrasonic Backscattering in Polycrystalline Materials of Pwr Components

    NASA Astrophysics Data System (ADS)

    Chassignole, B.; Dupond, O.; Fouquet, T.; Rupin, F.

    2011-06-01

    The ultrasonic examination of metallic components of Pressurized Water Reactors (PWR) is an important challenge for the nuclear industry. During the past decades, EDF R&D has undertaken numerous studies in order to improve the NDT process on these applications and to help to their qualification. The present paper deals with the problem of the structural noise which can potentially disturbs the ultrasonic inspection. In particular, this study proposes a modeling approach to simulate the ultrasonic scattering due to coarse grain structures of polycrystalline materials. The methodology is based on the mixing of a grain scale description of the material and a 2D finite element code (ATHENA) developed by EDF to simulate the ultrasonic propagation in isotropic and anisotropic elastic media. The modeling results are compared to experimental acquisitions on mock-ups containing artificial defects.

  8. 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.

  9. 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.

  10. 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)

  11. Reviews.

    ERIC Educational Resources Information Center

    Journal of Chemical Education, 1990

    1990-01-01

    Reviewed are two computer software packages: "Introduction to Spectroscopy, IR, NMR & CMR," and "ASYSTANT" (a mathematical and statistical analysis software tool). Discussed are the functions, strengths, weaknesses, hardware requirements, components, level, and cost for each package. (CW)

  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

    Science Teacher, 1988

    1988-01-01

    Reviews four software packages: (1) Agents of Infection; (2) The Earth Science Series; (3) Investigating Electric Fields; and (4) Aquatic Biology-PC. Describes the hardware needed, bibliographic data and specific details of the programs themselves. (CW)

  14. 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)

  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. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    SciTech Connect

    Kavaklioglu, K.; Ikonomopoulos, A. )

    1993-01-01

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint.

  17. 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.

  18. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    SciTech Connect

    M. Krug; R. Shogan; A. Fero; M. Snyder

    2004-11-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR.

  19. 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.

  20. Evaluation of thermal mixing data from a model cold leg and downcomer. [PWR

    SciTech Connect

    Rothe, P.H.; Fanning, M.W.

    1982-12-01

    This report describes an evaluation of thermal mixing data obtained in a 1/5-scale, transparent model of the cold leg and downcomer of a Pressurized Water Reactor (PWR). The data are relevant to the phenomenon of fluid and thermal mixing following HPI (High Pressure Injection) of coolant water in a PWR loop. The data are reduced, correlated and compared with theoretically derived values and scaling approaches.

  1. 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.

  2. 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.

  3. 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…

  4. 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)

  5. 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…

  6. 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…

  7. 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)

  8. Reviews

    NASA Astrophysics Data System (ADS)

    2002-07-01

    BOOK REVIEWS (353) Dr Dyer's Academy Further Advanced Physics Physics 11-14, with Biology 11-14 and Chemistry 11-14 Nelson Modular Science: Books 1 and 2 Key Science: Physics, 3rd Edition Nelson Science: Physics, 2nd Edition Physics for AQA: Separate Award, Coordinated Award Physical Processes: A Visual Approach WEB WATCH (359) Physics Favourites: John Miller's selection

  9. Reviews.

    ERIC Educational Resources Information Center

    Adoption & Fostering, 1994

    1994-01-01

    Reviews 17 publications covering a variety of topics including abused disabled children; child placement; helping attachment disordered children; open adoption; family empowerment; teenage fostering; the Children Act of 1989 (Great Britain); family law; social services and child care law; and challenging racism in the early years. (TJQ)

  10. 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)

  11. 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

  12. Reliability assessment for a steam-driven jet injector for the IS-BWR design

    SciTech Connect

    Aybar, H.S.

    1996-12-31

    In recent years, several advanced reactor designs that utilize passive safety systems have been proposed as a result of public concern about nuclear power plant safety. System simplicity and maintainability and the utilization of passive features are the main design and safety objectives of the advanced reactors. The inherently safe, natural circulation boiling water reactor (IS-BWR) concept with a steam-driven jet injector (SDJI) as a passive emergency core cooling system has been proposed by the faculty and students in the Nuclear Engineering Program at the Ohio State University. The SDJI is a device without moving parts in which steam is used as an energy source to pump cold water from a pressure much lower than the steam pressure to a pressure higher than the steam pressure. Thus, the SDJI can be considered equivalent to a turbine-driven pump. All thermodynamics processes in the SDJI rely on direct contact transport phenomena between water and steam. The SDJI has been used as an emergency feedwater device in ships and a steam-jet air ejector to remove noncondensable gases from the condenser in modern steam power plants where the pressures are <1.7 MPa. The preliminary probabilistic reliability assessment for the emergency core cooling system of the IS-BWR has been reported. After recent experimental studies on the SDJI, its reliability has been reevaluated for the design improvement of the IS-BWR. The purpose of this paper is to give a reliability assessment of the SDJI after the recent development and a comparison with conventional systems.

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. Parameterization of Buoyancy Effects in Generic PWR Boron Dilution Scenarios

    SciTech Connect

    Galindo-Garcia, Ivan F.; Cotton, Mark A.; Axcell, Brian P.

    2006-07-01

    A computational investigation is undertaken into the role of buoyancy in a PWR boron dilution transient following a postulated Small Break Loss of Coolant Accident (SB-LOCA). In the scenario envisaged there is flow of de-borated and relatively high temperature water from a single cold leg into the downcomer; flow rates are typical of natural circulation conditions. The study focuses upon the development of boron concentration distributions in the downcomer and adopts a 3D-unsteady formulation of the mean flow equations in combination with the standard high-Reynolds-number k-{epsilon} turbulence model. It is found that the Richardson number (Ri = Gr/Re{sup 2}) is the most important group parameterizing the course of a concentration transient. At Ri values characterizing a 'baseline' scenario the results indicate that there is a stable, circumferentially-uniform, descent through the downcomer of a stratified region of low-borated fluid. Qualitatively the same behaviour is found at higher Richardson number, although at Ri values of approximately one-fifth the baseline level there is evidence of large-scale mixing and a consequent absence of concentration stratification. (authors)

  18. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

  19. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

  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. 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.

  2. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    PubMed

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689

  3. Determining a lateral load specification for downcomers during chugging in a Mark II containment. [BWR

    SciTech Connect

    Lehner, J.R.; Sonin, A.A.

    1983-01-01

    During a postulated LOCA in a BWR containment, steam is vented to the suppression pool and condensation-driven pressure oscillations may occur. In a Mark II type of containment with vertical downcomers these pressure oscillations may give rise to lateral loads near the downcomer exit which are impulsive in nature and random in direction. Domestic BWR vendors found that a dynamic load having a half sinusoidal shape with maximum amplitude F and duration tau, when applied to an analytic model of a Mark II downcomer, reproduced well the structural response observed in domestic and foreign lateral load tests. Because of the relative time scales involved, this could be thought of as an impulse specification with an impulse of amplitude F. Originally, the highest load observed in a domestic test conducted for the Mark II Owners had been suggested as an appropriate specification. At the request of NRC, Brookhaven National Laboratory and its consultants evaluated the results of several foreign and domestic lateral load tests in order to recommend an appropriate load amplitude.

  4. 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.

  5. 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)

  6. BWR ex-vessel steam explosion analysis with MC3D code

    SciTech Connect

    Leskovar, M.

    2012-07-01

    A steam explosion may occur, during a severe reactor accident, when the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. To resolve the open issues in steam explosion understanding and modeling, the OECD program SERENA phase 2 was launched at the end of year 2007, focusing on reactor applications. To verify the progress made in the understanding and modeling of fuel coolant interaction key phenomena for reactor applications a reactor exercise has been performed. In this paper the BWR ex-vessel steam explosion study, which was carried out with the MC3D code in conditions of the SERENA reactor exercise for the BWR case, is presented and discussed. The premixing simulations were performed with two different jet breakup modeling approaches and the explosion was triggered also at the expected most challenging time. For the most challenging case, at the cavity wall the highest calculated pressure was {approx}20 MPa and the highest pressure impulse was {approx}90 kPa.s. (authors)

  7. Crack growth rate in core shroud horizontal welds using two models for a BWR

    NASA Astrophysics Data System (ADS)

    Arganis Juárez, C. R.; Hernández Callejas, R.; Medina Almazán, A. L.

    2015-05-01

    An empirical crack growth rate correlation model and a predictive model based on the slip-oxidation mechanism for Stress Corrosion Cracking (SCC) were used to calculate the crack growth rate in a BWR core shroud. In this study, the crack growth rate was calculated by accounting for the environmental factors related to aqueous environment, neutron irradiation to high fluence and the complex residual stress conditions resulting from welding. In estimating the SCC behavior the crack growth measurements data from a Boiling Water Reactor (BWR) plant are referred to, and the stress intensity factor vs crack depth throughout thickness is calculated using a generic weld residual stress distribution for a core shroud, with a 30% stress relaxation induced by neutron irradiation. Quantitative agreement is shown between the measurements of SCC growth rate and the predictions of the slip-oxidation mechanism model for relatively low fluences (5 × 1024 n/m2), and the empirical model predicted better the SCC growth rate than the slip-oxidation model for high fluences (>1 × 1025 n/m2). The relevance of the models predictions for SCC growth rate behavior depends on knowing the model parameters.

  8. 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.

  9. A comparison of HLW-glass and PWR-borate waste glass

    NASA Astrophysics Data System (ADS)

    Luo, Shanggeng; Sheng, Jiawei; Tang, Baolong

    2001-09-01

    Glass can incorporate a wide variety of wastes ranging from high level wastes (HLW) to low and intermediate level wastes (LILW). A comparison of HLW-Glass and PWR-borate waste glass is given in this paper. The HLW glass formulation named GC-12/9B and 90-19/U can incorporate 16-20 wt% HLW at 1100°C or 1150°C. The borate waste glass named SL-1 can incorporate 45 wt% borate waste generated from PWR. Their physical properties, characteristic temperatures, chemical durability and leach behavior are summarized here. The comparison indicates: the PWR-glass SL-1 can incorporate up to 45 wt% waste oxides at lower melting temperature (1000°C) in agreement with minimum additive waste stabilization (MAWS) approach; owing to the PWR-borate glass contain less Si and more B and Na, its mass loss is higher than HWR-glass; both HLW-glass and PWR-borate glass have favorable chemical durability and the same leaching phenomena, i.e., Na is mostly depleted, but Ca, Mg, Al and Ti are enriched in the leached surface layer.

  10. 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)

  11. Post-partum weight retention in women in Asia: a systematic review.

    PubMed

    Cheng, H-R; Walker, L O; Tseng, Y-F; Lin, P-C

    2011-10-01

    Obesity is a health issue for people in Asia, and many studies support that childbearing may contribute to women's obesity. However, most studies of post-partum weight retention (PWR) focus on Western countries. This paper aimed to review recent data on PWR and discuss the related factors of PWR for women in Asia. Three electronic databases - MEDLINE, CINAHL and PsycINFO - were searched for data from January 1990 to August 2010. Twelve studies were identified as containing information related to PWR. Chinese, Taiwanese and Korean women tend to have higher gestational weight gain and PWR than women in other Asian countries in this review. The average reported gestational weight gain in these studies ranged from 8.3 kg to 18.0 kg, and the average weight retention at 6 months post partum in various countries ranged from 1.56 kg to 4.1 kg. Gestational weight gain had a positive effect on PWR, but there were no consistent conclusions for other factors. Only a few psychosocial and behavioural factors have been studied. Moreover, no one has assessed changes in waist circumference and waist-to-hip ratio for measuring central obesity and ensuring risk of chronic disease. Focusing on changes in maternal weight and body composition, and considering the impact of psychological and behavioural factors are suggested for future studies in Asia. PMID:21545682

  12. Materials Reliability Program: Fracture Toughness Testing of Decommissioned PWR Core Internals Material Samples (MRP-160) Non-Proprietary Version

    SciTech Connect

    M. E. Krug; R. P. Shogan

    2005-09-30

    Pressurised water reactor (PWR) cores operate under extreme envrionmental conditions due to coolant chemistry, operating temperature and neutron exposure. Extending the life of PWRs requires detailed knowledge of teh changes in mechanical and corrosion properties of teh structural austenitic stainless steel components adjacent to the fuel. This report contains results of fracture toughness testing of samples machined from decommissioned PWR reactor internals.

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2010-06-01

    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 233U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium & uranium confinement in PWR.

  15. Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

    SciTech Connect

    Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G; Salko, Robert K; Evans, Thomas M; Turner, John A; Belcourt, Kenneth; Hooper, Russell; Schmidt, Rodney

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

  16. PWR containment structures license renewal industry report: Revision 1. Final report

    SciTech Connect

    Deng, D.; Renfro, J.; Statton, J.

    1994-07-01

    Reinforced concrete, prestressed concrete, and freestanding steel PWR containment structures and components 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 structures and components 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 containment structures and components for license renewal.

  17. Switching from deferred dismantling to immediate dismantling: the example of Chooz A, a French PWR

    SciTech Connect

    Grenouillet, Jean-Jacques

    2007-07-01

    Located in the north of France, close to Belgian border, Chooz A is the first PWR that was built in France from 1962 to 1967. When it was shutdown in 1991, a deferred dismantling strategy was selected. Further to an evolution of EDF decommissioning strategy in 2001, the decommissioning of the plant was accelerated by reducing the safe enclosure period to only a few years. Thus Chooz A will be the first PWR to be fully dismantled in France and it gives a good insight of what is needed to reactivate a plant for final dismantling after a safe enclosure period. (author)

  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. 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.

  20. Characterization of debris/concrete interactions for advanced research reactor and commercial BWR severe accidents

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.; Greene, S.R.

    1991-01-01

    The core concrete interaction (CCI) is an important phase of any severe accident where the reactor vessel has failed and core debris is relocated onto the containment basemat. In recent calculations performed at the Oak Ridge National Laboratory (ORNL), CCI has been studied for severe accidents occurring in a commercial Boiling Water Reactor (BWR) and in a high-power density Department of Energy (DOE) research reactor that is currently in the conceptual design stage. Because of differences in the debris decay heating level, core debris composition and inventory, and containment design, the characteristics of the resulting CCI and containment response are different for the two reactor types. Furthermore, proper selection of the basemat concrete type and the provision of an overlying water pool are found to be significant CCI mitigating factors for the research reactor and thus constitute important design considerations for any future reactor type. 10 refs., 4 figs., 1 tab.

  1. 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.

  2. 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.

  3. RAMONA-3B/MINET composite representation of BWR thermal-hydraulic systems

    SciTech Connect

    Van Tuyle, G.J.; Slovik, G.; Cazzoli, E.G.; Nepsee, T.C.; Guppy, J.G.

    1985-01-01

    The modification and interfacing of two computer codes, RAMONA-3B and MINET, for the thermal hydraulic transient analysis of a Boiling Water Reactor nuclear steam supply system, is described. The RAMONA-3B code provides for multi-channel thermal hydraulics and three-dimensional (or one-dimensional) neutron kinetics analysis of a boiling water reactor core. The RAMONA-3B system representation terminates at the end of the steam line and at the junction of the feedwater line at the vessel inlet. By interfacing RAMONA-3B with MINET, a generic balance-of-plant systems analysis code, a complete BWR systems code with detailed core modeling was obtained. The result is a code of particular importance to the analysis of transients such as ATWS. A comparison between the 3-D and 1-D neutronics representation is provided, along with a test case utilizing the composite RAMONA-3B/MINET code.

  4. 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.

  5. Probability of failure in BWR reactor coolant piping: Volume 1, Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1989-03-01

    This report summarizes a probabilistic reliability evaluation of BWR reactor coolant piping performed for the US Nuclear Regulatory Commission (NRC) by the Lawrence Livermore National Laboratory (LLNL). In this evaluation, LLNL estimated the probability of a double-ended guillotine break (DEGB) in the main steam, feedwater, and recirculation loop piping of a representative Mark I BWR plant. Two causes of pipe break were considered: crack growth at welded joints, and the earthquake-induced failure of supports for piping and components. A probabilistic fracture mechanics model, including intergranular stress corrosion cracking (IGSCC) in Types 304 and 316NG stainless steels, was used to estimate the probability of crack-induced pipe break. The probability of pipe break indirectly caused by support failure was estimated by applying reliability techniques to supports for ''heavy components'' such as the reactor pressure vessel, as well as to conventional pipe supports such as spring hangers and snubbers. Our probabilistic fracture mechanics evaluation found that the probability of crack-induced DEGB in main steam, feedwater and, if IGSCC is not a factor, recirculation piping is very low. In ISGCC-susceptible Type 304SS piping, stress corrosion dominates the probability of DEGB due mainly to cracks that initiate during the first few years of plant life; replacing Type 304 piping with IGSCC-resistant Type 316NG lowers DEGB probabilities by several orders of magnitude. We also found that the probability of pipe break caused by seismically-induced support failure is low regardless of whether ''heavy component'' supports or conventional pipe supports are being considered. 23 refs., 34 figs., 16 tabs.

  6. Effects of improved modeling on best estimate BWR severe accident analysis

    SciTech Connect

    Hyman, C.R.; Ott, L.J.

    1984-01-01

    Since 1981, ORNL has completed best estimate studies analyzing several dominant BWR accident scenarios. These scenarios were identified by early Probabilistic Risk Assessment (PRA) studies and detailed ORNL analysis complements such studies. In performing these studies, ORNL has used the MARCH code extensively. ORNL investigators have identified several deficiencies in early versions of MARCH with regard to BWR modeling. Some of these deficiencies appear to have been remedied by the most recent release of the code. It is the purpose of this paper to identify several of these deficiencies. All the information presented concerns the degraded core thermal/hydraulic analysis associated with each of the ORNL studies. This includes calculations of the containment response. The period of interest is from the time of permanent core uncovery to the end of the transient. Specific objectives include the determination of the extent of core damage and timing of major events (i.e., onset of Zr/H/sub 2/O reaction, initial clad/fuel melting, loss of control blade structure, etc.). As mentioned previously the major analysis tool used thus far was derived from an early version of MARCH. BWRs have unique features which must be modeled for best estimate severe accident analysis. ORNL has developed and incorporated into its version of MARCH several improved models. These include (1) channel boxes and control blades, (2) SRV actuations, (3) vessel water level, (4) multi-node analysis of in-vessel water inventory, (5) comprehensive hydrogen and water properties package, (6) first order correction to the ideal gas law, and (7) separation of fuel and cladding. Ongoing and future modeling efforts are required. These include (1) detailed modeling for the pressure suppression pool, (2) incorporation of B/sub 4/C/steam reaction models, (3) phenomenological model of corium mass transport, and (4) advanced corium/concrete interaction modeling. 10 references, 17 figures, 1 table.

  7. 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

  8. 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

  9. 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.

  10. Crack propagation in stainless steels and nickel base alloys in a commercial operating BWR

    SciTech Connect

    Jenssen, A.; Morin, U.; Bengtsson, B.; Jansson, C.

    1995-12-31

    Crack propagation was investigated to study critical stress intensity factors for intergranular stress corrosion cracking (IGSCC), and crack growth rates in various materials. Modified bolt loaded compact tension (CT) specimens were exposed to BWR normal water chemistry (NWC) in a commercially operating BWR. The test facility was a pressure vessel, originally designed for high temperature magnetite filters. Stainless steels (SS) of Types 304 SS and 316 SS were included in the test matrix, as well as the Ni base weld materials alloys 82 and 182. The SS were investigated both in sensitized and in cold worked condition. For alloy 182 various parameters were studied, such as the effect of the carbon stabilization parameter, and the as-welded condition versus a post weld heat treatment (PWHT). Crack growth was measured annually, during the normal outages, by an optical microscope. The results were evaluated as crack growth rate as a function of stress intensity. A few specimens have been removed from testing for fractographic examination. Most of the specimens were exposed to NWC for more than 30,000 hours. Alloy 82 in as welded condition was found to be susceptible to IGSCC, at least at stress intensities above 30 MPa{radical}m. For alloy 182, in as welded condition, significant crack growth was detected in all specimens. No beneficial effect of the carbon stabilization parameter could be found. PWHT had a beneficial effect on the IGSCC susceptibility of alloy 182, and at stress intensities below 30 MPa{radical}m the crack growth rates were one to two orders of magnitude lower, compared to alloy 182 in as welded condition. As expected, an increasing susceptibility to IGSCC with increasing degree of cold work was found for stainless steel. At 5% cold work Type 304 SS cracked at a higher rate than Type 316NG with the same degree of cold work. At 20% cold work Type 304 SS and Type 316NG cracked at essentially the same rate.

  11. 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

  12. 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.

  13. 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.

  14. Technical report on material selection and processing guidelines for BWR coolant pressure boundary piping. Draft report. Revision 2

    SciTech Connect

    Hazelton, W.S.

    1986-06-01

    This report updates and supersedes the technical positions NRC established in NUREG-0313, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' published in July 1977, and its subsequent revision published in July 1980. This report sets forth the NRC staff's revised acceptable methods to control the intergranular stress corrosion cracking susceptibility of BWR ASME Code Class 1, 2, and 3 pressure boundary piping and safe ends. For piping that does not fully comply with the material selection, testing, and processing guideline combinations of this document, varying degrees of augmented inservice inspection will be required, pursuant to 10 CFR 50.55a(g)(6)(ii). This revision also includes guidance regarding crack evaluation and weld overlay repair methods for long term operation or for continuing interim operation of plants until a more permanent solution is implemented.

  15. Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)

    SciTech Connect

    Burns, C. J. and Aumiler, D. L.

    2006-04-26

    The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF.

  16. Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

    SciTech Connect

    Dingman, S.; Camp, S.; LaChance, J.; Mary Drouin

    1995-12-31

    This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat removal are important contributors for the BWR 3/4 plants, while transients, station blackout sequences, and loss-of-coolant accidents are important for the Westinghouse 4-loop plants. The key factors that contribute to the variability in the results are discussed. The results are often driven by plant-specific design and operational characteristics, but differences in modeling approaches are also important for some accident classes.

  17. 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.

  18. 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)

  19. 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.

  20. Analysis of the dose rate produced by control rods discharged from a BWR into the irradiated fuel pool.

    PubMed

    Ródenas, J; Gallardo, S; Abarca, A; Juan, V

    2010-01-01

    BWR control rods become activated by neutron reactions into the reactor. Therefore, when they are withdrawn from the reactor, they must be stored into the storage pool for irradiated fuel at a certain depth under water. Dose rates on the pool surface and the area surrounding the pool should be lower than limits for workers. The MCNP code based on the Monte Carlo method has been applied to model this situation and to calculate dose rates at points of interest. PMID:19836252

  1. 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. PMID:21093278

  2. 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.

  3. Proceedings: 1983 Workshop on Secondary-Side Stress Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator Tubing

    SciTech Connect

    1986-03-01

    Participants in this international workshop discussed research investigating mechanisms and propagation rates of intergranular corrosion in PWR steam generators. Laboratory test results, which have been consistent with power plant experience, permitted preliminary definition of corrosion rates in alloy 600 tubing.

  4. 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

  5. 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.

  6. 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.

  7. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    SciTech Connect

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  8. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    SciTech Connect

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  9. 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.

  10. Hydrothermal deposition of zirconia coatings on pre-oxidized BWR structural materials

    NASA Astrophysics Data System (ADS)

    Zhou, Z. F.; Chalkova, E.; Lvov, S. N.; Chou, P. H.

    2008-09-01

    An in situ hydrothermal deposition process is being developed to apply a thin coating of zirconia onto the structural materials used in Boiling Water Reactors as a potential method for mitigating intergranular stress corrosion cracking. The process has successfully deposited ZrO 2 onto as-received interior surfaces of 304 stainless steel and Alloy 600 tubes [Z.F. Zhou, E. Chalkova, S.N. Lvov, P. Chou, R. Pathania, Corros. Sci. 49 (2007) 830]. This paper discusses the application of the coating on specimens with different surface conditions: as-received; ground to remove the as-received surface; and ground and pre-oxidized. For comparable deposition parameters and for a given substrate, the different surface conditions did not influence the morphology or the thickness of the coating, but had a substantial impact on adhesion. As in our previous study, electrochemical potentials of the coated specimens in simulated BWR environment were not clearly lower than those of uncoated specimens [Zhou et al., 2007].

  11. 100% MOX BWR experimental program design using multi-parameter representative

    SciTech Connect

    Blaise, P.; Fougeras, P.; Cathalau, S.

    2012-07-01

    A new multiparameter representative approach for the design of Advanced full MOX BWR core physics experimental programs is developed. The approach is based on sensitivity analysis of integral parameters to nuclear data, and correlations among different integral parameters. The representativeness method is here used to extract a quantitative relationship between a particular integral response of an experimental mock-up and the same response in a reference project to be designed. The study is applied to the design of the 100% MOX BASALA ABWR experimental program in the EOLE facility. The adopted scheme proposes an original approach to the problem, going from the initial 'microscopic' pin-cells integral parameters to the whole 'macroscopic' assembly integral parameters. This approach enables to collect complementary information necessary to optimize the initial design and to meet target accuracy on the integral parameters to be measured. The study has demonstrated the necessity of new fuel pins fabrication, fulfilling minimal costs requirements, to meet acceptable representativeness on local power distribution. (authors)

  12. Flow-induced vibration characteristics of the BWR/5-201 jet pump

    SciTech Connect

    LaCroix, L.V.

    1982-09-01

    A General Electric boiling water reactor BWR/5-201 jet pump was tested for flow-induced vibration (FIV) characteristics in the Large Steam Water Test Facility at Moss Landing, CA, during the period June-July 1978. High level periodic FIV were observed at reactor operating conditions (1027 psia, 532/sup 0/F and prototypical flow rates) for the specific single jet pump assembly tested. High level FIV of similar amplitude and character have been shown capable of damaging jet pump components and associated support hardware if allowed to continue unchecked. High level FIV were effectively suppressed in two special cases tested: (1) lateral load (>500 lb) at the mixer to diffuser slip joint; and (2) a labyrinth seal (5 small, circumferential grooves) on the mixer at the slip joint. Stability criteria for the particular jet pump tested were developed from test data. A cause-effect relationship between the dynamic pressure within the slip joint and the jet pump vibration was established.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. BWR (boiling-water reactor) radiation control: In-plant demonstration at Vermont Yankee: Final report

    SciTech Connect

    Palino, G.F.; Hobart, R.L.; Sawochka, S.G.

    1987-10-01

    Results of the RP1934 program, which was established by EPRI in 1981 to demonstrate the adequacy of BRAC program (RP819) principles for BWR radiation control at Vermont Yankee, are presented. Evaluations were performed of the effectiveness of optimization of purification system performance, control of feedwater dissolved oxygen concentrations, minimization of corrosion product and ionic transport, and improved startup, shutdown, and layup practices. The impact on shutdown radiation levels of these corrective actions was assessed based on extensive primary system radiation survey and component gamma scan data. Implementation of the BRAC recommendations was found to be insufficient to reduce the rate of activity buildup on out-of-core surfaces at Vermont Yankee, and additional corrective actions were found necessary. Specifically, replacement of cobalt-bearing materials in the control rod drive pins and rollers and feedwater regulating valves was pursued as was installation of electropolished 316 stainless steel during a recirculation piping replacement program. Aggressive programs to further reduce copper concentrations in the reactor water by improving condensate demineralizer efficiency and to minimize organic ingress to the power cycle by reducing organic concentrations in recycled radwaste also were undertaken. Evaluations of the impact on activity buildup of several pretreatment processes including prefilming in moist air, preexposure to high temperature water containing zinc, and electropolishing also were performed in a test loop installed in the reactor water cleanup system. A significant beneficial impact of electropolishing was shown to be present for periods up to 6000 hours.

  18. Identification and assessment of containment and release management strategies for a BWR Mark I containment

    SciTech Connect

    Lin, C.C.; Lehner, J.R. )

    1991-09-01

    This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark 1 type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark 1 containment could face during the course of a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenges. A safety objective tree is developed which provides the connection between the safety objectives, the safety functions, the challenges, and the strategies. The strategies were assessed by applying them to certain severe accident sequence categories which have one or more of the following characteristics: have high probability of core damage or high consequences, lead to a number of challenges, and involve the failure of multiple systems. 59 refs., 55 figs., 27 tabs.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

    SciTech Connect

    Courau, T.; Moustafa, S.; Plagne, L.; Poncot, A.

    2013-07-01

    As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 10{sup 9} degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S{sub 8} 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP{sub 5} DIABOLO fission production rate is lower than 0.2% compared to the S{sub 8} DOMINO reference for this 3D PWR core model. (authors)

  4. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3. International Agreeement Report

    SciTech Connect

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

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser.

  5. 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.

  6. Pressure-vessel-damage fluence reduction by low-leakage fuel management. [PWR

    SciTech Connect

    Cokinos, D.; Aronson, A.L.; Carew, J.F.; Kohut, P.; Todosow, M.; Lois, L.

    1983-01-01

    As a result of neutron-induced radiation damage to the pressure vessel and of an increased concern that in a PWR transient the pressure vessel may be subjected to pressurized thermal shock (PTS), detailed analyses have been undertaken to determine the levels of neutron fluence accumulation at the pressure vessels of selected PWR's. In addition, various methods intended to limit vessel damage by reducing the vessel fluence have been investigated. This paper presents results of the fluence analysis and the evaluation of the low-leakage fuel management fluence reduction method. The calculations were performed with DOT-3.5 in an octant of the core/shield/vessel configuration using a 120 x 43 (r, theta) mesh structure.

  7. 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)

  8. 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.

  9. 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)

  10. 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.

  11. 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

  12. 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.

  13. Design and analysis of a prestressed concrete reactor vessel for a 2000 MWe super large-sized BWR

    SciTech Connect

    James, R.J.; Rashid, Y.R.; Gou, P.F.; Sawyer, C.D.; Tanaka, S.; Shirai, Y.; Mori, M.; Takekuro, I.

    1996-06-01

    The super large-sized natural circulation BWR (Boiling Water Reactor) which eliminates the need for a short-term ECCS (Emergency Core Cooling System) is conceived as a sophisticated approach to future LWR (Light Water Reactor) plant design, promising simplified system design with improved safety and economics. This reactor vessel design requires the technology of prestressed concrete reactor vessels (PCRVs), because the large reactor vessel with a sufficient water inventory to eliminate ECCS is beyond the current technology of steel reactor vessels. A large power output of 2,000 MWe was selected for the initial scoping study.

  14. 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.

  15. 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.

  16. 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)

  17. Severe degradation of BWR fuel failures: Coolant activity analysis. Interim report

    SciTech Connect

    Klepfer, H.H.; Kjaer-Pedersen, N.; Rank, P.; Ozer, O.

    1993-11-01

    The Electric Power Research Institute (EPRI) has been investigating an apparent recent increase in the frequency of severely degraded failed fuel rods in BWRs. These incidents involve abnormal levels of coolant radioactivity from the failed rod, and are followed by a steep increase in the activity from tramp uranium, most of which is deposited on fuel rod surfaces and therefore tends to carry over to subsequent operating cycles. Analyses of coolant activity from several BWRs experiencing degradation of failed fuel rods were carried out with extensive use of the CHIRON code, developed for EPRI by S. Levy Incorporated. CHIRON is based on steady-state conditions, but its trending plot capability was extremely useful for studying the time variation of radiochemical parameters. It is observed that the failure degradation process usually starts three to eight months after a primary failure occurrence. It is possible from measurements of Np{sup 239} coolant activity to reach quite accurate estimates of the amount of fuel released from the failed rod(s) over time. The neptunium calculations can also be used to estimate the size/length of the cladding rupture. An equation has been developed to predict offgas at the start of the next cycle from offgas at end of cycle. It is concluded that a failed BWR fuel rod can deteriorate severely within a few months, and therefore all failure occurrences should be acted upon with the serious consequences of severe damage in mind. A set of recommended guidelines is established for monitoring and interpreting activity signals, and for acting upon failure observations.

  18. Thermalhydraulic processes in the reactor coolant system of a BWR (boiling water reactor) under severe accident conditions

    SciTech Connect

    Hodge, S.A.

    1989-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MSIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a brief discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermalhydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are briefly described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom Atomic Power Station is presented. 12 refs., 4 figs., 1 tab.

  19. 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.

  20. 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.

  1. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report.

    SciTech Connect

    Tentner, A.; Nuclear Engineering Division

    2009-10-13

    A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

  2. 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)

  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. MELCOR analyses of severe accident scenarios in Oconee, a B&W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S.; Foulds, R.

    1993-03-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock & Wilcox (B&W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  7. MELCOR analyses of severe accident scenarios in Oconee, a B W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S. ); Foulds, R. )

    1993-01-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock Wilcox (B W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  8. 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.

  9. Nuclear data uncertainties by the PWR MOX/UO{sub 2} core rod ejection benchmark

    SciTech Connect

    Pasichnyk, I.; Klein, M.; Velkov, K.; Zwermann, W.; Pautz, A.

    2012-07-01

    Rod ejection transient of the OECD/NEA and U.S. NRC PWR MOX/UO{sub 2} core benchmark is considered under the influence of nuclear data uncertainties. Using the GRS uncertainty and sensitivity software package XSUSA the propagation of the uncertainties in nuclear data up to the transient calculations are considered. A statistically representative set of transient calculations is analyzed and both integral as well as local output quantities are compared with the benchmark results of different participants. It is shown that the uncertainties in nuclear data play a crucial role in the interpretation of the results of the simulation. (authors)

  10. 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.

  11. User's guide for the PWR LOCA analysis capability of the WRAP-EM system

    SciTech Connect

    Beranek, F; Gregory, M V

    1980-02-01

    The Water Reactor Analysis Package (WRAP) has been expanded to provide the capability to analyze loss-of-coolant accidents (LOCAs) in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) by using evaluation models (EMs). The input specifications for modules in the WRAP-EM system are presented in this document along with the JOSHUA input templates. This document, along with the WRAP user's guide, provides a step-by-step procedure for setting up a PWR data base for the WRAP-EM system. 12 refs.

  12. 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.

  13. CRACK GROWTH RESPONSE OF ALLOY 152 AND 52 WELD METALS IN SIMULATED PWR PRIMARY WATER

    SciTech Connect

    Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2009-12-01

    The crack growth response of alloy 152 and 52 weld metals has been measured in simulated PWR primary water at both high (325-350 C) and low (50 C) temperatures. Tests were performed on samples machined from alloy 152 or 52 mockup welds. Propagation rates under cycle + hold and constant K conditions at high temperatures show stable, but extremely low SCC growth rates. The most significant intergranular cracking occurred during cycling at 50 C, particularly for the alloy 152 weld metal at high stress intensity.

  14. Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients. [PWR

    SciTech Connect

    Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

    1980-01-01

    The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations.

  15. Neutronics and safety characteristics of a 100% MOX fueled PWR using weapons grade plutonium

    SciTech Connect

    Biswas, D.; Rathbun, R.; Lee, Si Young; Rosenthal, P.

    1993-12-31

    Preliminary neutronics and safety studies, pertaining to the feasibility of using 100% weapons grade mixed-oxide (MOX) fuel in an advanced PWR Westinghouse design are presented in this paper. The preliminary results include information on boron concentration, power distribution, reactivity coefficients and xenon and control rode worth for the initial and the equilibrium cycle. Important safety issues related to rod ejection and steam line break accidents and shutdown margin requirements are also discussed. No significant change from the commercial design is needed to denature weapons-grade plutonium under the current safety and licensing criteria.

  16. Development of inspection systems for alloy 600 nozzles of PWR reactor vessel

    SciTech Connect

    Unate, K.; Ideo, M.; Sanagawa, T.; Shirai, T.; Araki, Y.

    1995-08-01

    PWR reactor vessels have alloy 600 nozzles at top and bottom heads. The former are head penetration nozzles for CRDM, and the latter are bottom mounted instrumentation nozzles. The authors have developed inspection systems of two types for each nozzle to confirm the soundness. ECT and UT Techniques are employed for both systems. These systems are controlled remotely and enable to reduce radiation exposure, inspection time and number of inspectors. Based on the functional tests using full scale mockups, the reliabilities and effectiveness of both systems were confirmed.

  17. 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

  18. 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

  19. 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

  20. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    SciTech Connect

    Chiang, R. T.

    2013-07-01

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  1. 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.

  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. Experiment data report for LOFT anticipated transient-without-scram Experiment L9-3. [PWR

    SciTech Connect

    Bayless, P.D.; Divine, J.M.

    1982-05-01

    Selected pertinent and uninterpreted data from the third anticipated transient with multiple failures experiment (Experiment L9-3) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large (approx. 1000 MW(e)), commercial PWR operations. Experiment L9-3 simulated a loss-of-feedwater anticipated transient without scram. The loss-of-feedwater accident led to an increase in the primary coolant system temperature and pressure. Both the experiment power-operated relief valve (PORV) and safety relief valve opened and were able to limit and control the pressure transient. The plant was then recovered with the control rods still withdrawn by injecting 7200-ppM borated water, manually cycling the PORV and feeding and bleeding the steam generator.

  4. 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)

  5. 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)

  6. Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

    NASA Astrophysics Data System (ADS)

    Ga¨rtner, M.; Fischer, G.

    1987-06-01

    To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

  7. 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.

  8. 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.

  9. The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

    NASA Astrophysics Data System (ADS)

    Urquidi-Macdonald, Mirna; Pitt, Jonathan; Macdonald, Digby D.

    2007-05-01

    A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH( T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields.

  10. Analysis of loss of off-site power with a PWR at shutdown

    SciTech Connect

    Chu, T.L.; Yoon, W.H.; Fitzpatrick, R.G.

    1987-01-01

    In many probabilistic risk assessments (PRAs), loss of offsite power (LOOP) when a nuclear power plant is operating was found to be a significant contributor to core damage. The purpose of this study is to provide an analysis of a LOOP event that occurs while a pressurized water reactor (PWR) is shut down. The importance of such an analysis was recognized as part of a study to evaluate the core damage frequency due to a loss of decay heat removal (DHR) capability during an outage. When a PWR is in a shutdown condition, there are relatively few technical specification requirements on the operability of safety systems. In fact, some safety systems are intentionally disabled, i.e., the safety injection system and nonoperating charging pumps. Another problem when the reactor is shut down is that the reactor coolant system (RCS) may be partially drained and the steam generators may be unavailable. To determine the time available for operator actions, given that a LOOP occurs during shutdown and the DHR capability is lost, a simple thermal model has been developed. Similar calculations have been performed for other phases of refueling and maintenance outages. A total core damage frequency due to LOOP while the plant is in shutdown has been calculated to be 5.9 x 10/sup -6//yr. This is approximately twice the core damage frequency calculated for LOOP when the plant is at power.

  11. Evaluation of the Sub-Channel Code COBRA-TF for Prediction of BWR Fuel Assembly Void Fraction Distribution

    SciTech Connect

    Aydogan, Fatih; Hochreiter, Lawrence E.; Ivanov, Kostadin; Rhee, Gene; Sartori, Enrico

    2006-07-01

    Good quality experimental data is needed to refine the thermal hydraulic models for the prediction of rod bundle void distribution and critical heat flux (CHF) or dry-out. The Nuclear Power Engineering Corporation (NUPEC) has provided a valuable database to evaluate the thermal hydraulic codes [1]. Part of this database was selected for the NUPEC BWR Full-size Fine-Mesh Bundle Tests (BFBT) benchmark sponsored by US NRC, METI-Japan, NEA/OECD and Nuclear Engineering Program of the Pennsylvania State University (PSU). Twenty-five organizations from ten countries have confirmed their intention to participate and will provide code predictions to be compared to the measured data for a series of defined exercises within the framework of the BFBT benchmark. This benchmark data includes both the fine-mesh high quality sub-channel void fraction and critical power data. Using a full BWR rod bundle test facility, the void distribution was measured at mesh sizes smaller than the sub-channel by using a state-of-the-art computer tomography (CT) technology [1]. Experiments were performed for different pressures, flow rates, exit qualities, inlet sub-cooling, power distributions, spacer types and assembly designs. There are microscopic and sub-channel averaged void fraction data from the CT scanner at the bundle exit as well as X-ray densitometer void distribution data at different elevation levels in the rod bundle. Each sub-channel's loss coefficient was calculated with using the Rehme method [2,3], and a COBRA-TF sub-channel model was developed for the NUPEC facility. The BWR assembly that was modeled with COBRA-TF includes two water rods at the center. The predicted sub-channel void fraction values from COBRA-TF are compared with the bundle exit void fraction values measured using the CT-scanner void fraction from the BFBT benchmark data. Different plots are used to examine the code prediction of the void distribution at a sub-channel level for the different sub-channels within

  12. 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.

  13. 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.

  14. Proceedings: 1984 Workshop on Secondary-Side Stress Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator Tubing

    SciTech Connect

    1986-03-01

    During 1984, research investigating intergranular corrosion and stress corrosion cracking in PWR steam generators provided data to formulate a corrosion-product transport theory. In addition, the research showed that changing the pH of liquids in generator crevices will retard and sometimes arrest the corrosion process.

  15. End-of-life destructive examinations of Zircaloy maximum depletion blanket fuel plates from the Shippingport PWR Core 2

    SciTech Connect

    Clayton, J.C.; Kammenzind, B.F.; Senio, P.; Sherman, J.

    1993-10-01

    Destructive examinations were performed on four Shippingport PWR Core 2 maximum fluence and depletion blanket plates for surface integrity, corrosion oxide thickness, and hydrogen absorption of the Zircaloy-4 cladding. The Shippingport PWR Core 2 operated for 23,360 effective full power hours (EFPH) (62,235 hot hours) at an average coolant temperature of 536{degrees}F (280{degrees}C) and a peak neutron flux of 0.6{times}10{sup 14}n/cm{sup 2}/s. The end-of-life examination program included measurements on three PWR-2 beta-quenched blanket fuel plates and one alpha-annealed blanket end plate. The examinations consisted of optical and scanning electron microscopy (SEM) inspections, direct metallographic oxide thickness measurements, and hydrogen extraction analyses on a joined element pair from the peak fluence (132{times}10{sup 20} n/cm{sup 2}), maximum depletion (13.5{times}10{sup 20} fissions/cc)PWR-2 blanket cluster.

  16. 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.

  17. 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.

  18. COBRA/TRAC analysis of the PKL reflood test K9. [PWR

    SciTech Connect

    Wilkins, C.A.; Thurgood, M.J.

    1982-08-01

    Experiments at the Primaerkreislaeufe (PKL) test facility in Erlangen, Germany, simulated the refill and reflood procedure after a loss-of-coolant accident (LOCA) in the primary coolant system of a 1300-MW pressurized water reactor (PWR). COBRA/TRAC, a thermal-hydraulics analysis code developed at the Pacific Northwest Laboratory, was used to model experiment K9 of the PKL test series (completed December 1979). The COBRA/TRAC code, which utilizes COBRA-TF as the vessel module and TRAC-P1A for the remaining components, was designed to analyze LOCAs in PWRs. PKL-K9 was characterized by a double-ended guillotine break in the cold leg with emergency core cooling water injected into the cold legs. COBRA/TRAC was able to successfully predict lower-core temperature profiles and quench times, upper-core temperature profiles until the quench, upper plenum and break pressures, and correct trends in collapsed water levels.

  19. 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.

  20. 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.

  1. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    SciTech Connect

    Pasichnyk, I.; Perin, Y.; Velkov, K.

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  2. 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.

  3. 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.

  4. 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

  5. Calculation of the neutron source distribution in the VENUS PWR Mockup Experiment

    SciTech Connect

    Williams, M.L.; Morakinyo, P.; Kam, F.B.K.; Leenders, L.; Minsart, G.; Fabry, A.

    1984-01-01

    The VENUS PWR Mockup Experiment is an important component of the Nuclear Regulatory Commission's program goal of benchmarking reactor pressure vessel (RPV) fluence calculations in order to determine the accuracy to which RPV fluence can be computed. Of particular concern in this experiment is the accuracy of the source calculation near the core-baffle interface, which is the important region for contributing to RPV fluence. Results indicate that the calculated neutron source distribution within the VENUS core agrees with the experimental measured values with an average error of less than 3%, except at the baffle corner, where the error is about 6%. Better agreement with the measured fission distribution was obtained with a detailed space-dependent cross-section weighting procedure for thermal cross sections near the core-baffle interface region. The maximum error introduced into the predicted RPV fluence due to source errors should be on the order of 5%.

  6. A predictive model for corrosion fatigue crack growth rates in RPV steels exposed to PWR environments

    SciTech Connect

    Atkinson, J.D.; Chen, Z.; Yu, J.

    1995-12-31

    Corrosion fatigue crack propagation rates have been measured in A533B Class 1 plate in stagnant PWR primary water for a range of steel sulphur contents, temperature and corrosion potential values. Parametric descriptions of the data collected under constant rig conditions give good correlations for each variable and are consistent with a crack tip environment controlled process related to sulphur chemistry. A modified crack velocity equation is proposed to include temperature, sulphur content, polarization potential, frequency and {Delta}K values and it is shown how the predictions compare with the proposed ASME XI revision. Critical fatigue situations are identified for 0.003% and 0.019% sulphur steels typical of modern and old plant. The use of the equation in assessing the synergistic effect of variables is discussed.

  7. 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.

  8. Reactor coolant pump startup under degraded conditions in a scaled OTSG lowered loop PWR

    SciTech Connect

    Tafreshi, A.M.; Marzo, M. di

    1996-12-31

    After a SB-LOCA or improper maintenance activities, the potential exists for a non-uniform distribution of boric acid in a PWR coolant system. This in turn presents the possibility of a reactivity excursion if sufficient volumes of boron-dilute water are transported into the core region without having first undergone substantial mixing. A research program is being conducted at the University of Maryland College Park (UMCP) 2 x 4 thermal-hydraulic test facility to assess the generation, transport and mixing of boron-dilute volumes. Start up of a pump and flow of a boron free slug of water in the cold leg and subsequent transport to the core downcomer in the facility is investigated here.

  9. Deposition of cobalt on surface-treated stainless steel under PWR conditions

    SciTech Connect

    Lister, D.H.; Anderson, P.G.; Barry, B.J.; Lavoie, R.G. . Chalk River Nuclear Labs.)

    1989-10-01

    As part of an on-going program aimed at reducing radiation exposures in light water reactors, the modification of surfaces to minimize their propensity to pick up radioactivity under reactor conditions has been studied. This report describes how stainless steel specimens, surface-treated with a variety of processes, picked up Co-60 from high-temperature water under PWR conditions in a high-pressure loop. The build-up of activity was monitored on-line with a movable gamma spectrometer. Off-line counting at the end of the experiment established the absolute activity levels, and selective examinations with SEM and metallography characterized the surface condition of the exceptional specimens. The effectiveness of the surface treatments was gauged by fitting simple parabolae to the activity build-up data and comparing the coefficients with those obtained from untreated control specimens. 10 refs., 23 figs., 4 tabs.

  10. 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.

  11. 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.

  12. 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

  13. Containment failure time and mode for a low-pressure short-term station blackout in a BWR-4 with Mark-I containment

    SciTech Connect

    Carbajo, J.J.; Greene, S.R. )

    1993-01-01

    This study investigates containment failure time and mode for a low-pressure, short-term station blackout severe accident sequence in a boiling water reactor (BWR-4) with a Mark-I containment. The severe accident analysis code MELCOR, version 1.8.1, was used in these calculations. Other results using the MELCOR/CORBH package and the BWRSAR and CONTAIN codes are also presented and compared to the MELCOR results. The plant analyzed is the Peach Bottom atomic station, a BWR-4 with a Mark-I containment. The automatic depressurization system was used to depressurize the vessel in accordance with the Emergency Procedure Guidelines. Two different variations of the station blackout were studied: one with a dry cavity and the other with a flooded cavity. For the flooded cavity, it is assumed that a control rod drive (CRD) pump becomes operational after vessel failure, and it is used to pump water into the cavity.

  14. 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).

  15. Criticality Safety and Sensitivity Analyses of PWR Spent Nuclear Fuel Repository Facilities

    SciTech Connect

    Maucec, Marko; Glumac, Bogdan

    2005-01-15

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based storage and dry transport containers under various loading patterns and moderating conditions. To comply with standard safety requirements, fresh 4.25% enriched nuclear fuel was assumed. The impact of potential optimum moderation due to water steam or foam formation as well as of different interpretations, of neutron multiplication through varying the system boundary conditions was elaborated. The simulations indicate that in the case of compact (all rack locations filled with fresh fuel) single or 'double tiering' loading, the supercriticality can occur under the conditions of enhanced neutron moderation, due to accidentally reduced density of cooling water. Under standard operational conditions the effective multiplication factor (k{sub eff}) of pool-based storage facility remains below the specified safety limit of 0.95. The nuclear safety requirements are fulfilled even when the fuel elements are arranged at a minimal distance, which can be initiated, for example, by an earthquake. The dry container in its recommended loading scheme with 26 fuel elements represents a safe alternative for the repository of fresh fuel. Even in the case of complete water flooding, the k{sub eff} remains below the specified safety level of 0.98. The criticality safety limit may however be exceeded with larger amounts of loaded fuel assemblies (i.e., 32). Additional Monte Carlo criticality safety analyses are scheduled to consider the 'burnup credit' of PWR spent nuclear fuel, based on the ongoing calculation of typical burnup activities.

  16. Initiation stress threshold irradiation assisted stress corrosion cracking criterion assessment for core internals in PWR environment

    SciTech Connect

    Tanguy, Benoit; Stern, Anthony; Bossis, Philippe; Pokor, Cedric

    2012-07-01

    Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors evolutions due to neutron irradiation. The potential contributors to IASCC in PWR primary water are: (i) radiation induced segregation (RIS) at grain boundaries, (ii) radiation induced microstructure (formation and growth of dislocations loops, voids, bubbles, phases), (iii) localized deformation under loading, (iv) irradiation creep and transmutations. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Parallel to fundamental understanding developments relative to IASCC, well controlled laboratory tests on neutron irradiated stainless steels are needed to assess the main mechanisms and also to establish an engineering criterion relative to the initiation of fracture due to IASCC. First part of this study describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop dedicated to SCC tests on neutron irradiated materials is then given. This autoclave recirculation loop has been started on July 2010 with the first SCC test on an irradiated stainless steel (grade 316) performed at CEA. The main steps of the interrupted SCC tests are then described. Second part of this paper reports the partial results of the first test performed on a highly neutron irradiated material. (authors)

  17. 3D Neutron Transport PWR Full-core Calculation with RMC code

    NASA Astrophysics Data System (ADS)

    Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien

    2014-06-01

    Nowadays, there are more and more interests in the use of Monte Carlo codes to calculate the detailed power density distributions in full-core reactors. With the Inspur TS1000 HPC Server of Tsinghua University, several calculations have been done based on the EDF 3D Neutron Transport PWR Full-core benchmark through large-scale parallelism. To investigate and compare the results of the deterministic method and Monte Carlo method, EDF R&D and Department of Engineering Physics of Tsinghua University are having a collaboration to make code to code verification. So in this paper, two codes are used. One is the code COCAGNE developed by the EDF R&D, a deterministic core code, and the other is the Monte Carlo code RMC developed by Department of Engineering Physics in Tsinghua University. First, the full-core model is described and a 26-group calculation was performed by these two codes using the same 26-group cross-section library provided by EDF R&D. Then the parallel and tally performance of RMC is discussed. RMC employs a novel algorithm which can cut down most of the communications. It can be seen clearly that the speedup ratio almost linearly increases with the nodes. Furthermore the cell-mapping method applied by RMC consumes little time to tally even millions of cells. The results of the codes COCAGNE and RMC are compared in three ways. The results of these two codes agree well with each other. It can be concluded that both COCAGNE and RMC are able to provide 3D-transport solutions associated with detailed power density distributions calculation in PWR full-core reactors. Finally, to investigate how many histories are needed to obtain a given standard deviation for a full 3D solution, the non-symmetrized condensed 2-group fluxes of RMC are discussed.

  18. Electrochemical behaviour of stainless steel in PWR primary coolant conditions: Effects of radiolysis

    NASA Astrophysics Data System (ADS)

    Muzeau, Benoist; Perrin, Stéphane; Corbel, Catherine; Simon, Dominique; Feron, Damien

    2011-12-01

    Few data are available in the literature on the role of the water radiolysis on the corrosion of stainless steel core components in PWR operating conditions (300 °C, 155 bar). The present approach uses a high energy proton beam to control the production of radiolytic species at the interface between a stainless steel sample and water in a high temperature and high pressure (HP-HT) electrochemical cell working in the range 25 °C/1 bar-300 °C/90 bar. The cell is designed to record the free corrosion potential of the AISI 316L/water interface mounted in line with a cyclotron delivering the proton beam. The evolution of the potential is compared before, during and after the proton irradiation. The first results are obtained with an aqueous solution containing boron, lithium and dissolved hydrogen, as in PWR primary coolant circuit. The stainless steel/water interfaces are irradiated between 25 °C and 300 °C with protons emerging at 22 MeV at the interface. The flux is varied by five orders of magnitude, from 6.6 × 10 11 to 6.6 × 10 15 H + m -2 s -1. The evolution of the free corrosion potential is highly dependent on the temperature and/or pressure. For a given temperature and pressure, it evolves with the flux and the ageing of the AISI 316L/water interfaces. An important role of the temperature of irradiation on the electrochemical response was observed. These results give a better understanding of the role of radiolysis on stainless steel corrosion in high temperature conditions.

  19. Single PWR spent fuel assembly heat transfer data for computer code evaluations

    SciTech Connect

    Bates, J.M.

    1986-01-01

    The descriptions and results of two separate heat transfer tests designed to investigate the dry storage of commercial PWR spent fuel assemblies are presented. Presented first are descriptions and selected results from the Fuel Temperature Test performed at the Engine Maintenance and Disassembly facility on the Nevada Test Site. An actual spent fuel assembly from the Turkey Point Unit Number 3 Reactor with a decay heat level of 1.17 KW, was installed vertically in a test stand mounted canister/liner assembly. The boundary temperatures were controlled and the canister backfill gases were alternated between air, helium and vacuum to investigate the primary heat transfer mechanisms of convection, conduction and radiation. The assembly temperature profiles were experimentally measured using installed thermocouple instrumentation. Also presented are the results from the Single Assembly Heat Transfer Test designed and fabricated by Allied General Nuclear Services, under contract to the Department of Energy, and ultimately conducted by the Pacific Northwest Laboratory. For this test, an electrically heated 15 x 15 rod assembly was used to model a single PWR spent fuel assembly. The electrically heated model fuel assembly permitted various ''decay heat'', levels to be tested; 1.0 KW and 0.5 KW were used for these tests. The model fuel assembly was positioned within a prototypic fuel tube and in turn placed within a double-walled sealed cask. The complete test assembly could be positioned at any desired orientation (horizontal, vertical, and 25/sup 0/ from horizontal for the present work) and backfilled as desired (air, helium, or vacuum). Tests were run for all combinations of ''decay heat,'' backfill, and orientation. Boundary conditions were imposed by temperature controlled guard heaters installed on the cask exterior surface.

  20. 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)

  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. 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.

  3. 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.

  4. 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.

  5. Comparison of Serpent and HELIOS-2 as applied for the PWR few-group cross section generation

    SciTech Connect

    Fridman, E.; Leppaenen, J.; Wemple, C.

    2013-07-01

    This paper discusses recent modifications to the Serpent Monte Carlo code methodology and related to the calculation of few-group diffusion coefficients and reflector discontinuity factors The new methods were assessed in the following manner. First, few-group homogenized cross sections calculated by Serpent for a reference PWR core were compared with those generated 1 commercial deterministic lattice transport code HELIOS-2. Second, Serpent and HELIOS-2 fe group cross section sets were later employed by nodal diffusion code DYN3D for the modeling the reference PWR core. Finally, the nodal diffusion results obtained using the both cross section sets were compared with the full core Serpent Monte Carlo solution. The test calculations show that Serpent can calculate the parameters required for nodal analyses similar to conventional deterministic lattice codes. (authors)

  6. Calculation of sample problems related to two-phase flow blowdown transients in pressure relief piping of a PWR pressurizer

    SciTech Connect

    Shin, Y.W.; Wiedermann, A.H.

    1984-02-01

    A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients.

  7. Nano-cavities observed in a 316SS PWR Flux Thimble Tube Irradiated to 33 and 70 dpa

    SciTech Connect

    Edwards, Danny J.; Garner, Francis A.; Bruemmer, Stephen M.; Efsing, Pal G.

    2009-02-28

    The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290ºC and 70 dpa at 315ºC were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during post-irradiation slow strain rate testing in PWR water conditions.

  8. Sample problem calculations related to two-phase flow transients in a PWR relief-piping network

    SciTech Connect

    Shin, Y.W.; Wiedermann, A.H.

    1981-03-01

    Two sample problems related with the fast transients of water/steam flow in the relief line of a PWR pressurizer were calculated with a network-flow analysis computer code STAC (System Transient-Flow Analysis Code). The sample problems were supplied by EPRI and are designed to test computer codes or computational methods to determine whether they have the basic capability to handle the important flow features present in a typical relief line of a PWR pressurizer. It was found necessary to implement into the STAC code a number of additional boundary conditions in order to calculate the sample problems. This includes the dynamics of the fluid interface that is treated as a moving boundary. This report describes the methodologies adopted for handling the newly implemented boundary conditions and the computational results of the two sample problems. In order to demonstrate the accuracies achieved in the STAC code results, analytical solutions are also obtained and used as a basis for comparison.

  9. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR

    SciTech Connect

    Not Available

    1982-12-01

    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. Standard Technical Specifications General Electric plants, BWR/4: Bases (Sections 2.0-3.3). Volume 2, Revision 1

    SciTech Connect

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved ST or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume I contains the Specifications for all chapters and sections of the improved STS. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4-3.10 of the improved STS.

  15. 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.

  16. 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)

  17. 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.

  18. Typical BWR/4 MSIV closure ATWS analysis using RAMONA-3B code with space-time neutron kinetics

    SciTech Connect

    Neymotin, L.; Saha, P.

    1984-01-01

    A best-estimate analysis of a typical BWR/4 MSIV closure ATWS has been performed using the RAMONA-3B code with three-dimensional neutron kinetics. All safety features, namely, the safety and relief valves, recirculation pump trip, high pressure safety injections and the standby liquid control system (boron injection), were assumed to work as designed. No other operator action was assumed. The results show a strong spatial dependence of reactor power during the transient. After the initial peak of pressure and reactor power, the reactor vessel pressure oscillated between the relief valve set points, and the reactor power oscillated between 20 to 50% of the steady state power until the hot shutdown condition was reached at approximately 1400 seconds. The suppression pool bulk water temperature at this time was predicted to be approx. 96/sup 0/C (205/sup 0/F). In view of code performance and reasonable computer running time, the RAMONA-3B code is recommended for further best-estimate analyses of ATWS-type events in BWRs.

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. 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)

  4. 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.

  5. 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.

  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. Large Break LOCA Safety Injection Sensitivity for a CE/ABB System 80+ PWR

    SciTech Connect

    Pottorf, J.; Bajorek, S.M.

    2002-07-01

    A WCOBRA/TRAC model of an evolutionary pressurized water reactor with direct vessel injection was constructed using publicly available information and a hypothetical double-ended guillotine break of a cold leg pipe was simulated. The model is an approximation of a ABB/Combustion Engineering System 80+ pressurized water reactor (PWR). WCOBRA/TRAC is the thermal-hydraulics code approved by the U.S. Nuclear Regulatory Commission for use in realistic large break LOCA analyses of Westinghouse 3- and 4-loop PWRs and the AP600 passive design. The AP600 design uses direct vessel injection, and the applicability of WCOBRA/TRAC to such designs is supported by comparisons to appropriate test data. This study extends the application of WCOBRA/TRAC to the investigation of the predicted behavior of direct vessel injection in an evolutionary design. A series of large break LOCA simulations were performed assuming a core power of 3914 MWt, and Technical Specification limits of 2.5 on total peaking factor and 1.7 on hot channel enthalpy rise factor. Two cladding temperature peaks were predicted during reflood, one following bottom of core recovery and a second caused by saturated boiling of water in the downcomer. This behavior is consistent with prior WCOBRA/TRAC calculations for some Westinghouse PWRs. The simulation results are described, and the sensitivity to failure assumptions for the safety injection system is presented. (authors)

  8. TRAB-3D/SMABRE Calculation of the OECD/NRC PWR MSLB Benchmark

    SciTech Connect

    Daavittila, A.; Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

    2001-06-17

    All three exercises of the OECD/NRC Pressurized Water Reactor (PWR) Main Steam Line Break (MSLB) Benchmark were calculated. The SMABRE thermal-hydraulics code was used for the first exercise, the plant simulation with point-kinetics neutronics. The second exercise was calculated with the TRAB-3D three-dimensional reactor dynamics code. The third exercise was calculated with the combination TRAB-3D/SMABRE. The results of all the exercises agree reasonably well with those of the other participants; therefore, instead of reporting results, this paper concentrates on describing the computational aspects of the calculation with the above-mentioned codes and on some observations of the sensitivity of the results. The variations calculated with SMABRE with modifications in the upper head, steam generators, and steam lines affect mainly the time of recriticality. During the fourth workshop of the benchmark, a decision was made to extrapolate the cross sections if the fuel temperature or moderator density was out of the range of the given cross section tables. In the TRAB-3D calculation, this extrapolation made a significant difference for the first scenario; there is a low power maximum after the scram, which is not seen in the calculation without the extrapolation.

  9. 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.

  10. 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)

  11. NDE and mechanical removal of sludge in PWR steam generators: Volume 2, Vendor practices: Final report

    SciTech Connect

    Kidd, C.C.; Scharton, T.D.; Spencer, R.B.; Taylor, G.B.; Stewart, D.R.; Gallagher, M.J.; Johnson, L.E.; Sapia, M.A.; Edwards, L.J.; Dashukewich, M.L.

    1988-01-01

    A study was made to identify the needs of utilities for detecting, measuring, and mechanically removing sludge and related corrosion products from PWR steam generators, both recirculating U-tube and once through designs. The study determining, from the utility-user viewpoint, how well these needs are being met by currently available technology; identified opportunities for improvement; and made recommendations for research efforts to realize these opportunities. Methods for chemically removing sludge and corrosion products from steam generators, i.e., use of chemical solvents, were not addressed. Reports from nuclear steam supply system vendors and independent service vendors on their current processes and prior developmental efforts to realize these opportunities. Methods for chemically removing sludge and corrosion products from steam generators, i.e., use of chemical solvents, were not addressed. Reports from nuclear steam supply system vendors and independent service vendors on their current processes and prior developmental efforts with mechanical removal methods and NDE techniques are included in the study. In addition, information was obtained from the technical literature and from discussions and visits with knowledgeable individuals at utilities, service vendors, and engineering and consulting firms. Current removal methods examined included sludge lancing, pressure pulse and water slap; current NDE techniques examined included eddy current, optical instruments, sludge sampling, and water balance measurements. Additional NDE techniques reported on by the service vendors included Hall effect and magnetic field sensing probes, ultrasonic, and radiation attenuation techniques.

  12. 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)

  13. {sup 252}Cf-source-driven frequency analysis measurements with subcritical arrays of PWR fuel pins

    SciTech Connect

    Mihalczo, J.T.; Valentine, T.E.; Blakeman, E.D.; King, W.T.

    1996-08-01

    Experiments with fresh PWR fuel assemblies were performed to assess the {sup 252}Cf-source-driven frequency analysis method for measuring the subcriticality of spent fuel. The measurements at the Babcox and Wilcox Critical Experiments Facility mocked up between 17x17 fuel pins (single assembly) and a full array of 4961 fuel pins (about 17 fuel assemblies) in borated water with a fixed B concentration. For the full array, the B content of the water was varied from 1511 at delayed criticality to 4303 ppM. Measurements were done for various source-detector-fuel pin configurations; they showed high sensitivity of frequency analysis parameters to B content and fissile mass. Parameters such as auto and cross power spectral densities can be calculated directly by a more general model of the Monte Carlo code (MCNP-DSP). Calculation-measurement comparisons are presented. This model permits the validation of neutron and gamma ray transport calculational methods with subcritical measurements using the {sup 252}Cf-source-driven frequency analysis method.

  14. 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.

  15. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    NASA Astrophysics Data System (ADS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  16. Library of PWR (pressurized-water reactor) steam generator tubing samples: Final report

    SciTech Connect

    Albertin, L.; Clark, W.G. Jr.; Junker, W.R.; Kuchirka, P.J.; Madeyski, A.; Metala, M.J.; Taszarek, B.J.

    1988-01-01

    The PWR Steam Generator Tubing Sample Library is a Steam Generator Owners Group-EPRI program whose objective is to compile a library of well-characterized tubing samples to be used for performance evaluation of inspection systems and for training and qualification of signal interpretation systems. The library was created through the preparation of samples intended to replicate degradation encountered in actual field tubes. A limited number of tube segments removed from actual steam generators are included. Degradation categories include wear, pitting and fatigue cracks, as well as stress corrosion cracking (SCC) and intergranular attack (IGA). Eddy current and ultrasonic inspection techniques, along with supplementary radiography, dye penetrant, and optical techniques were used to characterize the library candidates. Advanced computer-aided NDE data collection, analysis and display techniques were used to assess test results. This report provides details of the library program, with major emphasis on the sampling protocol, characterization of degradation and recommendations for the use and future growth of the library. Also included is a compendium of steam generator tube degradation field observation, describing past destructive examinations of tubes removed for inspection from steam generators, and a description of a physical modeling approach, using mercury (metal) to assess the discontinuity characterization capabilities of a pancake-type eddy current probe. Computerized data analysis and display techniques were used to reconstruct the test results in both two-dimensional color-coded maps and three-dimensional pseudo-isometric plots.

  17. 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.

  18. The stress corrosion cracking behavior of alloys 690 and 152 WELD in a PWR environment.

    SciTech Connect

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

    2009-01-01

    Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and 182, respectively. The latter two alloys are used as structural materials in pressurized water reactors (PWRs) and have been found to undergo stress corrosion cracking (SCC). The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for the replacement alloys. The study involved Alloy 690 cold-rolled by 26% and a laboratory-prepared Alloy 152 double-J weld in the as-welded condition. The experimental approach involved pre-cracking in a primary water environment and monitoring the cyclic CGRs to determine the optimum conditions for transitioning from the fatigue transgranular to intergranular SCC fracture mode. The cyclic CGRs of cold-rolled Alloy 690 showed significant environmental enhancement, while those for Alloy 152 were minimal. Both materials exhibited SCC of 10{sup -11} m/s under constant loading at moderate stress intensity factors. The paper also presents tensile property data for Alloy 690TT and Alloy 152 weld in the temperature range 25--870 C.

  19. 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.

  20. Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.

    Energy Science and Technology Software Center (ESTSC)

    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

  1. 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.

  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. )

    1992-01-01

    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.

  3. 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.

  4. Interpretation of the XR2-1 experiment and characteristics of the BWR lower plenum debris bed

    SciTech Connect

    Hodge, S.A.; Ott, L.J.

    1997-11-01

    The Ex-Reactor (XR) experiments have been conducted to advance the understanding of BWR severe accident melt progression events. The XR2-1 experiment addresses the fate of the initial large (code-predicted) movements of molten metals from the upper core to the lower core and core plate region. For this question, which has ramifications for blockage formation in the core region, the XR2-1 test results provide significant and perhaps definitive insights. Nevertheless, some events that occurred during this test are creatures of the special features of the test apparatus, and there is a potential for misconceptions with respect to the direct applicability of some of the results. This paper describes the conclusions that can be drawn from the XR2-1 experiment results and identifies those areas (such as fuel pellet stack collapse and core plate integrity) where care must be taken not to misconstrue the test events. Another important area where much recent work has been performed is the effort to analyze the potential for maintaining core debris within the reactor vessel lower plenum by cooling of the outer vessel wall. One of the first steps in such an analytical endeavor is to attempt to establish the pattern of energy transfers into the wall inner surface. As a prerequisite to determination of this pattern, it is necessary to first consider the nature of the debris within the lower plenum. Too often is an easily represented homogeneous circulating liquid pool incorporated without adequate consideration of the true material conditions. Basic considerations of the relative quantities of materials present, the potentials for eutectics formations, and the associated melting points dictate otherwise. This paper offers some insights as to the true nature of the lower plenum debris and discusses the need for some relatively simple experiments that would contribute much toward the basic understanding necessary for accurate debris characterization. 11 refs., 4 figs., 4 tabs.

  5. 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.

  6. 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)

  7. 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.

  8. Severe accident modeling of a PWR core with different cladding materials

    SciTech Connect

    Johnson, S. C.; Henry, R. E.; Paik, C. Y.

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  9. 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.

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. COMMIX-1A analysis of fluid and thermal mixing in a model cold leg and downcomer of a PWR

    SciTech Connect

    Chen, B.C.J.; Cha, B.K.; Sha, W.T.

    1984-06-01

    Fluid and thermal mixing in a model cold leg and downcomer of a PWR was analyzed using COMMIX-1A. The present analysis differs from previous analyses reported in EPRI NP-3321 in three major aspects. First, extremely fine meshes were used to minimize numerical diffusion in the analysis. Second, one-equation (k) turbulence model was used to better model the turbulent flow. Third, curved surfaces were modeled by several slanted planes to better represent the geometries. By using these improvements, CREARE 1/5-scale test No. 51 was reanalyzed. Significant improvements were achieved in the comparisons between the COMMIX-1A calculations and the experimental data.

  16. 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.

  17. 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.

  18. 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.

  19. COMMIX-1A analysis of fluid and thermal mixing in a model cold leg and downcomer of a PWR

    SciTech Connect

    Chen, B.C.J.; Cha, B.K.; Miao, C.C.; Sha, W.T.; Kim, J.H.; Sun, B.K.H.

    1983-01-01

    The issue of thermal shock of a PWR pressure vessel has been under considerable attention recently. A number of experimental as well as analytical studies have been performed to investigate the effect of the thermal transient on the pressure vessel due to the high pressure injection (HPI) of the cold fluid into the cold leg. This process has been called Pressurized Thermal Shock (PTS). This paper is an analytical study of PTS by using COMMIX-1A. Experimental investigations were performed at CREARE and SAI. In the CREARE experiment, a 1/5 scale model was set up to simulate a cold leg and downcomer of a PWR. Tests with several different ratios of hot loop flow versus cold HPI flow were performed to study the effect of the flow ratio on the fluid and thermal mixing process in the system, especially in the downcomer region. Analytical investigations also proceeded in parallel with the experiments. Quite a few analytical investigations were performed with the COMMIX-1A code. However, in this version of COMMIX, the effect of the numerical diffusion was not addressed.

  20. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Steady state and transient computations

    SciTech Connect

    Martin, A.; Alvarez, D.; Cases, F.

    1996-06-01

    After the Tchernobyl accident a working group was created to analyze the French PWR Safety with a respect to potential risk of reactivity accident. Potentially risky situations are those which can lead to heterogeneous boron concentration or temperature of the primary coolant fluid. This paper reports a Research and Development action based on numerical simulations and experiments on the primary coolant temperature or boron mixing capabilities in a PWR vessel. New numerical results obtained with the thermal hydraulic Finite Element (FE) Code N3S are presented. In these calculations the FE mesh takes into account the geometry of the lower plenum plates and columns. Two configurations have been investigated The first one is a steady state fluid flow mixing case. The reactor is cooled by free convection and the three loops, balanced in mass flow rate, are in operation. The second is a free boron plug transient case. It is related to the mixing of a clear plug injected in the vessel when a primary coolant pump starts-up. Two clear plug volumes have been investigated (3 and 8 m{sup 3}). Comparisons between these new numerical results and the data previously obtained (see Alvarez et al., 1992, Alvarez, Martin and Schneider, 1994) are presented in this paper.

  1. 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.

  2. A review of the (60)Co internal dosimetry at Devonport Royal Dockyard.

    PubMed

    Vickers, J M A; Collison, R; Collision, R

    2010-03-01

    The physico-chemical properties of (60)Co contaminants arising from the UK Naval Nuclear Propulsion Programme (NNPP) pressurised water reactor (PWR) plants have been investigated in order to review individual monitoring requirements at Devonport Royal Dockyard (DRD). This has been achieved through laboratory tests on NNPP primary component samples and interpretation of direct bioassay measurements using internal dosimetry modelling software. Interpretation of lung measurements was completed for two inhalation events involving material originating from a PWR plant and post-primary circuit decontamination. Initial estimates of intake and dose were calculated using International Commission on Radiological Protection default parameter values. However, a good fit could only be achieved by fitting the data to alternative absorption parameters where 90-95% of the material dissolved and absorbed rapidly at a rate of 1 day(-1). As a consequence of this review, a number of improvements have been made to monitoring arrangements at DRD. A minimum of three direct measurements are now taken during the 0-30 day period after an intake, the capability of the Canberra Accuscan has been enhanced and dissolution tests are being carried out by the Health Protection Agency (HPA) on samples taken from PWR plants. PMID:20220215

  3. 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

  4. 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

  5. High mechanical performance of Areva upgraded fuel assemblies for PWR in USA

    SciTech Connect

    Gottuso, Dennis; Canat, Jean-Noel; Mollard, Pierre

    2007-07-01

    The merger of the product portfolios of the former Siemens and Framatome fuel businesses gave rise to a new family of PWR products which combine the best features of the different technologies to enhance the main performance of each of the existing products. In this way, the technology of each of the three main fuel assembly types usually delivered by AREVA NP, namely Mark-BW{sup TM}, HTP{sup TM} and AFA 3G{sup TM} has been enriched by one or several components from the others which contributes to improve their robustness and to enhance their performance. The combined experience of AREVA's products shows that the ROBUST FUELGUARD{sup TM}, the HMP{sup TM} end grid, the MONOBLOC{sup TM} guide tube, a welded structure, M5{sup R} material for every zirconium component and an upper QUICK-DISCONNECT{sup TM} are key features for boosting fuel assembly robustness. The ROBUST FUELGUARD benefits from a broad experience demonstrating its high efficiency in stopping debris. In addition, its mechanical strength has been enhanced and the proven blade design homogenizes the downstream flow distribution to strongly reduce excitation of fuel rods. The resistance to rod-to-grid fretting resistance of AREVA's new products is completed by the use of a lower HMP grid with 8 lines of contact to insure low wear. The Monobloc guide tube with a diameter maximized to strengthen the fuel assembly stiffness, excludes through its uniform outer geometry any local condition which could weaken guide tube straightness. The application of a welded cage to all fuel assemblies of the new family of products in combination with stiffer guide tubes and optimized hold-down assures each fuel assembly enhanced resistance to distortion. The combination of these features has been widely demonstrated as an effective method to reduce the risk of incomplete RCCA insertion and significantly reduce assembly distortion. Thanks to its enhanced performance, M5 alloy insures that all fuel assemblies in the family

  6. 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.

  7. Identification of ryanodine receptor isoforms in prostate DU-145, LNCaP, and PWR-1E cells.

    PubMed

    Kobylewski, Sarah E; Henderson, Kimberly A; Eckhert, Curtis D

    2012-08-24

    The ryanodine receptor (RyR) is a large, intracellular calcium (Ca(2+)) channel that is associated with several accessory proteins and is an important component of a cell's ability to respond to changes in the environment. Three isoforms of the RyR exist and are well documented for skeletal and cardiac muscle and the brain, but the isoforms in non-excitable cells are poorly understood. The aggressiveness of breast cancers in women has been positively correlated with the expression of the RyR in breast tumor tissue, but it is unknown if this is limited to specific isoforms. Identification and characterization of RyRs in cancer models is important in understanding the role of the RyR channel complex in cancer and as a potential therapeutic target. The objective of this report was to identify the RyR isoforms expressed in widely used prostate cancer cell lines, DU-145 and LNCaP, and the non-tumorigenic prostate cell line, PWR-1E. Oligonucleotide primers specific for each isoform were used in semi-quantitative and real-time PCR to determine the identification and expression levels of the RyR isoforms. RyR1 was expressed in the highest amount in DU-145 tumor cells, expression was 0.48-fold in the non-tumor cell line PWR-1E compared to DU-145 cells, and no expression was observed in LNCaP tumor cells. DU-145 cells had the lowest expression of RyR2. The expression was 26- and 15-fold higher in LNCaP and PWR-1E cells, respectively. RyR3 expression was not observed in any of the cell lines. All cell types released Ca(2+) in response to caffeine showing they had functional RyRs. Total cellular RyR-associated Ca(2+) release is determined by both the number of activated RyRs and its accessory proteins which modulate the receptor. Our results suggest that the correlation between the expression of the RyR and tumor aggression is not related to specific RyR isoforms, but may be related to the activity and number of receptors. PMID:22846571

  8. 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.

  9. Experimental validation of CASMO-4E and CASMO-5M for radial fission rate distributions in a westinghouse SVEA-96 Optima2 BWR fuel assembly

    SciTech Connect

    Grimm, P.; Perret, G.

    2012-07-01

    Measured and calculated radial total fission rate distributions are compared for the three axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising 96, 92 and 84 fuel rods, respectively. The measurements were performed on a full-size fuel assembly in the PROTEUS zero-power experimental facility. The measured fission rates are compared to the results of the CASMO-4E and CASMO-5M fuel assembly codes. Detailed measured geometrical data were used in the models, and effects of the surrounding zones of the reactor were taken into account by correction factors derived from MCNPX calculations. The results of the calculations agree well with those of the experiments, with root-mean-square deviations between 1.2% and 1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4E and CASMO-5M is comparable. (authors)

  10. 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

  11. 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

  12. 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.

  13. Probabilistic assessment of the primary-coolant-loop pipe-fracture due to fatigue crack growth for a PWR plant

    SciTech Connect

    Chou, C.K.

    1981-06-01

    The work reported herein assesses the probability of a double-ended guillotine break of the hot leg, cold leg and cross-over line (for the purpose of this paper we defined it as a large LOCA) of a PWR plant subjected to the loads caused by plant transients and earthquakes. The work employs a fracture mechanics based fatigue model to propagate cracks from an initial flaw distribution. Flaw size and aspect ratio, material properties, operating transient and seismic stress histories, pre-service and in-service inspections as well as leak defections are considered random variables to be input into the fatigue crack growth fracture mechanics model. A brief description of the model and interrelationship between various steps are discussed.

  14. 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.

  15. Analysis of the performance of the Westinghouse reactor vessel level indicating system for tests at semiscale. [PWR

    SciTech Connect

    Hardy, J.E.; Miller, G.N.

    1982-10-01

    The Westinghouse Reactor Vessel Level Indicating System (RVLIS), a differential pressure level measurement system, was tested at SEMISCALE. This report contains the analyses of these tests and the conclusions of these analyses. The tests performed included small break and intermediate break tests. Also, frequency response and natural circulation tests were run and analyzed. The RVLIS always indicated a level less than the two phase froth level. The RVLIS output in early small break tests indicated a level 200 cm greater than actual collapsed liquid level. This discrepancy was caused by structural differences between SEMISCALE and a Westinghouse reactor. Once modifications were made so that SEMISCALE better simulated a Westinghouse PWR, the maximum difference between RVLIS and SEMISCALE instrumentation was 30 cm or 3% which is less than the stated uncertainty of the Westinghouse RVLIS.

  16. 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.

  17. Modeling the activity of 129I and 137Cs in the primary coolant and CVCS resin of an operating PWR

    NASA Astrophysics Data System (ADS)

    Hwang, K. H.; Lee, K. J.

    2006-04-01

    Mathematical models have been developed to describe the activities of 129I and 137Cs in the primary coolant and resin of the chemical and volume control system (CVCS) during constant power operation in a pressurized water reactor (PWR). The models, which account for the source releases from defective fuel rod(s) and tramp uranium, rely on the contribution of CVCS resin and boron recovery system as a removal process, and differences in behavior for each nuclide. The current models were validated through measured coolant activities of 137Cs. The resultant scaling factors agree reasonably well with the results of the test resin of the coolant and the actual resins from the PWRs of other countries.

  18. 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.

  19. 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).

  20. 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)