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Sample records for stone-webster reference pwr

  1. 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).

  2. 75 FR 57532 - In the Matter of: Stone & Webster Construction, Inc.; Confirmatory Order (Effective Immediately)

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-21

    ..., practices and programs that support Safety Conscious Work Environment (SCWE) and Safety Culture: At the... Culture & Safety Conscious Work Environment, modeled on NEI 09-12 and RIS 2005- 018, describes...

  3. Subchannel analysis of multiple CHF events. [PWR; BWR

    SciTech Connect

    Reddy, D.G.; Fighetti, C.F.

    1982-08-01

    The phenomenon of multiple CHF events in rod bundle heat transfer tests, referring to the occurrence of CHF on more than one rod or at more than one location on one rod is examined. The adequacy of some of the subchannel CHF correlations presently used in the nuclear industry in predicting higher order CHF events is ascertained based on local coolant conditions obtained with the COBRA IIIC subchannel code. The rod bundle CHF data obtained at the Heat Transfer Research Facility of Columbia University are examined for multiple CHF events using a combination of statistical analyses and parametric studies. The above analyses are applied to the study of three data sets of tests simulating both PWR and BWR reactor cores with uniform and non-uniform axial heat flux distributions. The CHF correlations employed in this study include: (1) CE-1 correlation, (2) B and W-2 correlation, (3) W-3 correlation, and (4) Columbia correlation.

  4. Horizontal Drop of 21- PWR Waste Package

    SciTech Connect

    A.K. Scheider

    2007-01-31

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

  5. Improving fuel-rod performance. [PWR; BWR

    SciTech Connect

    Ocken, H.; Knott, S.

    1981-03-01

    To reduce the risk of fuel-rod failures, utilities operate their nuclear reactors within conservative limits on power increases proposed by nuclear-fuel vendors. Of particular concern to US utilities is that adopting these limits results in an industrywide average plant capacity loss of 3% in BWR designs and 0.3% in PWR designs. To replace lost BWR capacity by other generating means currently costs the utilities $150 million annually, and losses for PWRs are about $20 million. Efforts are therefore being made to identify the factors responsible for Zircaloy degradation under PCI condition and to improve nuclear-fuel-rod design and reactor operation.

  6. Cavern/Vault Disposal Concepts and Thermal Calculations for Direct Disposal of 37-PWR Size DPCs

    SciTech Connect

    Hardin, Ernest; Hadgu, Teklu; Clayton, Daniel James

    2015-03-01

    This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

  7. Crevice chemistry control in PWR steam generators

    SciTech Connect

    Sawochka, S.G.; Choi, S.S.; Millett, P.J.; Bates, J.; Gardner, J.

    1995-12-31

    To establish a basis for predicting and eventually controlling crevice solution chemistry in PWR steam generators, hideout tests were performed at several units. Results indicated that impurity hideout rates varied with the species and with bulk water concentration. Field evaluations of crevice impurity inventory models based on the hideout rate data indicated that further model refinements were necessary, e.g., more frequent quantification of the relation of hideout rates and bulk water concentration. An alternate crevice inventory model based on a real-time mass balance approach also began to be pursued. Modeling results currently are being used at several PWRs to establish a chloride injection rate consistent with development of a near neutral crevice solution to minimize IGA/SCC. Hideout return data are being used to independently establish predictions of crevice chemistry and to substantiate the hideout rate and mass balance model predictions.

  8. PWR representative behavior during a LOCA

    SciTech Connect

    Allison, C.M.

    1981-01-01

    To date, there has been substantial analytical and experimental effort to define the margins between design basis loss-of-coolant accident (LOCA) behavior and regulatory limits on maximum fuel rod cladding temperature and deformation. As a result, there is extensive documentation on the modeling of fuel rod behavior in test reactors and design basis LOCA's. However, modeling of that behavior using representative, non-conservative, operating histories is not nearly as well documented in the public literature. Therefore, the objective of this paper is (a) to present calculations of LOCA induced behavior for Pressurized Water Reactor (PWR) core representative fuel rods, and (b) to discuss the variability in those calculations given the variability in fuel rod condition at the initiation of the LOCA. This analysis was limited to the study of changes in fuel rod behavior due to different power operating histories. The other two important parameters which affect that behavior, initial fuel rod design and LOCA coolant conditions were held invarient for all of the representative rods analyzed.

  9. 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)

  10. A PWR Thorium Pin Cell Burnup Benchmark

    SciTech Connect

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

    2000-05-01

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

  11. Impact of boron dilution accidents on low boron PWR safety

    SciTech Connect

    Papukchiev, A.; Liu, Y.; Schaefer, A.

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As an inadvertent reduction of the boron concentration during a boron dilution accident could introduce positive reactivity and have a negative impact on PWR safety, design changes to reduce boron concentration in the reactor coolant are of general interest. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) load has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) to 518 ppm. For the assessment of the potential safety advantages, a boron dilution accident due to small break loss-of-coolant-accident (SBLOCA) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The results from the comparative analyses showed that the impact of the boron dilution accident on the new PWR design safety is significantly lower in comparison with the standard design. The new reactor design provided at least 4, 4% higher reactivity margin to recriticality during the whole accident which is equivalent to the negative reactivity worth of additional 63% of all control rods fully inserted in to the core. (authors)

  12. 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)

  13. 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)

  14. Design of Recycle PWR with Heavy Water Moderation

    SciTech Connect

    Hibi, K.; Uchita, M.

    2002-07-01

    This study shows the conceptual plant design of the recycle PWR (RPWR), which is an innovative MOX-PWR with breeding ratios around 1.1 moderated by heavy water. Most of the plant systems of RPWR can employ the systems of PWRs. RPWR has no acid boron systems and has a small tritium removal system. The construction and operation costs are similar to the current PWRs. While, heavy water cost will be decreased drastically with up-to-date producing methods. The reliability for the plant systems of RPWR is high and R and D cost for realizing RPWR is very low because the core design of RPWR is fundamentally based on the current PWR technology. (authors)

  15. 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.

  16. PWR fuel features to preclude externally induced damage

    SciTech Connect

    Shallenberger, J.M.; Wilson, J.F.; Knott, R.P.

    1987-01-01

    Over the past several years there have been instances of pressurized water reactor (PWR) fuel damage attributed to factors external to the fuel. These externally induced causes include debris in the reactor coolant and baffle jetting. These causes of PWR fuel damage account for --50% of the total number of damaged rods. This paper discusses two features that significantly reduce the potential for fuel damage due to debris and baffle jetting. These two features are the debris filter bottom nozzle (DFBN) and the antivibration clip.

  17. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect

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

    2000-08-21

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

  18. 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.

  19. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect

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

    1997-12-01

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

  20. Update on the PWR axial burnup profile database

    SciTech Connect

    Cacciapouti, R.F.; Volkinburg, S.V.

    1995-12-01

    A pressurized water reactor database was developed to evaluate the axial burnup profiles of various reactor types. The data showed that the various types exhibit similar behavior, especially at the top and bottom of the assembly. From the existing data, bounding axial burnup profiles can be developed to envelope the various pressurized water reactor assembly deigns. The database encompasses most of the PWR fuel designs and contains sufficient data to provide reliable statistics.

  1. 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.

  2. 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.

  3. 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.

  4. Evaluation of zinc addition to PWR primary coolant

    SciTech Connect

    Pathania, R.; Yagnik, S.; Gold, R.E.; Dove, M.; Kolstad, E.

    1995-12-31

    Laboratory studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. Because of the potential benefits of zinc addition in reducing radiation fields and mitigating PWSCC of Alloy 600 a project was initiated to qualify zinc addition to a PWR. The objective of this work was to evaluate the effect of zinc addition on radiation fields, PWSCC of Alloy 600 and fuel cladding corrosion at the Farley-2 PWR. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor prior to zinc addition at Farley-2. This paper provides an overview of the scope of the zinc addition demonstration at Farley-2 and the fuel cladding corrosion tests at Halden. The zinc concentration in the Farley-2 coolant is approximately 40 ppb and that in Halden is 50 ppb. The paper presents initial results from these studies which are still in progress.

  5. 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

  6. 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.

  7. 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

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

    NASA Astrophysics Data System (ADS)

    Kang, Rixin; Zhang, Shucheng; Chen, Dianshan

    1991-02-01

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

  9. 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.

  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. 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.

  12. 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

  13. 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.

  14. 103. PWR2 CORE SUPPORT FLANGE BEING SEATED ON REACTOR VESSEL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    103. PWR-2 CORE SUPPORT FLANGE BEING SEATED ON REACTOR VESSEL FLANGE, APRIL 14, 1964 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

  15. Concept of Small Sized Integrated PWR with Double Pressure Vessels

    SciTech Connect

    Kinoshita, I.; Ueda, N.; Nishi, Y.; Matsumura, T.

    2002-07-01

    For early deployment of small sized nuclear reactors, it is better to reduce the BOP cost with new ideas than introducing innovative technologies for core, fuel and materials. In this report, a concept of the integrated, forced convective and small PWR with double pressure vessels has been proposed. The electric output of this reactor is 150 MW. Conventional technologies are adopted for core and fuel. Refueling, maintenance and repairing are made in a special ship with complete facilities and skilled experts. The pressure vessel with the core, control rod drive mechanisms (CRDM), main circulating pumps (MCP), steam generators (SG) and other reactor internals are transferred between the reactor building and the ship. Technical feasibility for safety and maintainability has been discussed qualitatively. The construction cost has been roughly estimated. (authors)

  16. 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.

  17. 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.

  18. 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.

  19. PWR and BWR spent fuel assembly gamma spectra measurements

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  20. PWR and BWR spent fuel assembly gamma spectra measurements

    SciTech Connect

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

    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) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  1. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGES

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

    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. 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.

  3. 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.

  4. 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

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

    SciTech Connect

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

    1993-04-01

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

  6. A Feasibility Study of an Integral PWR for Space Applications

    SciTech Connect

    Grandis, S. De; Finzi, E.; Lombardi, C.V.; Mandelli, D.; Padovani, E.; Passoni, M.; Ricotti, M.E.; Santini, L.

    2004-07-01

    Fission space power systems are well suited to provide safe, reliable, economic and robust energy sources, in the order of 100 KWe. A preliminary feasibility study of a nuclear fission reactor is here presented with the following requirements: i) high reliability, ii) R and D program of moderate cost, iii) to be deployed within a reasonable period of time (e.g. 2015), iv) to be operated and controlled for a long time (10 years) without human intervention, v) possibly to be also used as a byproduct for some particular terrestrial application (or at least to share common technologies), vi) to start with stationary application. The driving idea is to extend as much as possible the PWR technology, by recurring to an integral type reactor. Two options are evaluated for the electricity production: a Rankine steam cycle and a Rankine organic fluid cycle. The neutronics calculation is based on WIMS code benchmarked with MCNP code. The reactivity control is envisaged by changing the core geometry. The resulting system appears viable and of reasonable size, well fit to the present space vector capabilities. Finally, a set of R and D needs has been identified: cold well, small steam turbines, fluid leakage control, pumps, shielding, steam generator in low-gravity conditions, self pressurizer, control system. A R and D program of reasonable extent may yield the needed answers, and some demanding researches are of interest for the new generation Light Water Reactors. (authors)

  7. Effect of component aging on PWR control rod drive systems

    SciTech Connect

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-01-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock Wilcox (B W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging.

  8. Effect of component aging on PWR control rod drive systems

    SciTech Connect

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-06-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock & Wilcox (B & W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging.

  9. Evaluation of surface modification techniques for PWR steam generator channel heads. Final report

    SciTech Connect

    Spalaris, C.N.

    1986-06-01

    Surface modification which were developed under a previous EPRI program and then applied to Boiling Water Reactor replacement piping, were modified for treating PWR steam generator channel head surfaces. Surface modifications have been shown to reduce out-of-core activity build up in BWR and thought to be equally effective in PWR circuits as well. Prototypical surface test specimens were used to develop techniques appropriate to PWR alloy substrates which were then applied to treat the surfaces of a spare, full size PWR channel head in a field demonstration. Modified surfaces cut from test specimens and pieces removed from the field demonstration were submitted to metallurgical investigations. No damage to the substrate alloys was detected as a result of the surface modification processes. Combination of mechanical and electropolishing action improved the as fabricated finish by at least a factor of 3 for the Inconel plate and factors of 20 for the stainless weld overlay. Field demonstration yielded a factor of 10 improvement in the weld overlay and 30 to 40% in the divider plate. Because these surfaces are known to be responsible for 57% of the area radioactivity in PWR steam generators in service, prepolishing is expected to reduce radiation fields substantially. 31 figs.

  10. 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.

  11. 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.

  12. 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)

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

    SciTech Connect

    J.W. Davis

    1996-07-15

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

  14. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  15. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  16. 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

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

    SciTech Connect

    T.L. Lotz

    1997-01-29

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

  20. 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

  1. 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.

  2. Application of the RCP01 Code to Depletion of a PWR Spent Nuclear Fuel Sample

    SciTech Connect

    Joo, Hansem

    2002-01-01

    An essential component of a proposed burnup credit methodology for commercial PWR spent nuclear fuel (SNF) is the validation of the tools used for isotopic and criticality calculations. A number of benchmark experiments have been analyzed to establish the validation of the tools and to determine biases and corrections. To benchmark the RCP01 Monte Carlo computer code, an isotopic validation study was conducted for one of the benchmark experiments, a SNF sample taken from the Calvert Cliffs PWR Unit-1 (CCPU1). Modeling considerations and nuclear data associated with the RCP01 transport/depletion calculations are discussed. The accuracy of RCP01 calculations is demonstrated to be very good when RCP01 results are compared to destructive chemical assay data for major actinides and important fission products in the SNF sample.

  3. 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.

  4. An Empirical Approach to Bounding the Axial Reactivity Effects of PWR Spent Nuclear Fuel

    SciTech Connect

    P. M. O'Leary; J. M. Scaglione

    2001-04-04

    One of the significant issues yet to be resolved for using burnup credit (BUC) for spent nuclear fuel (SNF) is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters (such as local power, fuel temperature, moderator temperature, burnable poison rod history, and soluble boron concentration) affect the isotopic inventory of fuel that is depleted in a pressurized water reactor (PWR). However, obtaining the detailed operating histories needed to model all PWR fuel assemblies to which BUC would be applied is an onerous and costly task. Simplifications therefore have been suggested that could lead to using ''bounding'' depletion parameters that could be broadly applied to different fuel assemblies. This paper presents a method for determining a set of bounding depletion parameters for use in criticality analyses for SNF.

  5. 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.

  6. 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)

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Generation and behavior of metal oxide colloids in PWR steam systems

    SciTech Connect

    Varsanik, R.G.

    1984-10-01

    This work reviews the curently available literature and research work on the generation and behavior of metal oxide colloids in PWR steam systems. The work of E. Matijevic et al on the generation and adhesion of iron and copper oxides is described. The role of colloid chemistry in the control of plant sludge and corrosion products is described. Factors affecting the adherence and re-entrainment of colloidal metal oxides along with possible methods for the control of metal oxide deposition are reviewed.

  13. 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)

  14. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    SciTech Connect

    Hori, Keiichi; Miyazaki, Keiji; Akiyama, Yoshiei; Nishioka, Hiromasa; Takeda, Naoki

    1996-08-01

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method.

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

    SciTech Connect

    Naegeli, Robert Earl

    2004-06-01

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

  16. Determination of soluble chromium in simulated PWR coolant by differential-pulse adsorptive stripping voltammetry.

    PubMed

    Torrance, K; Gatford, C

    1987-11-01

    An analytical method has been developed for the determination of dissolved chromium at concentrations less than 2 mug/l. in PWR coolant by differential-pulse adsorptive stripping voltammetry at a hanging mercury drop electrode. Concentrations above 2 mug/l. can be determined by appropriate dilution of the sample. The method is based on measurement of the current associated with reduction of a chromium(III)-DTPA (diethylenetriaminepenta-acetic acid) complex adsorbed at the surface of the mercury drop. The effects of boric acid, pH, DTPA concentration, accumulation potential and time were investigated together with the oxidation state of the chromium. No interference was observed from other transition metal ions expected to be present in PWR coolant. No alternative chemical technique of similar sensitivity was available for comparison with the results obtained in solutions containing <1 mug/l. chromium. Recoveries from simulated coolant solutions were greater than 95% and the relative standard deviations for single determinations were in the range 12-25%. The statistical limit of detection at the 95% confidence level was 0.023 mug/l. This method of analysis should prove valuable in corrosion studies and is uniquely capable of following the changes in soluble chromium concentration in PWR coolant that follow operational changes in the reactor.

  17. Differential pulse stripping voltammetry for the determination of nickel and cobalt in simulated PWR coolant.

    PubMed

    Torrance, K; Gatford, C

    1985-04-01

    The determination of ionic nickel and cobalt in simulated PWR coolant at concentrations below 1 microg/l. by differential pulse stripping voltammetry at a hanging mercury-drop electrode has been investigated. The high sensitivity for these ions results from the adsorptive accumulation of their dimethylglyoximate complexes on the mercury drop. Boric acid does not interfere and if the samples are adjusted to pH 9 with an ammonia-ammonium chloride buffer, both nickel and cobalt can be determined in the same run. The relative standard deviations at concentrations below 2 microg/l. are of the order of 5-7% and the limits of detection for nickel and cobalt are about 8 and 2 ng/l. respectively. These performance statistics show that this method is the most sensitive method currently available for determination of soluble nickel and cobalt in PWR coolant and it should prove to be most valuable in any corrosion studies of the materials of construction of the primary circuit of a PWR.

  18. 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.

  19. 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

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

    SciTech Connect

    Bailey, W.J.; Tokar, M.

    1982-12-01

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

  1. 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

  2. Reference Services.

    ERIC Educational Resources Information Center

    Bunge, Charles A.

    1999-01-01

    Discusses library reference services. Topics include the historical development of reference services; instruction in library use, particularly in college and university libraries; guidance; information and referral services and how they differ from traditional question-answering service; and future concerns, including user fees and the planning…

  3. Reference Assessment

    ERIC Educational Resources Information Center

    Bivens-Tatum, Wayne

    2006-01-01

    This article presents interesting articles that explore several different areas of reference assessment, including practical case studies and theoretical articles that address a range of issues such as librarian behavior, patron satisfaction, virtual reference, or evaluation design. They include: (1) "Evaluating the Quality of a Chat Service"…

  4. Reference Revolutions.

    ERIC Educational Resources Information Center

    Mason, Marilyn Gell

    1998-01-01

    Describes developments in Online Computer Library Center (OCLC) electronic reference services. Presents a background on networked cataloging and the initial implementation of reference services by OCLC. Discusses the introduction of OCLC FirstSearch service, which today offers access to over 65 databases, future developments in integrated…

  5. 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)

  6. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    NASA Astrophysics Data System (ADS)

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  7. IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC

    NASA Astrophysics Data System (ADS)

    PetroviĆ, B. G.

    1991-01-01

    The following sections are included: * INTRODUCTION * IN-CORE FUEL MANAGEMENT CALCULATIONS * In-Core Fuel Management * Methodological Problems of In-Core Fuel Management * In-Core Fuel Management Analytical Tools * PENN STATE FUEL MANAGEMENT PACKAGE * Penn State Fuel Management Package (PFMP) * Assembly Data Description (ADD) * Linking PSU-LEOPARD and MCRAC: An Example * MULTICYCLE REACTOR ANALYSIS CODE (MCRAC) * Main Features and Options of MCRAC code * Core geometry * Diffusion equations * 1.5-group model * Multicycle neutronic analysis * Multicycle cost analysis * Criticality search * Power-dependent xenon feedback calculations * Control rod and burnable absorber simulation * Search for LP with flat BOC power distribution * Artificial ADD option * Variable dimensioning technique * RBI version of MCRAC code * Programming changes in PC version * Fuel interchange option * MCRAC Input/Output * General input description * Sample input * Sample output * EXPERIENCE WITH MCRAC CODE * CONCLUSIONS * REFERENCES

  8. Neutralization of steam generator denting. Volume 1. Final report. [PWR

    SciTech Connect

    Wolfe, C.R.; Esposito, J.N.; Wozniak, S.M.; Whyte, D.D.

    1983-09-01

    This report deals with experimental laboratory work, the purpose of which was to reproduce the denting phenomenon observed in some steam generator heat transfer tubing and then to determine the effectiveness of selected candidate additives with regard to their ability to inhibit the denting phenomenon. Denting was shown to be dependent on a synergism between materials, oxidants, chloride, crevice geometry, superheats and temperature. A reference denting environment was developed and candidate inhibitors to this environment were subsequently added in an attempt to prevent and/or stop the denting process. Boric acid applications in a soak/on-line mode were found to be completely effective with no indications of deleterious side effects. Boric acid has been recommended for field application on a plant-specific basis. Calcium hydroxide and sodium phosphate additions in an on-line mode were effective inhibitors.

  9. Neutralization of steam generator denting. Volume 2. Final report. [PWR

    SciTech Connect

    Wolfe, C.R.; Esposito, J.N.; Wozniak, S.M.; Whyte, D.D.

    1983-09-01

    This report deals with experimental laboratory work, the purpose of which was to reproduce the denting phenomenon observed in some steam generator heat transfer tubing and then to determine the effectiveness of selected candidate additives with regard to their ability to inhibit the denting phenomenon. Denting was shown to be dependent on a synergism between materials, oxidants, chloride, crevice geometry, superheats and temperature. A reference denting environment was developed and candidate inhibitors to this environment were subsequently added in an attempt to prevent and/or stop the denting process. Boric acid applications in a soak/on-line mode were found to be completely effective with no indications of deleterious side effects. Boric acid has been recommended for field application on a plant-specific basis. Calcium hydroxide and sodium phosphate additions in an on-line mode were effective inhibitors.

  10. Ready Reference.

    ERIC Educational Resources Information Center

    Koltay, Emery

    1999-01-01

    Includes the following ready reference information: "Publishers' Toll-Free Telephone Numbers"; "How to Obtain an ISBN (International Standard Book Number)"; "How to Obtain an ISSN (International Standard Serial Number)"; and "How to Obtain an SAN (Standard Address Number)". (AEF)

  11. 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.

  12. 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.

  13. 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.

  14. THERMAL HISTORY OF CLADDING IN A 21 PWR WASTE PACKAGE LOADED WITH AVERAGE FUEL

    SciTech Connect

    H.M. Wade

    2000-01-25

    The purpose of this calculation is to evaluate a mid-assembly axial fuel cladding temperature profile of a 21 pressurized water reactor (PWR) spent nuclear fuel (SNF) waste package (WP) loaded with average fuel assemblies and emplaced in a monitored geologic repository. This calculation is intended to evaluate Viability Assessment (VA) and Enhanced Design Alternatives (EDA) II design configurations in support of performance assessment. This calculation was developed by Waste Package Operations (WPO) under Office of Civilian Radioactive Waste Management (OCRWM) procedure AP-3.12Q, Revision 0.

  15. 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.

  16. URSULA2 computer program. Volume 2. Applications (sensitivity studies and demonstration calculations). Final report. [PWR

    SciTech Connect

    Keeton, L.W.; Marchland, E.O.; Singhal, A.K.; Spalding, D.B.

    1980-01-01

    The URSULA2 computer program has been developed for the thermal-hydraulic analysis of steam generators for PWR nuclear power plants. It computes three-dimensional distributions of velocity, pressure, enthalpy, etc., in the shell of the generator, and the distributions of primary-fluid temperature within the tubes. The code is applicable to both steady and unsteady flows and is equiped with three physical models: the equal velocity homogeneous model, a slip (or two-fluid) model, and an algebraic slip model. Applications, sensitivity studies, and demonstration calculations are presented.

  17. 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.

  18. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    SciTech Connect

    Billone, M. C.; Burtseva, T. A.

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  19. Generic Crystalline Disposal Reference Case

    SciTech Connect

    Painter, Scott Leroy; Chu, Shaoping; Harp, Dylan Robert; Perry, Frank Vinton; Wang, Yifeng

    2015-02-20

    A generic reference case for disposal of spent nuclear fuel and high-level radioactive waste in crystalline rock is outlined. The generic cases are intended to support development of disposal system modeling capability by establishing relevant baseline conditions and parameters. Establishment of a generic reference case requires that the emplacement concept, waste inventory, waste form, waste package, backfill/buffer properties, EBS failure scenarios, host rock properties, and biosphere be specified. The focus in this report is on those elements that are unique to crystalline disposal, especially the geosphere representation. Three emplacement concepts are suggested for further analyses: a waste packages containing 4 PWR assemblies emplaced in boreholes in the floors of tunnels (KBS-3 concept), a 12-assembly waste package emplaced in tunnels, and a 32-assembly dual purpose canister emplaced in tunnels. In addition, three failure scenarios were suggested for future use: a nominal scenario involving corrosion of the waste package in the tunnel emplacement concepts, a manufacturing defect scenario applicable to the KBS-3 concept, and a disruptive glaciation scenario applicable to both emplacement concepts. The computational approaches required to analyze EBS failure and transport processes in a crystalline rock repository are similar to those of argillite/shale, with the most significant difference being that the EBS in a crystalline rock repository will likely experience highly heterogeneous flow rates, which should be represented in the model. The computational approaches required to analyze radionuclide transport in the natural system are very different because of the highly channelized nature of fracture flow. Computational workflows tailored to crystalline rock based on discrete transport pathways extracted from discrete fracture network models are recommended.

  20. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    SciTech Connect

    De Rosa, Felice

    2006-07-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when {delta}Tsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

  1. 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.

  2. 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.

  3. 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.

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

    SciTech Connect

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

    1995-12-31

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

  5. The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes

    SciTech Connect

    Pignatel, Jean-Francois

    2002-07-01

    Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, and leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor. (author)

  6. Poroelastic references

    SciTech Connect

    Morency, Christina

    2014-12-12

    This file contains a list of relevant references on the Biot theory (forward and inverse approaches), the double-porosity and dual-permeability theory, and seismic wave propagation in fracture porous media, in RIS format, to approach seismic monitoring in a complex fractured porous medium such as Brady?s Geothermal Field.

  7. Ready Reference.

    ERIC Educational Resources Information Center

    Koltay, Emery

    2001-01-01

    Includes four articles that relate to ready reference, including a list of publishers' toll-free telephone numbers and Web sites; how to obtain an ISBN (International Standard Book Number) and an ISSN (International Standard Serial Number); and how to obtain an SAN (Standard Address Number), for organizations that are involved in the book…

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

    SciTech Connect

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

    1985-04-01

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

  9. Field Artillery Reference Data.

    DTIC Science & Technology

    1980-01-01

    wrist 1 Test set elec tube7 Camouflage screen sys (2) SURVEY SECTIONS 1 Tool kit elec equip 1 Carrier CP 1 TIr 11/2T 7 Camouflage screen spt sys Personnel...units. SUMMARY - MAJOR ITEMS OF EQUIPMEN TRANSPORTATION 0 Carrier, cargo, tracked, 6-ton 4 Cable, tel, WD-1/TT, DR-8, 1/4 mi 131-. Carrier, CP , It tracked...regen repeater 3 Pwr sup assy 4 Tel set TA-312 6 Radio set GRC-1 60 2 TIr 1AT 9 Radio set PRC-77 2 Trk 1AT 3 Radio set VRC-47 2 Carrier CP 6 Radio

  10. Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

    SciTech Connect

    Simion, G.P.; VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B.; Bulmahn, K.D.

    1993-06-01

    This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement.

  11. Effect of single aging on stress corrosion cracking susceptibility of INCONEL X-750 under PWR conditions

    NASA Astrophysics Data System (ADS)

    Mishra, B.; Moore, J. J.

    1988-05-01

    Unfavorable morphology of precipitates and inclusions has been thought to be the cause of severe intergranular stress corrosion cracking (IGSCC) in double aged INCONEL* X-750 alloy used in reactor water environments. A single step aging treatment of 200 hours at 811 °C followed by furnace cooling after solution treating for 2 hours at 1075 °C has been found to provide an improved combination of strength, ductility, and resistance to SCC under simulated PWR test conditions. In this single aged condition a reprecipitated secondary carbide, together with γ' was produced at the grain boundary which resulted in a mixed fracture mode comprising dimple rupture and microvoid coalescence compared with a predominantly intergranular mode for the fully age hardened specimens. This improvement has been explained in terms of the morphology of the second phase precipitates which are produced in these heat treatment regimes.

  12. Grid-to-rod flow-induced impact study for PWR fuel in reactor

    SciTech Connect

    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 effective way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.

  13. Source term experiment STEP-3 simulating a PWR severe station blackout

    SciTech Connect

    Simms, R.; Baker, L. Jr.; Ritzman, R.L.

    1987-05-21

    For a severe PWR accident that leads to a loss of feedwater to the steam generators, such as might occur in a station blackout, fission product decay heating will cause a water boiloff. Without effective cooling of the core, steam will begin to oxidize the Zircaloy cladding. The noble gases and volatile fission products, such as Cs and I, that are major contributors to the radiological source term, will be released from the damaged fuel shortly after cladding failure. The accident environment when these volatile fission products escape was simulated in STEP-3 using four fuel elements from the Belgonucleaire BR3 reactor. The primary objective was to examine the releases in samples collected as close to the test zone as possible. In this paper, an analysis of the temperatures and hydrogen generation is compared with the measurements. The analysis is needed to estimate releases and characterize conditions at the source for studies of fission product transport.

  14. 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%.

  15. 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.

  16. Development of cement solidification process for sodium borate waste generated from PWR plants

    SciTech Connect

    Hirofumi Okabe; Tatsuaki Sato; Yuichi Shoji; Yoshiko Haruguchi; Masaaki Kaneko; Michitaka Saso; Masumitsu Toyohara

    2013-07-01

    A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. To obtain high volume reduction and high mechanical strength of the waste, simulated concentrated borate liquid waste with a sodium / boron (Na/B) mole ratio of 0.27 was dehydrated and powdered by using a wiped film evaporator. To investigate the effect of the Na/B mole ratio on the solidification process, a sodium tetraborate decahydrate reagent with a Na/B mole ratio of 0.5 was also used. Ordinary portland cement (OPC) and some additives were used for the solidification. Solidified cement prepared from powdered waste with a Na/B mole ratio 0.24 and having a high silica sand content (silica sand/cement>2) showed to improved uniaxial compressive strength. (authors)

  17. Investigation of radial power and temperature effects in large-scale reflood experiments. [PWR

    SciTech Connect

    Motley, F.

    1983-01-01

    The largest reflood test facility in the world has been designed and constructed by the Japan Atomic Energy Research Institute (JAERI). The experimental test facility, known as the Cylindrical Core Test Facility (CCTF), models a full-height core section and the four primary loops of a Pressurized Water Reactor (PWR). The radial power distribution and temperature distribution were varied during the testing program. The test results indicate that the radial effects, while noticeable, do not appreciably alter the overall quenching behavior of the facility. The Transient Reactor Analysis Code (TRAC) correctly predicted the experimental results of several of the tests. The code results indicate that the core flow pattern adjusts multidimensionally to mitigate the effects of increased power or stored energy.

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

    SciTech Connect

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

    1983-04-01

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

  19. Fog inerting effects on hydrogen combustion in a PWR ice condenser contaminant

    SciTech Connect

    Luangdilok, W.; Bennett, R.B.

    1995-05-01

    A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the upward lean flammability limits of the H{sub 2}-air-steam mixture in the ice condenser upper plenum region of a pressurized water reactor (PWR) ice condenser contaminant during postulated large loss of coolant accident (LOCA) conditions indicate that combustion may be suppressed beyond the downward flammability limit (8 percent H{sub 2} by volume). 18 refs., 3 tabs.

  20. 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.

  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. 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.

  3. Effect of coolant chemistry on PWR radiation transport processes. Progress report on reactor loop studies

    SciTech Connect

    Brown, D.J.; Flynn, G.; Haynes, J.W.; Kitt, G.P.; Large, N.R.; Lawson, D.; Mead, A.P.; Nichols, J.L.; Woodwark, D.R.

    1986-05-01

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. There are strong indications that the in-core deposition behavior of corrosion product species is not fully accounted for by the solubility model based on nickel ferrite; boric acid plays a role apart from its influence on pH, and corrosion products are adsorbed to some extent in the zirconium oxide film on the fuel cladding. In DWL, soluble species appear to be dominant in deposition processes. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. 13 figs.

  4. The key to superior water chemistry at a PWR nuclear station

    SciTech Connect

    Dolan, R.; Miller, L.K.; Olejar, L.L.; Salem, E.

    1983-01-01

    This paper demonstrates how a condensate polishing unit can be successfully used to treat the feedwater for circulating-type pressurized water reactors (PWRs). Water chemistry at the Salem Generating Station, a two-unit, four-loop Westinghouse PWR located in New Jersey, is discussed. Topics considered include a plant description and the history of early operation, the role of constant surveillance, makeup water quality, the effect of freezing on gel-type anion exchange resin, a total organic carbon (TOC) survey, steam generator chemistry, steam generator inspection, condensate polisher operation, and management philosophy. The SEPREX condensate polishing process, in which the complete separation of the anion exchange resin from the cation exchange resin is achieved by flotation separation, is examined. It is concluded that the utilization of a condensate polishing process such as SEPREX provides the operating personnel at the plant with the necessary means to maintain the minimum desired level of contaminants within the steam generator.

  5. 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.

  6. Grid-to-rod flow-induced impact study for PWR fuel in reactor

    DOE PAGES

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

    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

  7. 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.

  8. Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

    SciTech Connect

    Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.; Behling, S.R.

    1981-01-01

    RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of a hypothetical loss-of-coolant accident (LOCA).

  9. 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

  10. 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)

  11. 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)

  12. 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).

  13. 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.

  14. 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.

  15. Constraints on silicates formation in the Si-Al-Fe system: Application to hard deposits in steam generators of PWR nuclear reactors

    NASA Astrophysics Data System (ADS)

    Berger, Gilles; Million-Picallion, Lisa; Lefevre, Grégory; Delaunay, Sophie

    2015-04-01

    Introduction: The hydrothermal crystallization of silicates phases in the Si-Al-Fe system may lead to industrial constraints that can be encountered in the nuclear industry in at least two contexts: the geological repository for nuclear wastes and the formation of hard sludges in the steam generator of the PWR nuclear plants. In the first situation, the chemical reactions between the Fe-canister and the surrounding clays have been extensively studied in laboratory [1-7] and pilot experiments [8]. These studies demonstrated that the high reactivity of metallic iron leads to the formation of Fe-silicates, berthierine like, in a wide range of temperature. By contrast, the formation of deposits in the steam generators of PWR plants, called hard sludges, is a newer and less studied issue which can affect the reactor performance. Experiments: We present here a preliminary set of experiments reproducing the formation of hard sludges under conditions representative of the steam generator of PWR power plant: 275°C, diluted solutions maintained at low potential by hydrazine addition and at alkaline pH by low concentrations of amines and ammoniac. Magnetite, a corrosion by-product of the secondary circuit, is the source of iron while aqueous Si and Al, the major impurities in this system, are supplied either as trace elements in the circulating solution or by addition of amorphous silica and alumina when considering confined zones. The fluid chemistry is monitored by sampling aliquots of the solution. Eh and pH are continuously measured by hydrothermal Cormet© electrodes implanted in a titanium hydrothermal reactor. The transformation, or not, of the solid fraction was examined post-mortem. These experiments evidenced the role of Al colloids as precursor of cements composed of kaolinite and boehmite, and the passivation of amorphous silica (becoming unreactive) likely by sorption of aqueous iron. But no Fe-bearing was formed by contrast to many published studies on the Fe

  16. Results of small break LOCA experiments in the LOFT reactor system with comparison to code calculations. [PWR

    SciTech Connect

    Adams, J.P.; Linebarger, J.H.; Leach, L.P.

    1980-01-01

    The results are presented of three small break loss-of-coolant experiments performed in the LOFT Pressurized Water Reactor (PWR) system. Experiment L3-0, performed without reactor power, represented a loss of coolant from the power operated relief valve on the top of the pressurizer. Experiments L3-1 and L3-2 were initiated with the reactor at full power (maximum linear heat generation rate approximately 52 kW/m) and represented 4-in and 1-in diameter breaks, respectively, in the reactor inlet piping of a commercial PWR. Comparisons of data to analytical model calculations with a number of different models indicate that most major phenomena were correctly calculated, but that improvements in modeling small break behavior are necessary.

  17. 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.

  18. Subchannel Thermal-Hydraulic Experimental Program (STEP). Volume 1. Mixing in a pressurized water reactor (PWR) rod bundle. Final report

    SciTech Connect

    Barber, A.R.; Zielke, L.A.

    1980-08-01

    This volume describes an experiment that was performed to determine the mixing characteristics of a pressurized water reactor (PWR) rod bundle. The objective of this project was to improve the subchannel computer code models of the reactor core. The experimental technique was isokinetic subchannel withdrawal of the entire flow from two sample subchannels. Once withdrawn, the sample fluid was condensed and its enthalpy was measured by regenerative heat exchange calorimetry. The test bundle was a 4 x 6 electrically heated array with a 50% power upset. The COBRA IIIC code was used to model the experiment and to determine the value of the thermal mixing coefficient, ..beta.., that was necessary to predict the measured results. Both single- and two-phase data were obtained over a range of PWR operating conditions. The results indicate that both single- and two-phase mixing is small. The COBRA model predicts the enthalpy data using a turbulent mixing coefficient, ..beta.. approx. = 0.002.

  19. 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.

  20. 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.

  1. 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.

  2. PwrSoC (integration of micro-magnetic inductors/transformers with active semiconductors) for more than Moore technologies

    NASA Astrophysics Data System (ADS)

    Mathuna, Cian Ó.; Wang, Ningning; Kulkarni, Santosh; Roy, Saibal

    2013-07-01

    This paper introduces the concept of power supply on chip (PwrSoC) which will enable the development of next-generation, functionally integrated, power management platforms with applications in dc-dc conversion, gate drives, isolated power transmission and ultimately, high granularity, on-chip, power management for mixed-signal, SOC chips. PwrSoC will integrate power passives with the power management IC, in a 3D stacked or monolithic form factor, thereby delivering the performance of a highefficiency dc-dc converter within the footprint of a low-efficiency linear regulator. A central element of the PwrSoC concept is the fabrication of power micro-magnetics on silicon to deliver micro-inductors and micro-transformers. The paper details the magnetics on silicon process which combines thin film magnetic core technology with electroplated copper conductors. Measured data for micro-inductors show inductance operation up to 20 MHz, footprints down to 0.5 mm2, efficiencies up to 93% and dc current carrying capability up to 600 mA. Measurements on micro-transformers show voltage gain of approximately - 1 dB at between 10 MHz and 30 MHz. Contribution to the Topical Issue “International Semiconductor Conference Dresden-Grenoble - ISCDG 2012”, Edited by Gérard Ghibaudo, Francis Balestra and Simon Deleonibus.

  3. Development and Application of Laser Peening System for PWR Power Plants

    SciTech Connect

    Masaki Yoda; Itaru Chida; Satoshi Okada; Makoto Ochiai; Yuji Sano; Naruhiko Mukai; Gaku Komotori; Ryoichi Saeki; Toshimitsu Takagi; Masanori Sugihara; Hirokata Yoriki

    2006-07-01

    Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion cracking (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J

  4. Key Issues for the control of refueling outage duration and costs in PWR Nuclear Power Plants

    SciTech Connect

    Degrave, Claude

    2002-07-01

    For several years, EDF, within the framework of the CIDEM1 project and in collaboration with some German Utilities, has undertaken a detailed review of the operating experience both of its own NPP and of foreign units, in order to improve the performances of future units under design, particularly the French-German European Pressurized Reactor (EPR) project. This review made it possible to identify the key issues allowing to decrease the duration of refueling and maintenance outages. These key issues can be classified in 3 categories Design, Maintenance and Logistic Support, Outage Management. Most of the key issues in the design field and some in the logistic support field have been studied and could be integrated into the design of any future PWR unit, as for the EPR project. Some of them could also be adapted to current plants, provided they are feasible and profitable. The organization must be tailored to each country, utility or period: it widely depends on the power production environment, particularly in a deregulation context. (author)

  5. Iodine partition coefficient measurements at simulated PWR steam generator conditions: Interim data report

    SciTech Connect

    Clinton, S.D.; Simmons, C.M.

    1987-05-01

    Iodine partition coefficients (defined as the ratio of the concentration of iodine species in the aqueous solution to the iodine concentration in the vapor phase) were measured at simulated PWR steam generator conditions (285C and 6.9 MPa), using carrier-free radioactive T I in the form of sodium iodide. The iodine tracer concentration was maintained at approx.6 x 10 mol/L; boric acid concentration was varied from 0 to 0.4 mol/L; and the solution pH (measured at 25C) was adjusted from 4 to 9 by the addition of lithium hydroxide. Iodine partition coefficients decrease with increasing boric acid concentration; however, the iodine volatility is essentially independent of the solution pH for a given boric acid concentration. Sparging the solutions with air at room temperature increases the iodine volatility by an order of magnitude, compared to that achieved with argon sparging. Iodine partition coefficient measurements ranged from a low of 200 (in 0.2 M boric acid sparged with air) to 400,000 (in purified water sparged with argon).

  6. TREAT source-term experiment STEP-1 simulating a PWR LOCA

    SciTech Connect

    Simms, R.; Baker, L. Jr.; Blomquist, C.A.; Ritzman, R.L.

    1986-01-01

    In a hypothetical pressurized water reactor (PWR) large-break loss-of-coolant accident (LOCA) in which the emergency core cooling system fails, fission product decay heating causes water boil-off and reduced heat removal. Zircaloy cladding is oxidized by the steam. The noble gases and volatile fission products such as cesium and iodine that constitute a principal part of the source term will be released from the damaged fuel at or shortly after the time of cladding failure. TREAT test STEP-1 simulated the LOCA environment when the volatile fission products would be released using four fuel elements from the Belgonucleaire BR3 reactor. The principal objective was to collect a portion of the releases carried by the flow stream in a region as close as possible to the test zone. In this paper, the test is described and the results of an analysis of the thermal and steam/hydrogen environment are compared with the test measurements in order to provide a characterization for analysis of fission product releases and aerosol formation. The results of extensive sample examinations are reported separately.

  7. 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.

  8. 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.

  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. 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.

  11. 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.

  12. Validation of the scale system for PWR spent fuel isotopic composition analyses

    SciTech Connect

    Hermann, O.W.; Bowman, S.M.; Parks, C.V.; Brady, M.C.

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  13. Evaluation of on-line chelant addition to PWR steam generators. Steam generator cleaning project

    SciTech Connect

    Tvedt, T.J.; Wallace, S.L.; Griffin, F. Jr.

    1983-09-01

    The investigation of chelating agents for continuous water treatment of secondary loops of PWR steam generators were conducted in two general areas: the study of the chemistry of chelating agents and the study of materials compatability with chelating agents. The thermostability of both EDTA and HEDTA metal chelates in All Volatile Treatment (AVT) water chemistry were shown to be greater than or equal to the thermostability of EDTA metal chelates in phosphate-sulfite water chemistry. HEDTA metal chelates were shown to have a much greater stability than EDTA metal chelates. Using samples taken from the EDTA metal chelate thermostability study and from the Commonwealth Research Corporation (CRC) model steam generators (MSG), EDTA decomposition products were determined. Active metal surfaces were shown to become passivated when exposed to EDTA and HEDTA concentrations as high as 0.1% w/w in AVT. Trace amounts of iron in the water were found to increase the rate of passivation. Material balance and visual inspection data from CRC model steam generators showed that metal was transported through and cleaned from the MSG's. The Inconel 600 tubes of the salt water fouled model steam generators experienced pitting corrosion. Results of this study demonstrates the feasibility of EDTA as an on-line water treatment additive to maintain nuclear steam generators in a clean condition.

  14. Endurance tests of valves with cobalt-free hardfacing alloys: PWR phase

    SciTech Connect

    Murphy, E.V.; Inglis, I. )

    1992-05-01

    Atomic Energy of Canada Limited (AECL) is conducting endurance tests on valves hard-faced with four cobalt-free alloys. The first phase of the program, in which PWR primary heat transport conditions were simulated in AECL's valve test loop, has been completed. The candidate alloys are NOREM 01, NOREM 04, EB 5183 and EVERIT 50. One valve with Stellite 6 trim served as the standard. Prior to loop testing, a baseline inaugural inspection was performed. During testing the loop was shutdown at approximately 500 cycle intervals, and the valves were disassembled for examination. The examinations included seat leak tests, profilometry, nondestructive inspection and finally destructive examination. Corrosion coupons in the loop were used to monitor any material loss due solely to corrosion mechanisms. This report summarizes the final examination results and discusses the relative performance of the candidate alloys. The results indicate that, based upon the sliding wear damage assessment and seat leakage test results, all the candidate alloys perform better than the Stellite 6 control sample. On the same basis, NOREM 04 and EB 5183 are the best of the candidate alloys, although there are only minor differences in performance among the four alloys.

  15. Endurance tests of valves with cobalt-free hardfacing alloys: PWR phase. Final report

    SciTech Connect

    Murphy, E.V.; Inglis, I.

    1992-05-01

    Atomic Energy of Canada Limited (AECL) is conducting endurance tests on valves hard-faced with four cobalt-free alloys. The first phase of the program, in which PWR primary heat transport conditions were simulated in AECL`s valve test loop, has been completed. The candidate alloys are NOREM 01, NOREM 04, EB 5183 and EVERIT 50. One valve with Stellite 6 trim served as the standard. Prior to loop testing, a baseline inaugural inspection was performed. During testing the loop was shutdown at approximately 500 cycle intervals, and the valves were disassembled for examination. The examinations included seat leak tests, profilometry, nondestructive inspection and finally destructive examination. Corrosion coupons in the loop were used to monitor any material loss due solely to corrosion mechanisms. This report summarizes the final examination results and discusses the relative performance of the candidate alloys. The results indicate that, based upon the sliding wear damage assessment and seat leakage test results, all the candidate alloys perform better than the Stellite 6 control sample. On the same basis, NOREM 04 and EB 5183 are the best of the candidate alloys, although there are only minor differences in performance among the four alloys.

  16. 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.

  17. Experimental investigation on denting in PWR steam generators: causes and corrective actions

    SciTech Connect

    Nordmann, F.; Brunet, J.P.; Duret, J.; Pinard-Legry, G.

    1983-10-01

    Denting studies have been undertaken in order to assess the influence of the most important parameters which could initiate corrosion of the carbon steel occurring in the tube-tube support plate crevices of some PWR steam generators. Tests have been carried out in model boilers where feedwater was polluted with sea or river water. Specific effects of chloride or sulfate and influence of oxygen content, magnetite addition and pH value were investigated. In magnetite prepacked crevices, denting is obtained within 1000 hrs for seawater pollution of 0.3 ppm chloride at the blowdown. In neutral chloride or in river water, denting is observed only with oxygen addition. Denting prevention is effective in the case of an on-line addition of phosphate, boric acid, or calcium hydroxide. For denting stopping, boric acid or calcium hydroxide is efficient even with a high seawater pollution. Soaks cannot stop denting if they are not followed by an on-line treatment (boric acid, calcium hydroxide). With quadrifoil holes, denting doesn't occur. In very severe test conditions, 13 percent Cr steel can be corroded, but the corrosion rate is low and oxide morphology is different from that growing on carbon steel.

  18. Survey of the literature on low-alloy steel fastener corrosion in PWR power plants

    SciTech Connect

    Hall, J.F.

    1984-12-01

    This report presents the results of a literature survey of low alloy steel fastener corrosion in PWR applications. The report addresses boric acid corrosion (accelerated general corrosion) and stress corrosion cracking of threaded fasteners used in primary pressure boundary closures, in secondary, auxiliary, and safety system closures and in component support applications. The report reviews and summarizes corrosion events that have occurred in domestic PWRs since 1968. Information provided for each event includes plant identification, year of event, major component or part affected, fastener material, fastener diameter, number of corroded studs, the service environments, the number of degraded fasteners and the results of post-service failure analyses. Possible corrective actions that are available to eliminate or mitigate the effects of the two types of corrosion are also identified. Laboratory test data, including some recent unpublished data, that are related to fastener corrosion are also discussed. The report also includes recommended additional work in the areas of boric acid corrosion, stress corrosion cracking and analytical methodologies to solve these fastener corrosion problems.

  19. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    SciTech Connect

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-07-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  20. Brief account of the effect of overcooling accidents on the integrity of PWR pressure vessels

    SciTech Connect

    Cheverton, R.D.

    1982-01-01

    The occurrence in recent years of several (PWR) accident initiating events that could lead to severe thermal shock to the reactor pressure vessel, and the growing awareness that copper and nickel in the vessel material significantly enhance radiation damage in the vessel, have resulted in a reevaluation of pressure-vessel integrity during postulated overcooling accidents. Analyses indicate that the accidents of concern are those involving both thermal shock and pressure loadings, and that an accident similar to that at Rancho Seco in 1978 could, under some circumstances and at a time late in the normal life of the vessel, result in propagation of preexistent flaws in the vessel wall to the extent that they might completely penetrate the wall. More severe accidents have been postulated that would result in even shorter permissible lifetimes. However, the state-of-the-art fracture-mechanics analysis may contain excessive conservatism, and this possibility is being investigated. Furthermore, there are several remedial measures, such as fuel shuffling, to reduce the damage rate, and vessel annealing, to restore favorable material properties, that may be practical and used if necessary. 5 figures.

  1. A new advanced fixed in-core instrumentation for a PWR reactor

    NASA Astrophysics Data System (ADS)

    Barbet, M.; Guillery, M.

    1981-06-01

    Gamma thermometer studies have been done at E.D.F. for four years. These studies started in France with a feasibility study in 1975. E.D.F.'s scope was to develop a new fixed "in-core" instrumentation for PWR based on the gamma heat measurements. The advanced gamma thermometer design has been done in such a way to be able to manufacture strings of 6 to 9 detectors each. The results of gamma thermometer make up in 1976 were encouraging and E.D.F. went on to develop a gamma thermometer assembly for a reactor application. Before being mounted on the reactor vessel, the gamma thermometer strings are calibrated in a loop test by means of an electrical current giving the ΔT versus the specific power ( W/ g). The loop test simulates the thermohydraulic conditions in the reactor tube guide. Two gamma thermometer strings have been installed in the BUGEY 5 reactor since June 1979. Four gamma thermometer strings are provided for insertion in the TRICASTIN 2 reactor and four more gamma thermometer strings are manufactured to be ready for the start up of the TRICASTIN 3 reactor in 1980.

  2. Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

    SciTech Connect

    Fast, Ivan; Bosbach, Dirk; Aksyutina, Yuliya; Tietze-Jaensch, Holger

    2015-07-01

    A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

  3. Influence of noncondensible gas on heat removal from the primary of a PWR

    SciTech Connect

    Umminger, K.; Mandl, R.; Schoen, B.

    1990-01-01

    Under loss-of-coolant accident conditions, there is a possibility that noncondensible gas (i.e., nitrogen, hydrogen, or fission gas) will enter the primary system, which can adversely affect the capability to remove decay heat. Small- and medium-sized breaks cause depressurization and lead to release of N{sub 2} dissolved in the primary coolant and accumulator inventories. Failure to close of an accumulator isolation valve after the accumulator content has emptied into the primary can result in significant amounts of propellant gas entering the primary system. In the event of a total loss of on- and off-site power, the feedwater is also lost. With the main steam isolation valves close, the secondaries boil dry through relief valves. The core decay heat leads to pressurization of the primary system, opening of the pressurizer safety relief valve, and loss of primary inventory. Without operator intervention, this scenario results in core uncovery and core damage as the primary inventory is depleted. At temperatures >800{degree}C (1500{degree}F), zircon/water reaction will take place accompanied by formation of substantial amounts of hydrogen. At this stage, even restored heat transfer (e.g., resumption of feedwater flow) will be impeded by the presence of the hydrogen. The influence of noncondensible gases on the heat transfer capability of a four-loop pressurized water reactor (PWR) was investigated in several parametric studies carried out in the PKL test facility.

  4. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  5. 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)

  6. Reference Frames and Relativity.

    ERIC Educational Resources Information Center

    Swartz, Clifford

    1989-01-01

    Stresses the importance of a reference frame in mechanics. Shows the Galilean transformation in terms of relativity theory. Discusses accelerated reference frames and noninertial reference frames. Provides examples of reference frames with diagrams. (YP)

  7. The effect of stainless steel overlay cladding on corrosion fatigue crack propagation in pressure vessel steel in PWR primary coolant

    SciTech Connect

    Bramwell, I.L.; Tice, D.R.; Worswick, D.; Heys, G.B.

    1995-12-31

    The growth of sub-critical cracks in pressure boundary materials in light water reactors is assessed using codified procedures, but the presence of the overlay-welded stainless steel cladding on the pressure vessel is not normally taken into consideration because of the difficulty in demonstrating clad integrity for the lifetime of the plant. In order to investigate any possible effect of the cladding layer on crack propagation, tests have been performed using two types of specimen. The first was sputter ion plated with a thin layer of austenitic stainless steel to simulate the electrochemical and oxide effects due to the cladding, whilst the second used an overlay clad specimen to investigate the behavior of a crack propagating from the austenitic into the ferritic material. Testing was carried out under cyclic loading conditions in well controlled simulated PWR primary water. At 288 C, the presence of stainless steel in contact with the low alloy steel did not enhance crack propagation in PWR primary coolant compared to unclad or unplated specimens. There was limited evidence that at 288 C under certain loading conditions, in both air and PWR water, there may be an effect of the cladding which reduces crack growth rates, at least for a short distance of crack propagation into the low alloy steel. Crack growth rates in the ferritic steel at 130 C were higher for both the plated and clad specimens than found in previous tests under similar conditions on the unclad material. However, the crack growth rates were bounded by current ASME 11 Appendix A recommendations for defects exposed to water and at low R ratio. There was no evidence of environmental enhancement of crack propagation in the stainless steel in clad specimens. The results indicate that the current approach of ignoring the cladding for assessment purposes is conservative at plant operating temperature.

  8. ATHOS: a computer program for thermal-hydraulic analysis of steam generators. Volume 3. User's manual. [PWR

    SciTech Connect

    Singhal, A.K.; Keeton, L.W.; Przekwas, A.J.; Weems, J.S.

    1982-10-01

    ATHOS (Analysis of the Thermal Hydraulics of Steam Generators) is a computer code developed by CHAM of North America Incorporated, under the contract RP 1066-1 from the Electric Power Research Institute, Palo Alto, California. ATHOS supersedes the earlier code URSULA2. ATHOS is designed for three-dimensional, steady state and transient analyses of PWR steam generators. The current version of the code has been checked out for: three different configurations of the recirculating-type U-tube steam generators; the homogeneous and algebraic-slip flow models; and full and part load operating conditions.

  9. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect

    Osamu KAawabata; Mitsuhiro Kajimoto

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

  10. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    NASA Astrophysics Data System (ADS)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (E<0.625 eV) are required as these affect the CR integrity through stresses/strains induced by coupled clad embrittlement / absorber swelling phenomena. The concept of the PSI scheme under development is to provide from validated core analysis models, the volumetric neutron source to a full core MCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  11. 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)

  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. 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.

  14. 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)

  15. 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.

  16. 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

  17. Parametric study of CHF data. Volume 2. A generalized subchannel CHF correlation for PWR and BWR fuel assemblies. Final report

    SciTech Connect

    Reddy, D.G.; Fighetti, C.F.

    1983-01-01

    The primary objective of this research was to develop a generalized subchannel CHF correlation based on the local fluid conditions obtained with the COBRA-IIIC thermal hydraulic subchannel code and covering PWR and BWR normal operating conditions as well as hypothetical loss-of-coolant accident (LOCA) conditions. In view of the importance of the local conditions predicted by the COBRA-IIIC code in the development of CHR correlation, the secondary objective was to improve the predictive capability of the COBRA-IIIC subchannel code. In the first phase of this study, the sensitivity of local enthalpies and local mass fluxes predicted by the COBRA-IIIC subchannel code to subcooled void correlation, bulk void correlation, two-phase friction multiplier correlation and turbulent mixing parameter was determined. In the second phase, based on the local conditions obtained with the COBRA-IIIC subchannel code, an accurate generalized subchannel CHF correlation was developed utilizing 3607 CHF data points from 65 test sections simulating PWR and BWR fuel assemblies.

  18. 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.

  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. 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.

  1. Reference Service Policy Statement.

    ERIC Educational Resources Information Center

    Young, William F.

    This reference service policy manual provides general guidelines to encourage reference service of the highest possible quality and to insure uniform practice. The policy refers only to reference service in the University Libraries and is intended for use in conjunction with other policies and procedures issued by the Reference Services Division.…

  2. 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.

  3. 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

  4. 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

  5. Reach for Reference. Four Recent Reference Books

    ERIC Educational Resources Information Center

    Safford, Barbara Ripp

    2004-01-01

    This article provides descriptions of four new science and technology encyclopedias that are appropriate for inclusion in upper elementary and/or middle school reference collections. "The Macmillan Encyclopedia of Weather" (Stern, Macmillan Reference/Gale), a one-volume encyclopedia for upper elementary and middle level students, is a…

  6. Fundamentals of Reference

    ERIC Educational Resources Information Center

    Mulac, Carolyn M.

    2012-01-01

    The all-in-one "Reference reference" you've been waiting for, this invaluable book offers a concise introduction to reference sources and services for a variety of readers, from library staff members who are asked to work in the reference department to managers and others who wish to familiarize themselves with this important area of…

  7. Live, Digital Reference.

    ERIC Educational Resources Information Center

    Kenney, Brian

    2002-01-01

    Discusses digital reference services, also known as virtual reference, chat reference, or online reference, based on a round table discussion at the 2002 American Library Association annual conference in Atlanta. Topics include numbers and marketing; sustainability; competition and models; evaluation methods; outsourcing; staffing and training;…

  8. Statistical Reference Datasets

    National Institute of Standards and Technology Data Gateway

    Statistical Reference Datasets (Web, free access)   The Statistical Reference Datasets is also supported by the Standard Reference Data Program. The purpose of this project is to improve the accuracy of statistical software by providing reference datasets with certified computational results that enable the objective evaluation of statistical software.

  9. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    NASA Astrophysics Data System (ADS)

    Hartini, Entin; Andiwijayakusuma, Dinan

    2014-09-01

    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.

  10. The evaluation of iron-base hardfacing alloys on gate valves after cycling under simulated PWR conditions for one year

    SciTech Connect

    Murphy, E.V.; Inglis, I.; Ocken, H.

    1992-12-31

    Gate valves hardfaced with iron-base alloys were exposed for about one year to simulated PWR conditions. The hardfacing alloys tested were EB 5183, EVERIT 50, NOREM 01 and NOREM 04. A gate valve with Satellite 6 was included in the test program as a control standard. During the test period the valves were opened and closed 2000 times. The performance of the valves was assessed by periodic leak tests and visual and profilometric characterisation of sealing surfaces. At the end of the test program, the seats and discs were destructively examined. The various examinations indicated all the iron-base alloys were superior to Satellite 6. Based on the results of hot leakage tests, one valve with EB 5183 and the valve with NOREM 04 were the best performers.

  11. 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.

  12. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Maldonado, G I; Burdo, J; He, T

    2006-10-10

    A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the

  13. Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

    SciTech Connect

    Hsu, M.T.; Davis, C.B.; Behling, S.R.

    1981-11-01

    This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio was maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).

  14. 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.

  15. Genetics Home Reference

    MedlinePlus

    Skip Navigation Bar Home Current Issue Past Issues Genetics Home Reference Past Issues / Spring 2007 Table of ... of this page please turn Javascript on. The Genetics Home Reference (GHR) Web site — ghr.nlm.nih. ...

  16. Herbal reference standards.

    PubMed

    Schwarz, Michael; Klier, Bernhard; Sievers, Hartwig

    2009-06-01

    This review describes the current definitions and regulatory requirements that apply to reference standards that are used to analyse herbal products. It also describes and discusses the current use of reference substances and reference extracts in the European and United States pharmacopoeias.

  17. Academic Library Reference Services.

    ERIC Educational Resources Information Center

    Batt, Fred

    This examination of the philosophy and objectives of academic library reference services provides an overview of the major reference approaches to fulfilling the following primary objectives of reference services: (1) providing accurate answers to patrons' questions and/or helping patrons find sources to pursue their research needs; (2) building…

  18. References for marine science

    NASA Astrophysics Data System (ADS)

    1990-06-01

    Standard and Reference Materials for Marine Science, National Oceanic and Atmospheric Administration Technical Memo OMA-51 (2nd edition, 434 pp.), by A. Y. Cantillo, is now available. This compilation of reference materials was prepared at the request of the Group of Experts on Standards and Reference Materials and was printed by NOAA. GESREM is sponsored by the International Atomic Energy Agency, the Intergovernmental Oceanographic Commission, and the United Nations Program.Reference materials are included on ashes, gases, instrument performance materials, oils, physical properties, rocks, sediments, sludges, tissues and waters. For each reference material, source, description and preparation, analyses and values, cost, references, and comments are given. Indices are included for elements, isotopes and organic compounds. Cross references to Chemical Abstracts Service registry numbers and alternate names and chemical structures of organic compounds are also provided.

  19. Evaluation of storing Shippingport Core II spent blanket fuel assemblies in the T Plant PWR Core II fuel pool without active cooling

    SciTech Connect

    Gilbert, E.R.; Lanning, D.D.; Dana, C.M.; Hedengren, D.C.

    1994-10-01

    PWR Core II fuel pool chiller-off test was conducted because it appeared possible that acceptable pool-water temperatures could be maintained without operating the chillers, thus saving hundreds of thousands of dollars in maintenance and replacement costs. Test results showed that the water-cooling capability is no longer needed to maintain pool temperature below 38{degrees}C (100{degrees}F).

  20. Scoping design analyses for optimized shipping casks containing 1-, 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect

    Bucholz, J.A.

    1983-01-01

    This report details many of the interrelated considerations involved in optimizing large Pb, Fe, or U-metal spent fuel shipping casks containing 1, 2, 3, 5, 7, or 10-year-old PWR fuel assemblies. Scoping analyses based on criticality, shielding, and heat transfer considerations indicate that some casks may be able to hold as many as 18 to 21 ten-year-old PWR fuel assemblies. In the criticality section, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. In another section, details of many n/..gamma.. shielding optimization studies are presented, including the optimal n/..gamma.. design points and the actual shielding requirements for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. Multigroup source terms based on ORIGEN2 calculations at these and other decay times are also included. Lastly, the numerical methods and experimental correlations used in the steady-state and transient heat transfer analyses are fully documented, as are pertinent aspects of the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. (While only casks for square, intact PWR fuel assemblies were considered in this study, the SCOPE code may also be used to design and analyze casks containing canistered spent fuel or other waste material. An abbreviated input data guide is included as an appendix).

  1. Summary report on optimized designs for shipping casks containing 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect

    Bucholz, J.A.

    1983-04-01

    The purpose of this study was to develop new conceptual designs for large Pb, Fe, and U-shielded spent fuel casks which have been optimized for the shipment of 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel assemblies. Design specifications for about 100 cases of potential interest are presented along with a brief 20-page synopsis of the associated analyses. Optimized shielding requirements are presented for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. With respect to criticality, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. Steady-state and transient heat transfer analyses for casks under nominal and accident conditions were performed using the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. Based on criticality, shielding, and heat transfer considerations, it appears that optimized cask designs could be developed to carry 15 to 18 five-year-old PWR fuel assemblies or as many as 18 to 21 ten-year-old PWR fuel assemblies. 4 figures, 4 tables.

  2. Radionuclide release calculations for selected severe accident scenarios. Volume 3. PWR, subatmospheric containment design

    SciTech Connect

    Denning, R.S.; Gieseke, J.A.; Cybulskis, P.; Lee, K.W.; Jordan, H.; Curtis, L.A.; Kelly, R.F.; Kogan, V.; Schumacher, P.M.

    1986-07-01

    This report presents results of analyses of the enviromental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with a subatmospheric containment design. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the US Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. the Surry plant has been used in this study as the reference plant for a subatmospheric design.

  3. High frequency reference electrode

    DOEpatents

    Kronberg, James W.

    1994-01-01

    A high frequency reference electrode for electrochemical experiments comprises a mercury-calomel or silver-silver chloride reference electrode with a layer of platinum around it and a layer of a chemically and electrically resistant material such as TEFLON around the platinum covering all but a small ring or "halo" at the tip of the reference electrode, adjacent to the active portion of the reference electrode. The voltage output of the platinum layer, which serves as a redox electrode, and that of the reference electrode are coupled by a capacitor or a set of capacitors and the coupled output transmitted to a standard laboratory potentiostat. The platinum may be applied by thermal decomposition to the surface of the reference electrode. The electrode provides superior high-frequency response over conventional electrodes.

  4. High frequency reference electrode

    DOEpatents

    Kronberg, J.W.

    1994-05-31

    A high frequency reference electrode for electrochemical experiments comprises a mercury-calomel or silver-silver chloride reference electrode with a layer of platinum around it and a layer of a chemically and electrically resistant material such as TEFLON around the platinum covering all but a small ring or halo' at the tip of the reference electrode, adjacent to the active portion of the reference electrode. The voltage output of the platinum layer, which serves as a redox electrode, and that of the reference electrode are coupled by a capacitor or a set of capacitors and the coupled output transmitted to a standard laboratory potentiostat. The platinum may be applied by thermal decomposition to the surface of the reference electrode. The electrode provides superior high-frequency response over conventional electrodes. 4 figs.

  5. Preparing the references.

    PubMed

    Peh, W C G; Ng, K H

    2009-07-01

    In a scientific paper, the references serve to provide background information and allow the researcher to compare and contrast the work of others in relation to his own study. Authors are responsible for the accuracy of all references cited. The references quoted should be easily accessible and retrievable by anyone wishing to obtain further information. There is a strong preference for citing journal articles listed in PubMed. The two major reference format systems are the Vancouver and Harvard systems, with increasing preference for the Vancouver system. Authors should adhere exactly to the instructions to authors of the target journal.

  6. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    NASA Astrophysics Data System (ADS)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  7. Reference Point Heterogeneity

    PubMed Central

    Terzi, Ayse; Koedijk, Kees; Noussair, Charles N.; Pownall, Rachel

    2016-01-01

    It is well-established that, when confronted with a decision to be taken under risk, individuals use reference payoff levels as important inputs. The purpose of this paper is to study which reference points characterize decisions in a setting in which there are several plausible reference levels of payoff. We report an experiment, in which we investigate which of four potential reference points: (1) a population average payoff level, (2) the announced expected payoff of peers in a similar decision situation, (3) a historical average level of earnings that others have received in the same task, and (4) an announced anticipated individual payoff level, best describes decisions in a decontextualized risky decision making task. We find heterogeneity among individuals in the reference points they employ. The population average payoff level is the modal reference point, followed by experimenter's stated expectation of a participant's individual earnings, followed in turn by the average earnings of other participants in previous sessions of the same experiment. A sizeable share of individuals show multiple reference points simultaneously. The reference point that best fits the choices of the individual is not affected by a shock to her income. PMID:27672374

  8. Marketing Reference Services.

    ERIC Educational Resources Information Center

    Norman, O. Gene

    1995-01-01

    Relates the marketing concept to library reference services. Highlights include a review of the literature and an overview of marketing, including research, the marketing mix, strategic plan, marketing plan, and marketing audit. Marketing principles are applied to reference services through the marketing mix elements of product, price, place, and…

  9. An Online Reference System.

    ERIC Educational Resources Information Center

    Chisman, Janet; Treat, William

    1984-01-01

    Describes a computer aid developed to assist in academic library reference service using the DataPhase Circulation System, an automated system that features full cataloging records in database and permits local programing. Access points (subject, type of reference work, course) and database structure and user screens are highlighted. (EJS)

  10. China Connections Reference Book.

    ERIC Educational Resources Information Center

    Kalat, Marie B.; Hoermann, Elizabeth F.

    This reference book focuses on six aspects of the geography of the People's Republic of China. They are: territory, governing units, population and land use, waterways, land forms, and climates. Designed as a primary reference, the book explains how the Chinese people and their lifestyles are affected by China's geography. Special components…

  11. Rethinking Virtual Reference

    ERIC Educational Resources Information Center

    Tenopir, Carol

    2004-01-01

    Virtual reference services seem a natural extension of libraries digital collections and the emphasis on access to the library anytime, anywhere. If patrons use the library from home, it makes sense to provide them with person-to-person online reference. The Library of Congress (LC), OCLC, and several large library systems have developed and…

  12. Reference Point Heterogeneity.

    PubMed

    Terzi, Ayse; Koedijk, Kees; Noussair, Charles N; Pownall, Rachel

    2016-01-01

    It is well-established that, when confronted with a decision to be taken under risk, individuals use reference payoff levels as important inputs. The purpose of this paper is to study which reference points characterize decisions in a setting in which there are several plausible reference levels of payoff. We report an experiment, in which we investigate which of four potential reference points: (1) a population average payoff level, (2) the announced expected payoff of peers in a similar decision situation, (3) a historical average level of earnings that others have received in the same task, and (4) an announced anticipated individual payoff level, best describes decisions in a decontextualized risky decision making task. We find heterogeneity among individuals in the reference points they employ. The population average payoff level is the modal reference point, followed by experimenter's stated expectation of a participant's individual earnings, followed in turn by the average earnings of other participants in previous sessions of the same experiment. A sizeable share of individuals show multiple reference points simultaneously. The reference point that best fits the choices of the individual is not affected by a shock to her income.

  13. Comparison of PWR - Burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results

    SciTech Connect

    Oberle, P.; Broeders, C. H. M.; Dagan, R.

    2006-07-01

    The increasing tendency towards fuel lifetime extension in thermal nuclear reactors motivated validation work for available evaluation tools for nuclear fuel burnup calculations. In this study two deterministic codes with different transport solvers and one Monte Carlo method are investigated. The code system KAPROS/KARBUS uses the classical deterministic First Collision Probability method utilizing a cylinderized Wigner-Seitz cell. In the SCALES.0/TRITON/NEWT code the Extended Step Characteristic method is applied. In a first step the two deterministic codes are compared with experimental results from the KWO-Isotope Correlation Experiment up to 30 MWD/kg HM burnup, published in 1981. Two pin cell calculations are analyzed by comparison of calculated and experimental results for important heavy isotope vectors. The results are very satisfactory. Subsequently, further validation at higher burnup (< 80 MWD/kg HM) is provided by comparison of the two deterministic codes and the Monte Carlo based burnup code MONTEBURNS for PWR UO{sub 2} fuel assembly calculations. Possible reasons for differences in the results are analyzed and discussed. Especially the influence of cross section data and processing is presented. (authors)

  14. Effect of surface state on the oxidation behavior of welded 308L in simulated nominal primary water of PWR

    NASA Astrophysics Data System (ADS)

    Ming, Hongliang; Zhang, Zhiming; Wang, Jiazhen; Zhu, Ruolin; Ding, Jie; Wang, Jianqiu; Han, En-Hou; Ke, Wei

    2015-05-01

    The oxidation behavior of 308L weld metal (WM) with different surface state in the simulated nominal primary water of pressurized water reactor (PWR) was studied by scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), X-ray diffraction (XRD) analyzer and X-ray photoelectron spectroscopy (XPS). After 480 h immersion, a duplex oxide film composed of a Fe-rich outer layer (Fe3O4, Fe2O3 and a small amount of NiFe2O4, Ni(OH)2, Cr(OH)3 and (Ni, Fe)Cr2O4) and a Cr-rich inner layer (FeCr2O4 and NiCr2O4) can be formed on the 308L WM samples with different surface state. The surface state has no influence on the phase composition of the oxide films but obviously affects the thickness of the oxide films and the morphology of the oxides (number & size). With increasing the density of dislocations and subgrain boundaries in the cold-worked superficial layer, the thickness of the oxide film, the number and size of the oxides decrease.

  15. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    SciTech Connect

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  16. Criticality evaluation of control component credited mixed zone spent and fresh fuel storage in high density PWR racks

    SciTech Connect

    Bilovsky, V.; Redmond, E.; Walker, C.; Ivanov, K.

    2006-07-01

    To expand the set of assemblies that qualify for storage in high-density racks, a mixed zone analysis may be performed where repeating pattern configurations within the rack are prescribed. In a mixed zone analysis, assemblies that are more reactive (low burnup) are stored adjacent to less reactive (highly burned) assemblies, thereby meeting the same overall criticality requirements as with the uniform burnup/enrichment analysis. The Arkansas Nuclear One (ANO) Plant has faced several challenges with respect to their spent fuel storage that reach beyond simply the number of spent fuel assemblies and available storage cells. These issues have resulted in the need for ANO to use an advanced storage strategy. In addition to using the mixed zone burnup approach in the high-density racks, ANO also proposed a new solution involving credit for control components in the spent fuel pool. ANO submitted an amendment of their spent fuel pool technical specifications to the Nuclear Regulatory Commission (NRC) based on the evaluation performed by Holtec International that was subsequently approved. This paper presents a description of the overall methodology used for supporting the submittal, and provides further discussion regarding the reactivity effect of control rods in a PWR spent fuel pool. (authors)

  17. Linking Grain Boundary Microstructure to Stress Corrosion Cracking of Cold Rolled Alloy 690 in PWR Primary Water

    SciTech Connect

    Bruemmer, Stephen M.; Olszta, Matthew J.; Toloczko, Mychailo B.; Thomas, Larry E.

    2012-10-01

    Grain boundary microstructures and microchemistries are examined in cold-rolled alloy 690 tubing and plate materials and comparisons are made to intergranular stress corrosion cracking (IGSCC) behavior in PWR primary water. Chromium carbide precipitation is found to be a key aspect for materials in both the mill annealed and thermally treated conditions. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG carbide distribution. For the same degree of cold rolling, alloys with few IG precipitates exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal that cracked carbides appeared to blunt propagation of IGSCC cracks in many cases. Preliminary results suggest that the localized grain boundary strains and stresses produced during cold rolling promote IGSCC susceptibility and not the cracked carbides and voids.

  18. Improvement of the thermal margins in the Swedish Ringhals-3 PWR by introducing new fuel assemblies with thorium

    SciTech Connect

    Lau, C. W.; Demaziere, C.; Nylen, H.; Sandberg, U.

    2012-07-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)

  19. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    SciTech Connect

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  20. Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

    DTIC Science & Technology

    1987-06-01

    Work 98 References 104 Appendices I 7 A Compatibility of Liquid lead at 750 Degrees Fahrenheit with Zircaloy-2, Inconel, and 316 Stainless Steel . 108...zircaloy, low carbon steel and 316 stainless steel in a liquid lead bath at 750 degrees F (398.89 degrees C). 2.3.2 Titanium Test Tube The Titanium Test...Charging system is connected to the loop through 1/8 inch/3.18 mm OD 316 stainless steel tubing using a 1/8 inch X 3/8 inch X 3/8 inch (3.18 mm X 9.53 mm X

  1. Genetics Home Reference: nephronophthisis

    MedlinePlus

    ... these are often referred to as nephronophthisis -associated ciliopathies. For example, Senior-Løken syndrome is characterized by ... Nephronophthisis Patient Support and Advocacy Resources (2 links) Ciliopathy Alliance National Kidney Foundation GeneReviews (1 link) Nephronophthisis ...

  2. Value of Information References

    SciTech Connect

    Morency, Christina

    2014-12-12

    This file contains a list of relevant references on value of information (VOI) in RIS format. VOI provides a quantitative analysis to evaluate the outcome of the combined technologies (seismology, hydrology, geodesy) used to monitor Brady's Geothermal Field.

  3. EPA QUICK REFERENCE GUIDES

    EPA Science Inventory

    EPA Quick Reference Guides are compilations of information on chemical and biological terrorist agents. The information is presented in consistent format and includes agent characteristics, release scenarios, health and safety data, real-time field detection, effect levels, samp...

  4. Selecting a reference object.

    PubMed

    Miller, Jared E; Carlson, Laura A; Hill, Patrick L

    2011-07-01

    One way to describe the location of an object is to relate it to another object. Often there are many nearby objects, each of which could serve as a candidate to be the reference object. A common theoretical assumption is that features that make a given object salient relative to the candidate set are instrumental in determining which is selected. The current research tests this assumption, assessing the relative importance of spatial, perceptual, and functional-interactive features. Three experiments demonstrated that spatial features have the strongest influence on reference object selection, with the perceptual feature of color playing no significant role. Functional-interactive features were shown to be spatially dependent, having an influence only when the spatial configuration enabled an interaction between the located object and the reference object. These findings challenge the common perspective that salience in and of itself dictates reference object selection and argue for a reliance on spatial features.

  5. Genetics Home Reference

    MedlinePlus

    ... MENU Toggle navigation Home Page Search Share: Email Facebook Twitter Home Health Conditions Genes Chromosomes & mtDNA Resources Help Me Understand Genetics Genetics Home Reference provides consumer-friendly information about the effects of genetic variation ...

  6. Enterprise Reference Library

    NASA Technical Reports Server (NTRS)

    Bickham, Grandin; Saile, Lynn; Havelka, Jacque; Fitts, Mary

    2011-01-01

    Introduction: Johnson Space Center (JSC) offers two extensive libraries that contain journals, research literature and electronic resources. Searching capabilities are available to those individuals residing onsite or through a librarian s search. Many individuals have rich collections of references, but no mechanisms to share reference libraries across researchers, projects, or directorates exist. Likewise, information regarding which references are provided to which individuals is not available, resulting in duplicate requests, redundant labor costs and associated copying fees. In addition, this tends to limit collaboration between colleagues and promotes the establishment of individual, unshared silos of information The Integrated Medical Model (IMM) team has utilized a centralized reference management tool during the development, test, and operational phases of this project. The Enterprise Reference Library project expands the capabilities developed for IMM to address the above issues and enhance collaboration across JSC. Method: After significant market analysis for a multi-user reference management tool, no available commercial tool was found to meet this need, so a software program was built around a commercial tool, Reference Manager 12 by The Thomson Corporation. A use case approach guided the requirements development phase. The premise of the design is that individuals use their own reference management software and export to SharePoint when their library is incorporated into the Enterprise Reference Library. This results in a searchable user-specific library application. An accompanying share folder will warehouse the electronic full-text articles, which allows the global user community to access full -text articles. Discussion: An enterprise reference library solution can provide a multidisciplinary collection of full text articles. This approach improves efficiency in obtaining and storing reference material while greatly reducing labor, purchasing and

  7. Membrane reference electrode

    DOEpatents

    Redey, L.; Bloom, I.D.

    1988-01-21

    A reference electrode utilizes a small thin, flat membrane of a highly conductive glass placed on a small diameter insulator tube having a reference material inside in contact with an internal voltage lead. When the sensor is placed in a non-aqueous ionic electrolytic solution, the concentration difference across the glass membrane generates a low voltage signal in precise relationship to the concentration of the species to be measured, with high spatial resolution. 2 figs.

  8. Membrane reference electrode

    DOEpatents

    Redey, Laszlo; Bloom, Ira D.

    1989-01-01

    A reference electrode utilizes a small thin, flat membrane of a highly conductive glass placed on a small diameter insulator tube having a reference material inside in contact with an internal voltage lead. When the sensor is placed in a non-aqueous ionic electrolytic solution, the concentration difference across the glass membrane generates a low voltage signal in precise relationship to the concentration of the species to be measured with high spatial resolution.

  9. Precision displacement reference system

    DOEpatents

    Bieg, Lothar F.; Dubois, Robert R.; Strother, Jerry D.

    2000-02-22

    A precision displacement reference system is described, which enables real time accountability over the applied displacement feedback system to precision machine tools, positioning mechanisms, motion devices, and related operations. As independent measurements of tool location is taken by a displacement feedback system, a rotating reference disk compares feedback counts with performed motion. These measurements are compared to characterize and analyze real time mechanical and control performance during operation.

  10. Reference Man anatomical model

    SciTech Connect

    Cristy, M.

    1994-10-01

    The 70-kg Standard Man or Reference Man has been used in physiological models since at least the 1920s to represent adult males. It came into use in radiation protection in the late 1940s and was developed extensively during the 1950s and used by the International Commission on Radiological Protection (ICRP) in its Publication 2 in 1959. The current Reference Man for Purposes of Radiation Protection is a monumental book published in 1975 by the ICRP as ICRP Publication 23. It has a wealth of information useful for radiation dosimetry, including anatomical and physiological data, gross and elemental composition of the body and organs and tissues of the body. The anatomical data includes specified reference values for an adult male and an adult female. Other reference values are primarily for the adult male. The anatomical data include much data on fetuses and children, although reference values are not established. There is an ICRP task group currently working on revising selected parts of the Reference Man document.

  11. Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data

    SciTech Connect

    1997-11-01

    Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

  12. 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

  13. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    SciTech Connect

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  14. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    SciTech Connect

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  15. Setting reference targets

    SciTech Connect

    Ruland, R.E.

    1997-04-01

    Reference Targets are used to represent virtual quantities like the magnetic axis of a magnet or the definition of a coordinate system. To explain the function of reference targets in the sequence of the alignment process, this paper will first briefly discuss the geometry of the trajectory design space and of the surveying space, then continue with an overview of a typical alignment process. This is followed by a discussion on magnet fiducialization. While the magnetic measurement methods to determine the magnetic centerline are only listed (they will be discussed in detail in a subsequent talk), emphasis is given to the optical/mechanical methods and to the task of transferring the centerline position to reference targets.

  16. Rod consolidation of RG and E's (Rochester Gas and Electric Corporation) spent PWR (pressurized water reactor) fuel

    SciTech Connect

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister.

  17. 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

  18. Crack initiation testing of thimble tube material under PWR conditions to determine a stress threshold for IASCC

    NASA Astrophysics Data System (ADS)

    Bosch, R. W.; Vankeerberghen, M.; Gérard, R.; Somville, F.

    2015-06-01

    IASCC (Irradiation Assisted Stress Corrosion Cracking) crack initiation tests have been carried out on thimble tube material retrieved from a Belgian PWR. The crack initiation tests were carried out by constant load testing of thimble tube specimens at different stress levels. The time-to-failure was determined as a function of the applied stress to find a stress threshold under which no stress corrosion cracking will take place. The thimble tube was made of 316L cold-worked stainless steel and the dose profile along the thimble tube ranges from 45 to 80 dpa. This allows adding crack initiation data for dose values that have not been significantly reported, i.e. in the range of 45-55 dpa and at 80 dpa. The results can be used to determine whether the stress under which no IASCC occurs saturates for a dose larger than 30 dpa or whether a small further threshold decrease with dose can be observed. Over a period of four years, more than 40 specimens have been tested with doses ranging from 45 to 80 dpa at stress levels between 40% and 70% of the irradiated yield stress. Fracture occurred at all stress levels (but not all specimens) although the time-to-failure increased with decreasing stress. The results show that intergranular cracking was the main fracture mode in all failed O-rings. Three of six 80 dpa O-rings subjected to 40% and 45% of the yield stress did not fail after six months of testing. Based on these results and a comparison with literature data, an apparent stress limit for IASCC could be estimated at 40% of the irradiated yield stress.

  19. Multifunctional reference electrode

    DOEpatents

    Redey, L.; Vissers, D.R.

    1981-12-30

    A multifunctional, low mass reference electrode of a nickel tube, thermocouple means inside the nickel tube electrically insulated therefrom for measuring the temperature thereof, a housing surrounding the nickel tube, an electrolyte having a fixed sulfide ion activity between the housing and the outer surface of the nickel tube forming the nickel/nickel sulfide/sulfide half-cell are described. An ion diffusion barrier is associated with the housing in contact with the electrolyte. Also disclosed is a cell using the reference electrode to measure characteristics of a working electrode.

  20. Multifunctional reference electrode

    DOEpatents

    Redey, Laszlo; Vissers, Donald R.

    1983-01-01

    A multifunctional, low mass reference electrode of a nickel tube, thermocouple means inside the nickel tube electrically insulated therefrom for measuring the temperature thereof, a housing surrounding the nickel tube, an electrolyte having a fixed sulfide ion activity between the housing and the outer surface of the nickel tube forming the nickel/nickel sulfide/sulfide half-cell. An ion diffusion barrier is associated with the housing in contact with the electrolyte. Also disclosed is a cell using the reference electrode to measure characteristics of a working electrode.

  1. IERS Reference System.

    NASA Astrophysics Data System (ADS)

    Yokoyama, K.

    Present circumstances related to IERS activities are described from various points of view. The NASA Dynamics of Solid Earth (DOSE) program and the IERS intensive campaign proposed by J. Dickey of JPL are particularly interesting. It is important to implement international cooperation to establish a fundamental radio reference frame by carrying out global solution based on all geodetic observations, past and future. A precession and nutation model may be determined observationally with an accuracy of 0.2 - 0.3 mas in a few years. Then it will become possible to establish the radio reference frame with this accuracy.

  2. Aluminum reference electrode

    DOEpatents

    Sadoway, D.R.

    1988-08-16

    A stable reference electrode is described for use in monitoring and controlling the process of electrolytic reduction of a metal. In the case of Hall cell reduction of aluminum, the reference electrode comprises a pool of molten aluminum and a solution of molten cryolite, Na[sub 3]AlF[sub 6], wherein the electrical connection to the molten aluminum does not contact the highly corrosive molten salt solution. This is accomplished by altering the density of either the aluminum (decreasing the density) or the electrolyte (increasing the density) so that the aluminum floats on top of the molten salt solution. 1 fig.

  3. Aluminum reference electrode

    DOEpatents

    Sadoway, Donald R.

    1988-01-01

    A stable reference electrode for use in monitoring and controlling the process of electrolytic reduction of a metal. In the case of Hall cell reduction of aluminum, the reference electrode comprises a pool of molten aluminum and a solution of molten cryolite, Na.sub.3 AlF.sub.6, wherein the electrical connection to the molten aluminum does not contact the highly corrosive molten salt solution. This is accomplished by altering the density of either the aluminum (decreasing the density) or the electrolyte (increasing the density) so that the aluminum floats on top of the molten salt solution.

  4. NASCAP programmer's reference manual

    NASA Technical Reports Server (NTRS)

    Mandell, M. J.; Stannard, P. R.; Katz, I.

    1993-01-01

    The NASA Charging Analyzer Program (NASCAP) is a computer program designed to model the electrostatic charging of complicated three-dimensional objects, both in a test tank and at geosynchronous altitudes. This document is a programmer's reference manual and user's guide. It is designed as a reference to experienced users of the code, as well as an introduction to its use for beginners. All of the many capabilities of NASCAP are covered in detail, together with examples of their use. These include the definition of objects, plasma environments, potential calculations, particle emission and detection simulations, and charging analysis.

  5. Isotopic Generation and Confirmation of the PWR Application Model 

    SciTech Connect

    L.B. Wimmer

    2003-11-10

    The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section 3.5.2.1.1 of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.

  6. Microstructural characterization on intergranular stress corrosion cracking of Alloy 600 in PWR primary water environment

    NASA Astrophysics Data System (ADS)

    Lim, Yun Soo; Kim, Hong Pyo; Hwang, Seong Sik

    2013-09-01

    Stress corrosion cracks in Alloy 600 compact tension specimens tested at 325 °C in a simulated primary water environment of a pressurized water reactor were analyzed using microscopic equipment. Oxygen diffused into the grain boundaries just ahead of the crack tips from the external primary water. As a result of oxygen penetration, Cr oxides were precipitated on the crack tips and the attacked grain boundaries. The oxide layer in the crack interior was revealed to consist of double (inner and outer) layers. Cr oxides were found in the inner layer, with NiO and (Ni,Cr) spinels in the outer layer. Cr depletion (or Ni enrichment) zones were created in the attacked grain boundary, the crack tip, and the interface between the crack and matrix, which means that the formation of Cr oxides was due to the Cr diffusion from the surrounding matrix. The oxygen penetration and resultant metallurgical changes around the crack tip are believed to be significant factors affecting the PWSCC initiation and growth behaviors of Alloy 600. For interpretation of color in Fig. 4, the reader is referred to the web version of this article.

  7. Clay Generic Disposal System Model - Sensitivity Analysis for 32 PWR Assembly Canisters (+2 associated model files).

    SciTech Connect

    Morris, Edgar

    2014-10-01

    The Used Fuel Disposition Campaign (UFDC), as part of the DOE Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technology program (FCT) is investigating the disposal of high level radioactive waste (HLW) and spent nuclear fuela (SNF) in a variety of geologic media. The feasibility of disposing SNF and HLW in clay media has been investigated and has been shown to be promising [Ref. 1]. In addition the disposal of these wastes in clay media is being investigated in Belgium, France, and Switzerland. Thus, Argillaceous media is one of the environments being considered by UFDC. As identified by researchers at Sandia National Laboratory, potentially suitable formations that may exist in the U.S. include mudstone, clay, shale, and argillite formations [Ref. 1]. These formations encompass a broad range of material properties. In this report, reference to clay media is intended to cover the full range of material properties. This report presents the status of the development of a simulation model for evaluating the performance of generic clay media. The clay Generic Disposal System Model (GDSM) repository performance simulation tool has been developed with the flexibility to evaluate not only different properties, but different waste streams/forms and different repository designs and engineered barrier configurations/ materials that could be used to dispose of these wastes.

  8. Multimedia Reference Tools.

    ERIC Educational Resources Information Center

    Holzberg, Carol S.

    2001-01-01

    Presents suggestions for content-rich classroom encyclopedias on CO-ROM and DVD, including: the Encarta Reference Suite 2001; the 2001 Grolier Multimedia Encyclopedia, School Edition; the Britannica 2001 DVD; and the World Book 2001 Deluxe Edition, v5.0. (SM)

  9. Digital Reference Service.

    ERIC Educational Resources Information Center

    Mon, Lorri

    2000-01-01

    Discusses the increasing demand for digital reference services from government Web sites via email, and describes a partnership between the Government Printing Office and the federal depository library at the University of Illinois at Chicago to create electronic access to the Department of State Foreign Affairs Network (DOSFAN). (Author/LRW)

  10. Reflections on Reference Services.

    ERIC Educational Resources Information Center

    Brandt, Kerryn A.; And Others

    1996-01-01

    Describes programmatic changes in reference services at the Johns Hopkins University (Maryland) medical library and speculates on the future. Topics include institutional restructuring and consolidation; improvements in technology infrastructure; external economic pressure; and fiscal accountability, including library funding and cost center…

  11. Questions in Reference Interviews.

    ERIC Educational Resources Information Center

    White, Marilyn Domas

    1998-01-01

    Characterizes the questioning behavior in reference interviews preceding delegated online searches of bibliographic databases and relates it to questioning behavior in other types of interviews/settings. Compares questions asked by the information specialist and those asked by the client; findings show the information specialist dominates the…

  12. The Reference Encounter Model.

    ERIC Educational Resources Information Center

    White, Marilyn Domas

    1983-01-01

    Develops model of the reference interview which explicitly incorporates human information processing, particularly schema ideas presented by Marvin Minsky and other theorists in cognitive processing and artificial intelligence. Questions are raised concerning use of content analysis of transcribed verbal protocols as methodology for studying…

  13. Reference-Dependent Sympathy

    ERIC Educational Resources Information Center

    Small, Deborah A.

    2010-01-01

    Natural disasters and other traumatic events often draw a greater charitable response than do ongoing misfortunes, even those that may cause even more widespread misery, such as famine or malaria. Why is the response disproportionate to need? The notion of reference dependence critical to Prospect Theory (Kahneman & Tversky, 1979) maintains that…

  14. Virtual Reference Services.

    ERIC Educational Resources Information Center

    Brewer, Sally

    2003-01-01

    As the need to access information increases, school librarians must create virtual libraries. Linked to reliable reference resources, the virtual library extends the physical collection and library hours and lets students learn to use Web-based resources in a protected learning environment. The growing number of virtual schools increases the need…

  15. Reference Collections and Standards.

    ERIC Educational Resources Information Center

    Winkel, Lois

    1999-01-01

    Reviews six reference materials for young people: "The New York Public Library Kid's Guide to Research"; "National Audubon Society First Field Guide. Mammals"; "Star Wars: The Visual Dictionary"; "Encarta Africana"; "World Fact Book, 1998"; and "Factastic Book of 1001 Lists". Includes ordering information.(AEF)

  16. A GUJARATI REFERENCE GRAMMAR.

    ERIC Educational Resources Information Center

    CARDONA, GEORGE

    THIS REFERENCE GRAMMAR WAS WRITTEN TO FILL THE NEED FOR AN UP-TO-DATE ANALYSIS OF THE MODERN LANGUAGE SUITABLE FOR LANGUAGE LEARNERS AS WELL AS LINGUISTS. THE AUTHOR LISTS IN THE INTRODUCTION THOSE STUDIES PREVIOUS TO THIS ONE WHICH MAY BE OF INTEREST TO THE READER. INCLUDED IN HIS ANALYSIS OF THE LANGUAGE ARE MAJOR CHAPTERS ON--(1) PHONOLOGY, (2)…

  17. Reference Sources for Nursing

    ERIC Educational Resources Information Center

    Nursing Outlook, 1976

    1976-01-01

    The ninth revision (including a Canadian supplement) of a list of nursing reference works lists items in the following sections: abstract journals, audiovisuals, bibliographies, dictionaries, directories, drug lists and pharmacologies, educational programs, histories, indexes, legal guides, library administration and organization, research grants,…

  18. Reference Sources for Nursing

    ERIC Educational Resources Information Center

    Nursing Outlook, 1978

    1978-01-01

    The tenth revision of a list of reference works for nurses, revised by a committee of the Interagency Council on Library Resources for Nursing, listed by type of publication as abstract journals, audiovisuals, bibliographies, books, dictionaries, directories, pharmacologies, indexes, guides, and so on. (MF)

  19. Selecting a Reference Object

    ERIC Educational Resources Information Center

    Miller, Jared E.; Carlson, Laura A.; Hill, Patrick L.

    2011-01-01

    One way to describe the location of an object is to relate it to another object. Often there are many nearby objects, each of which could serve as a candidate to be the reference object. A common theoretical assumption is that features that make a given object salient relative to the candidate set are instrumental in determining which is selected.…

  20. Isotope reference materials

    USGS Publications Warehouse

    Coplen, Tyler B.

    2010-01-01

    Measurement of the same isotopically homogeneous sample by any laboratory worldwide should yield the same isotopic composition within analytical uncertainty. International distribution of light element isotopic reference materials by the International Atomic Energy Agency and the U.S. National Institute of Standards and Technology enable laboratories to achieve this goal.

  1. Reference Model Development

    SciTech Connect

    Jepsen, Richard

    2011-11-02

    Presentation from the 2011 Water Peer Review in which principal investigator discusses project progress to develop a representative set of Reference Models (RM) for the MHK industry to develop baseline cost of energy (COE) and evaluate key cost component/system reduction pathways.

  2. Hospitality Services Reference Book.

    ERIC Educational Resources Information Center

    Texas Tech Univ., Lubbock. Home Economics Curriculum Center.

    This reference book provides information needed by employees in hospitality services occupations. It includes 29 chapters that cover the following topics: the hospitality services industry; professional ethics; organization and management structures; safety practices and emergency procedures; technology; property maintenance and repair; purchasing…

  3. An Amharic Reference Grammar.

    ERIC Educational Resources Information Center

    Leslau, Wolf

    This reference grammar presents a structural description of the orthography, phonology, morphology, and syntax of Amharic, the national language of Ethiopia. The Amharic material in this work, designed to prepare the student for speaking and reading the language, appears in both Amharic script and phonetic transcription. See ED 012 044-5 for the…

  4. The Unreliability of References

    ERIC Educational Resources Information Center

    Barden, Dennis M.

    2008-01-01

    When search consultants, like the author, are invited to propose their services in support of a college or university seeking new leadership, they are generally asked a fairly standard set of questions. But there is one question that they find among the most difficult to answer: How do they check a candidate's references to ensure that they know…

  5. Generating Multimodal References

    ERIC Educational Resources Information Center

    van der Sluis, Ielka; Krahmer, Emiel

    2007-01-01

    This article presents a new computational model for the generation of multimodal referring expressions (REs), based on observations in human communication. The algorithm is an extension of the graph-based algorithm proposed by Krahmer, van Erk, and Verleg (2003) and makes use of a so-called Flashlight Model for pointing. The Flashlight Model…

  6. International reference ionosphere 1990

    NASA Technical Reports Server (NTRS)

    Bilitza, Dieter; Rawer, K.; Bossy, L.; Kutiev, I.; Oyama, K.-I.; Leitinger, R.; Kazimirovsky, E.

    1990-01-01

    The International Reference Ionosphere 1990 (IRI-90) is described. IRI described monthly averages of the electron density, electron temperature, ion temperature, and ion composition in the altitude range from 50 to 1000 km for magnetically quiet conditions in the non-auroral ionosphere. The most important improvements and new developments are summarized.

  7. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  8. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  9. A High Fidelity Multiphysics Framework for Modeling CRUD Deposition on PWR Fuel Rods

    NASA Astrophysics Data System (ADS)

    Walter, Daniel John

    Corrosion products on the fuel cladding surfaces within pressurized water reactor fuel assemblies have had a significant impact on reactor operation. These types of deposits are referred to as CRUD and can lead to power shifts, as a consequence of the accumulation of solid boron phases on the fuel rod surfaces. Corrosion deposits can also lead to fuel failure resulting from localized corrosion, where the increased thermal resistance of the deposit leads to higher cladding temperatures. The prediction of these occurrences requires a comprehensive model of local thermal hydraulic and chemical processes occurring in close proximity to the cladding surface, as well as their driving factors. Such factors include the rod power distribution, coolant corrosion product concentration, as well as the feedbacks between heat transfer, fluid dynamics, chemistry, and neutronics. To correctly capture the coupled physics and corresponding feedbacks, a high fidelity framework is developed that predicts three-dimensional CRUD deposition on a rod-by-rod basis. Multiphysics boundary conditions resulting from the coupling of heat transfer, fluid dynamics, coolant chemistry, CRUD deposition, neutron transport, and nuclide transmutation inform the CRUD deposition solver. Through systematic parametric sensitivity studies of the CRUD property inputs, coupled boundary conditions, and multiphysics feedback mechanisms, the most important variables of multiphysics CRUD modeling are identified. Moreover, the modeling framework is challenged with a blind comparison of plant data to predictions by a simulation of a sub-assembly within the Seabrook nuclear plant that experienced CRUD induced fuel failures. The physics within the computational framework are loosely coupled via an operator-splitting technique. A control theory approach is adopted to determine the temporal discretization at which to execute a data transfer from one physics to another. The coupled stepsize selection is viewed as a

  10. Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

    SciTech Connect

    Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

    2012-07-01

    The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

  11. In-situ measurement of the effect of LiOH on the stability of fuel cladding oxide film in simulated PWR primary water environment

    SciTech Connect

    Saario, T.; Taehtinen, S.; Piippo, J.; Kukkonen, J.J.V.

    1995-12-31

    Development of new improved fuel cladding materials is a long process, partly because of the lack of fast and reliable in-situ techniques for investigations of cladding degradation in high temperature water environments. This paper describes results gained with the Contact Electric Resistance (CER) technique on the electric resistance of oxides growing on zirconium based fuel cladding materials. LiOH decreased the electric resistance of the oxides when about 70 ppm was injected in PWR water at 300 C. When PWR water contains boric acid and LiOH from the beginning of the exposure the fuel cladding material is covered by a hydroxide layer that protects the amorphous oxide layer and later hinders the increase of the resistance of the crystalline oxide layer. The dependency of electric resistance of the oxides on LiOH concentration is shown to correlate inversely with the effect of LiOH on weight gain. The kinetics of the breakdown process of electric resistance indicate that a phase transformation rather than a diffusion limited process is the mechanism of degradation. The growth rate of the electric resistance of the oxide in the early stage of oxide formation is shown to correlate well with the in-reactor weight gain of similar alloys. In-situ monitoring of the electric resistance of the oxide during growth is shown to give the same ranking order as long term in-reactor weight gain tests, but in a fraction of the testing time needed for weight gain tests.

  12. OSH technical reference manual

    SciTech Connect

    Not Available

    1993-11-01

    In an evaluation of the Department of Energy (DOE) Occupational Safety and Health programs for government-owned contractor-operated (GOCO) activities, the Department of Labor`s Occupational Safety and Health Administration (OSHA) recommended a technical information exchange program. The intent was to share written safety and health programs, plans, training manuals, and materials within the entire DOE community. The OSH Technical Reference (OTR) helps support the secretary`s response to the OSHA finding by providing a one-stop resource and referral for technical information that relates to safe operations and practice. It also serves as a technical information exchange tool to reference DOE-wide materials pertinent to specific safety topics and, with some modification, as a training aid. The OTR bridges the gap between general safety documents and very specific requirements documents. It is tailored to the DOE community and incorporates DOE field experience.

  13. Open SHMEM Reference Implementation

    SciTech Connect

    Pritchard, Howard; Curtis, Anthony; Welch, Aaron; Fridley, Andrew

    2016-05-12

    OpenSHMEM is an effort to create a specification for a standardized API for parallel programming in the Partitioned Global Address Space. Along with the specification the project is also creating a reference implementation of the API. This implementation attempts to be portable, to allow it to be deployed in multiple environments, and to be a starting point for implementations targeted to particular hardware platforms. It will also serve as a springboard for future development of the API.

  14. Range Reference Notebook

    DTIC Science & Technology

    2006-10-15

    rifle grenade (inert), tin can lid, 15” tent peg 3 Table FRD-7. Fort Ritchie Sector 3 Representative Examples of Non-MEC Clutter Description 1/2...Appendix B—Indirect Fire Range Examples SITES ( ADI ) Adak Naval Air Facility, AK, Mitt Lake Mortar Range (FRI) Fort Ritchie...example range. B- ADI -1 Indirect-Fire Range,: Adak, AK, Mitt Lake Mortar Range Impact Area Site-Specific References – Adak NAF Foster Wheeler

  15. Precision optical reference frequencies

    NASA Astrophysics Data System (ADS)

    Riehle, Fritz; Schnatz, Harald; Zinner, G.; Trebst, Tilmann; Helmcke, Juergen

    1999-05-01

    Optical reference frequencies are provided by lasers of which the frequencies are stabilized to suitable absorption lines. Presently, twelve reference frequencies/wavelengths within the wavelengths range from 243 nm to 10.3 micrometers are recommended by the International Committee of Weights and Measures as references for the realization of the meter and scientific applications. As typical examples, we describe a diode-pumped, frequency doubled YAG-laser stabilized to an absorption line of molecular iodine and a Ca-stabilized laser. The latter one has been developed in two versions, a transportable system utilizing a small beam of thermal Ca atoms and a stationary standard based on laser cooled and trapped Ca atoms. The frequency of the Ca standard based on cold Ca atoms has been measured by a frequency chain allowing a phase-coherent comparison against the primary standard of time and frequency, the caesium clock. Its value is vCa equals 455 986 240 494.13 kHz with a relative standard uncertainty of 2.5 (DOT) 10-13.

  16. Celestial Reference Frames

    NASA Astrophysics Data System (ADS)

    Jacobs, Christopher S.

    2013-03-01

    Concepts and Background: This paper gives an overview of modern celestial reference frames as realized at radio frequencies using the Very Long baseline Interferometry (VLBI) technique. We discuss basic celestial reference frame concepts, desired properties, and uses. We review the networks of antennas used for this work. We briefly discuss the history of the science of astrometry touching upon the discovery of precession, proper motion, nutation, and parallax, and the field of radio astronomy. Building Celestial Frames: Next, we discuss the multi-step process of building a celestial frame: First candidate sources are identified based on point-like properties from single dish radio telescopes surveys. Second, positions are refined using connected element interferometers such as the Very Large Array, and the ATCA. Third, positions of approximately milli-arcsecond (mas) accuracy are determined using intercontinental VLBI surveys. Fourth, sub-mas positions are determined by multiyear programs using intercontinental VLBI. These sub-mas sets of positions are then verified by multiple teams in preparation for release to non-specialists in the form of an official IAU International Celestial Reference Frame (ICRF). The process described above has until recently been largely restricted to work at S/X-band (2.3/8.4 GHz). However, in the last decade sub-mas work has expanded to include celestial frames at K-band (24 GHz), Ka-band (32 GHz), and Q-band (43 GHz). While these frames currently have the disadvantage of far smaller data sets, the astrophysical quality of the sources themselves improves at these higher frequencies and thus make these frequencies attractive for realizations of celestial reference frames. Accordingly, we review progress at these higher frequency bands. Path to the Future: We discuss prospects for celestial reference frames over the next decade. We present an example of an error budget for astrometric VLBI and discuss the budget's use as a tool for

  17. Celestial Reference Frame

    NASA Astrophysics Data System (ADS)

    Jacobs, Christopher S.

    2013-09-01

    Concepts and Background: This paper gives an overview of modern celestial reference frames as realized at radio frequencies using the Very Long baseline Interferometry (VLBI) technique. We discuss basic celestial reference frame concepts, desired properties, and uses. We review the networks of antennas used for this work. We briefly discuss the history of the science of astrometry touching upon the discovery of precession, proper motion, nutation, and parallax, and the field of radio astronomy. Building Celestial Frames: Next, we discuss the multi-step process of building a celestial frame: First candidate sources are identified based on point-like properties from single dish radio telescopes surveys. Second, positions are refined using connected element interferometers such as the Very Large Array, and the ATCA. Third, positions of approximately milli-arcsecond (mas) accuracy are determined using intercontinental VLBI surveys. Fourth, sub-mas positions are determined by multiyear programs using intercontinental VLBI. These sub-mas sets of positions are then verified by multiple teams in preparation for release to non-specialists in the form of an official IAU International Celestial Reference Frame (ICRF). The process described above has until recently been largely restricted to work at S/X-band (2.3/8.4 GHz). However, in the last decade sub-mas work has expanded to include celestial frames at K-band (24 GHz), Ka-band (32 GHz), and Q-band (43 GHz). While these frames currently have the disadvantage of far smaller data sets, the astrophysical quality of the sources themselves improves at these higher frequencies and thus make these frequencies attractive for realizations of celestial reference frames. Accordingly, we review progress at these higher frequency bands. Path to the Future: We discuss prospects for celestial reference frames over the next decade. We present an example of an error budget for astrometric VLBI and discuss the budget's use as a tool for

  18. Comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe PWR vessel

    SciTech Connect

    Fabry, A.; Chaouadi, R.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rosinski, S.T.; Carter, R.G.

    1999-10-01

    The sister pressure vessels at the BR3 and Yankee Rowe PWR plants were operated at lower-than-usual temperature ({approx}260 C) and their plates were austenitized at higher-than-usual temperature ({approx}970 C) -- a heat treatment leading to a coarser microstructure than is typical for the fine grain plates considered in development of USNRC Regulatory Guide 1.99. The surveillance programs provided by Westinghouse for the two plants were limited to the same A302-B plate representative of the Rowe vessel upper shell plate; this material displayed outlier behavior characterized by a 41J. Charpy-V Notch shift significantly larger than predicted by Regulatory Guide 1.99. Because lower irradiation temperature and nickel alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements embodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: (1) The accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively; (2) The BR3 surveillance and vessel testing program; this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, and was trepanned in early 1995; (3) The accelerated irradiations in the Belgian test reactor BR2 of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is contended that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the

  19. The OSMOSE program for the qualification of integral cross sections of actinides: Preliminary results in a PWR-UOx spectrum

    SciTech Connect

    Hudelot, J. P.; Antony, M.; Bernard, D.; Fougeras, P.

    2006-07-01

    -worth of the individual samples. The first experimental results were obtained with a very good reproducibility in 2005 and 2006 in the R1-UO{sub 2} core configuration representative of a PWR UOx standard spectrum. The preliminary results of measurements and comparison to calculational models are reported. (authors)

  20. Terminal Forecast Reference Notebook.

    DTIC Science & Technology

    1979-11-07

    Figure 18 , Surface Map 1230Z, October 24, 1948). 4I 7Cm 19 (2) During late spring, suuer or early autumn high oells some- times move southward into the...TFRF, is an excellent reference for this subject. 0 06 0070 I.N ATC 1. A- 2 AREAI p/ A-2 AREA I 1. General: Most flights briefed by Det 18 are within or...in close proxi- mity to Area I. For this reason, Det 18 personnel must become extremely knowledgeable of Area I. 2. Location and Background: The ROK

  1. Coal data: A reference

    SciTech Connect

    Not Available

    1995-02-01

    This report, Coal Data: A Reference, summarizes basic information on the mining and use of coal, an important source of energy in the US. This report is written for a general audience. The goal is to cover basic material and strike a reasonable compromise between overly generalized statements and detailed analyses. The section ``Supplemental Figures and Tables`` contains statistics, graphs, maps, and other illustrations that show trends, patterns, geographic locations, and similar coal-related information. The section ``Coal Terminology and Related Information`` provides additional information about terms mentioned in the text and introduces some new terms. The last edition of Coal Data: A Reference was published in 1991. The present edition contains updated data as well as expanded reviews and additional information. Added to the text are discussions of coal quality, coal prices, unions, and strikes. The appendix has been expanded to provide statistics on a variety of additional topics, such as: trends in coal production and royalties from Federal and Indian coal leases, hours worked and earnings for coal mine employment, railroad coal shipments and revenues, waterborne coal traffic, coal export loading terminals, utility coal combustion byproducts, and trace elements in coal. The information in this report has been gleaned mainly from the sources in the bibliography. The reader interested in going beyond the scope of this report should consult these sources. The statistics are largely from reports published by the Energy Information Administration.

  2. Molecular biology references.

    PubMed

    2003-05-01

    Many of the units in this manual describe methods and techniques for the cloning, expression, and structural analysis of neural genes and proteins. We assume that users of these protocols have at least some introductory background in recombinant DNA technology (or are working with a collaborator who does); therefore, we have not provided comprehensive coverage of all of these topics, but rather have concentrated on presenting selected techniques that will be of the most interest and use to the general neuroscience laboratory. More comprehensive coverage of these topics can be found in Current Protocols in Molecular Biology (CPMB), which is extensively cross-referenced throughout this manual. These cross-references are summarized in this appendix.

  3. Bulk Site Reference Materials

    SciTech Connect

    Barich, J.J. III; Jones, R.R. Sr.

    1996-12-31

    The selection, manufacture and use of Bulk Site Reference Materials (BSRMs) at hazardous waste sites is discussed. BSRMs are useful in preparing stabilization/solidification (S/S) formulations for soils, ranking competing S/S processes, comparing S/S alternatives to other technologies, and in interpreting data from different test types. BSRMs are large volume samples that are representative of the physical and chemical characteristics of a site soil, and that contain contaminants at reasonably high levels. A successful BSRM is extremely homogeneous and well-characterized. While not representative of any point on the site, they contain the contaminants of the site in the matrices of the site. Design objectives for a BSRM are to produce a material that (1) maintains good fidelity to site matrices and contaminants, and (2) exhibits the lowest possible relative standard deviation.

  4. PASCAL/48 reference manual

    NASA Technical Reports Server (NTRS)

    Knight, J. C.; Hamm, R. W.

    1984-01-01

    PASCAL/48 is a programming language for the Intel MCS-48 series of microcomputers. In particular, it can be used with the Intel 8748. It is designed to allow the programmer to control most of the instructions being generated and the allocation of storage. The language can be used instead of ASSEMBLY language in most applications while allowing the user the necessary degree of control over hardware resources. Although it is called PASCAL/48, the language differs in many ways from PASCAL. The program structure and statements of the two languages are similar, but the expression mechanism and data types are different. The PASCAL/48 cross-compiler is written in PASCAL and runs on the CDC CYBER NOS system. It generates object code in Intel hexadecimal format that can be used to program the MCS-48 series of microcomputers. This reference manual defines the language, describes the predeclared procedures, lists error messages, illustrates use, and includes language syntax diagrams.

  5. Long life reference electrode

    DOEpatents

    Yonco, R.M.; Nagy, Z.

    1987-07-30

    An external, reference electrode is provided for long term use with a high temperature, high pressure system. The electrode is arranged in a vertical, electrically insulative tube with an upper portion serving as an electrolyte reservoir and a lower portion in electrolytic communication with the system to be monitored. The lower end portion includes a flow restriction such as a porous plug to limit the electrolyte release into the system. A piston equalized to the system pressure is fitted into the upper portion of the tube to impart a small incremental pressure to the electrolyte. The piston is selected of suitable size and weight to cause only a slight flow of electrolyte through the porous plug into the high pressure system. This prevents contamination of the electrolyte but is of such small flow rate that operating intervals of a month or more can be achieved. 2 figs.

  6. Long life reference electrode

    DOEpatents

    Yonco, R.M.; Nagy, Z.

    1989-04-04

    An external, reference electrode is provided for long term use with a high temperature, high pressure system. The electrode is arranged in a vertical, electrically insulative tube with an upper portion serving as an electrolyte reservoir and a lower portion in electrolytic communication with the system to be monitored. The lower end portion includes a flow restriction such as a porous plug to limit the electrolyte release into the system. A piston equalized to the system pressure is fitted into the upper portion of the tube to impart a small incremental pressure to the electrolyte. The piston is selected of suitable size and weight to cause only a slight flow of electrolyte through the porous plug into the high pressure system. This prevents contamination of the electrolyte but is of such small flow rate that operating intervals of a month or more can be achieved. 2 figs.

  7. Long life reference electrode

    DOEpatents

    Yonco, Robert M.; Nagy, Zoltan

    1989-01-01

    An external, reference electrode is provided for long term use with a high temperature, high pressure system. The electrode is arranged in a vertical, electrically insulative tube with an upper portion serving as an electrolyte reservior and a lower portion in electrolytic communication with the system to be monitored. The lower end portion includes a flow restriction such as a porous plug to limit the electrolyte release into the system. A piston equalized to the system pressure is fitted into the upper portion of the tube to impart a small incremental pressure to the electrolyte. The piston is selected of suitable size and weight to cause only a slight flow of electrolyte through the porous plug into the high pressure system. This prevents contamination of the electrolyte but is of such small flow rate that operating intervals of a month or more can be achieved.

  8. Tank characterization reference guide

    SciTech Connect

    De Lorenzo, D.S.; DiCenso, A.T.; Hiller, D.B.; Johnson, K.W.; Rutherford, J.H.; Smith, D.J.; Simpson, B.C.

    1994-09-01

    Characterization of the Hanford Site high-level waste storage tanks supports safety issue resolution; operations and maintenance requirements; and retrieval, pretreatment, vitrification, and disposal technology development. Technical, historical, and programmatic information about the waste tanks is often scattered among many sources, if it is documented at all. This Tank Characterization Reference Guide, therefore, serves as a common location for much of the generic tank information that is otherwise contained in many documents. The report is intended to be an introduction to the issues and history surrounding the generation, storage, and management of the liquid process wastes, and a presentation of the sampling, analysis, and modeling activities that support the current waste characterization. This report should provide a basis upon which those unfamiliar with the Hanford Site tank farms can start their research.

  9. On establishing reference values.

    PubMed Central

    Lumsden, J H; Mullen, K

    1978-01-01

    In order to establish a range of reference values for any characteristic one can use Gaussian or nonparametric techniques, whichever are most appropriate. One has the choice of calculating tolerance intervals or percentile intervals. A tolerance interval is said to contain, say 95% of the population with probability, say 0.90. A percentile interval simply simply calculates the values between which 95% of the observations fall. If the data can be said to have a Gaussian distribution, the same precision can be obtained with smaller sample sizes than using the nonparametric techniques. In some cases, data which are not Gaussian can be transformed into a Gaussian form and hence make use of the more efficient Gaussian techniques. In both cases, the data should be checked for outliers or rogue observations and these should be eliminated if the testing procedure fails to imply that they are an integral part of the data. PMID:688072

  10. Nuclear Science References Database

    SciTech Connect

    Pritychenko, B.; Běták, E.; Singh, B.; Totans, J.

    2014-06-15

    The Nuclear Science References (NSR) database together with its associated Web interface, is the world's only comprehensive source of easily accessible low- and intermediate-energy nuclear physics bibliographic information for more than 210,000 articles since the beginning of nuclear science. The weekly-updated NSR database provides essential support for nuclear data evaluation, compilation and research activities. The principles of the database and Web application development and maintenance are described. Examples of nuclear structure, reaction and decay applications are specifically included. The complete NSR database is freely available at the websites of the National Nuclear Data Center (http://www.nndc.bnl.gov/nsr) and the International Atomic Energy Agency (http://www-nds.iaea.org/nsr)

  11. International Reference Ionosphere -2010

    NASA Astrophysics Data System (ADS)

    Bilitza, Dieter; Reinisch, Bodo

    The International Reference Ionosphere 2010 includes several important improvements and ad-ditions. This presentation introduces these changes and discusses their benefits. The electron and ion density profiles for the bottomside ionosphere will be significantly improved by using more ionosonde data as well as photochemical considerations. As an additional lower iono-sphere parameter IRI-2010 will include the transition height from molecular to cluster ions. At the F2 peak Neural Net models for the peak density and the propagation factor M3000F2, which is related to the F2 peak height, are introduced as new options. At high latitudes the model will benefit from the introduction of auroral oval boundaries and their variation with magnetic activity. Regarding the electron temperature, IRI-2010 now models variations with solar activity. The homepage for the IRI project is at http://IRI.gsfc.nasa.gov/.

  12. Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly

    NASA Astrophysics Data System (ADS)

    da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2014-04-01

    Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

  13. 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.

  14. PWR FLECHT SEASET 21-rod bundle flow blockage task data and analysis report. NRC/EPRI/Westinghouse Report No. 11. Appendices K-P

    SciTech Connect

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  15. On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

    NASA Astrophysics Data System (ADS)

    Restani, R.; Horvath, M.; Goll, W.; Bertsch, J.; Gavillet, D.; Hermann, A.; Martin, M.; Walker, C. T.

    2016-12-01

    Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

  16. Effects of long-term thermal aging on the stress corrosion cracking behavior of cast austenitic stainless steels in simulated PWR primary water

    NASA Astrophysics Data System (ADS)

    Li, Shilei; Wang, Yanli; Wang, Hui; Xin, Changsheng; Wang, Xitao

    2016-02-01

    The stress corrosion cracking (SCC) behavior of cast austenitic stainless steels of unaged and thermally aged at 400 °C for as long as 20,000 h were studied by using a slow strain rate testing (SSRT) system. Spinodal decomposition in ferrite during thermal aging leads to hardening in ferrite and embrittlement of the SSRT specimen. Plastic deformation and thermal aging degree have a great influence on the oxidation rate of the studied material in simulated PWR primary water environments. In the SCC regions of the aged SSRT specimen, the surface cracks, formed by the brittle fracture of ferrite phases, are the possible locations for SCC. In the non-SCC regions, brittle fracture of ferrite phases also occurs because of the effect of thermal aging embrittlement.

  17. ATHOS: a computer program for thermal-hydraulic analysis of steam generators. Volume 1. Mathematical and physical models and method of solution. [PWR

    SciTech Connect

    Singhal, A.K.; Keeton, L.W.; Spalding, D.B.; Srikantiah, G.S.

    1982-10-01

    ATHOS (Analysis of the Thermal Hydraulics of Steam Generators) is a computer code developed by CHAM of North America Incorporated, under the contract RP 1066-1 from the Electric Power Research Institute, Palo Alto, California. ATHOS supersedes the earlier code URSULA2. ATHOS is designed for three-dimensional, steady state and transient analyses of PWR steam generators. The current version of the code has been checked out for: three different configurations of the recirculating-type U-tube steam generators; the homogeneous and algebraic-slip flow models; and full and part load operating conditions. The description of ATHOS is divided into four volumes. Volume 1 includes the mathematical and physical models and method of solution.

  18. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  19. International comparison of a depletion calculation benchmark devoted to fuel cycle issues results from the phase 1 dedicated to PWR-UOx fuels

    SciTech Connect

    Roque, B.; Kilger, R.; Laugier, F.; Marimbeau, P.; Riffard, C.; Thro, J. F.; Yudkevich, M.; Hesketh, K.; Sartori, E.

    2006-07-01

    This paper presents the results from the first phase of an international depletion calculations comparison devoted to PWR-UOx fuel cycle issues. This 'benchmark' has been defined within the NEA/OECD Working Party on Scientific Issues in Reactors Systems (WPRS). The aim is to investigate a large range of isotopes, physics quantities and fuel types applied to fuel and back-end cycle configurations. The results analyses have shown that there is a good agreement between participants for the mass calculation of many isotopes. However, it is interesting to observe that better agreement is obtained for isotopes which benefit from experimental validation. In this benchmark, the poorest agreement is obtained in calculating activation products originating from fuel impurities. Some discrepancies on neutron emission rates were also observed, mainly due to the discrepancies on masses calculations. Good agreement was obtained for the total decay heat calculation. (authors)

  20. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

  1. Some Lessons Learned From the SIPACT Simulations on the Design of PWR and Improvement of AM Measures

    SciTech Connect

    Pochard, R.; Jedrzejewski, F.; Nilsuwankosit, S.

    2002-07-01

    the vessel was maintained and the safety margin time was increased. For the scenario that was related to a small break without HPIS, the concept of the safety time margin was still applicable. The time window was observed to be narrower for the bleeding on the secondary side if the core uncover was to be avoided, however. By observing the distribution of the mass in the primary loop, its behavior, which was directly related to the design, was fully demonstrated. One important finding showed that the current PWR design presented some disadvantage under the BDBA condition. Due to the way the water was accumulated in various components, sometime as much as that that was still remained in the pressure vessel, not all the water already presented or injected into the primary loop could reach the pressure vessel to be effectively utilized for core cooling. In order to characterize the availability of the water to cool the core, which related to the NPP BDBA robustness, a simple mass distribution criterion was proposed. Some improvements for the future design were also suggested. (authors)

  2. 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.

  3. Iodine volatility. [PWR; BWR

    SciTech Connect

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

    1984-01-01

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

  4. Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

    SciTech Connect

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

    2008-05-05

    In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.

  5. Preliminary reference Earth model

    NASA Astrophysics Data System (ADS)

    Dziewonski, Adam M.; Anderson, Don L.

    1981-06-01

    A large data set consisting of about 1000 normal mode periods, 500 summary travel time observations, 100 normal mode Q values, mass and moment of inertia have been inverted to obtain the radial distribution of elastic properties, Q values and density in the Earth's interior. The data set was supplemented with a special study of 12 years of ISC phase data which yielded an additional 1.75 × 10 6 travel time observations for P and S waves. In order to obtain satisfactory agreement with the entire data set we were required to take into account anelastic dispersion. The introduction of transverse isotropy into the outer 220 km of the mantle was required in order to satisfy the shorter period fundamental toroidal and spheroidal modes. This anisotropy also improved the fit of the larger data set. The horizontal and vertical velocities in the upper mantle differ by 2-4%, both for P and S waves. The mantle below 220 km is not required to be anisotropic. Mantle Rayleigh waves are surprisingly sensitive to compressional velocity in the upper mantle. High S n velocities, low P n velocities and a pronounced low-velocity zone are features of most global inversion models that are suppressed when anisotropy is allowed for in the inversion. The Preliminary Reference Earth Model, PREM, and auxiliary tables showing fits to the data are presented.

  6. Reference change values.

    PubMed

    Fraser, Callum G

    2011-09-30

    Reference change values (RCV) provide objective tools for assessment of the significance of differences in serial results from an individual. The concept is simple and the calculation easy, since all laboratories know their analytical imprecision (CV(A)) and estimates of within-subject biological variation (CV(I)) are available for a large number of quantities. Generally, CV(I) are constant over time, geography, methodology and in health and chronic stable disease. The formula is RCV=2(1/2) · Z · (CV(A)(2) + CV(I)(2))(1/2), where Z is the number of standard deviations appropriate to the probability. Correct interpretation of the semantics describing the clinical use of RCV is vital for selection of the Z-score. Many quantities of clinically importance exist for which good estimates of RCV are unavailable. Derivation of CV(I) may be difficult for such quantities: flair and imagination are required in selecting populations with chronic but stable disease on whom CV(I) can be determined. RCV can be used for delta-checking and auto-verification and laboratory information management systems (LIMS) can be adapted to do this. Recently, log-normal transformation to obtain unidirectional RCV has been used. Gaps in knowledge of RCV still require filling since the need for measures of change is clearly expressed in guidelines.

  7. Sensor Characteristics Reference Guide

    SciTech Connect

    Cree, Johnathan V.; Dansu, A.; Fuhr, P.; Lanzisera, Steven M.; McIntyre, T.; Muehleisen, Ralph T.; Starke, M.; Banerjee, Pranab; Kuruganti, T.; Castello, C.

    2013-04-01

    The Buildings Technologies Office (BTO), within the U.S. Department of Energy (DOE), Office of Energy Efficiency and Renewable Energy (EERE), is initiating a new program in Sensor and Controls. The vision of this program is: • Buildings operating automatically and continuously at peak energy efficiency over their lifetimes and interoperating effectively with the electric power grid. • Buildings that are self-configuring, self-commissioning, self-learning, self-diagnosing, self-healing, and self-transacting to enable continuous peak performance. • Lower overall building operating costs and higher asset valuation. The overarching goal is to capture 30% energy savings by enhanced management of energy consuming assets and systems through development of cost-effective sensors and controls. One step in achieving this vision is the publication of this Sensor Characteristics Reference Guide. The purpose of the guide is to inform building owners and operators of the current status, capabilities, and limitations of sensor technologies. It is hoped that this guide will aid in the design and procurement process and result in successful implementation of building sensor and control systems. DOE will also use this guide to identify research priorities, develop future specifications for potential market adoption, and provide market clarity through unbiased information

  8. World Reference Center for Arboviruses.

    DTIC Science & Technology

    1977-08-01

    multiple sclerosis . Lyme disease was associated in distribution with Ixodes ticks but the etiologic agent was not isolated. The reference center distributed 566 ampoules of reference sera, viruses, and antigens during 1977; mosquito and vertebrate cell lines were also distributed.

  9. Knowledge Management and Reference Services

    ERIC Educational Resources Information Center

    Gandhi, Smiti

    2004-01-01

    Many corporations are embracing knowledge management (KM) to capture the intellectual capital of their employees. This article focuses on KM applications for reference work in libraries. It defines key concepts of KM, establishes a need for KM for reference services, and reviews various KM initiatives for reference services.

  10. Web Reference: A Virtual Reality.

    ERIC Educational Resources Information Center

    Foster, Janet

    1999-01-01

    Presents ideas and strategies to enhance digital reference services available via the Internet in public libraries. Describes print publications which include Web reference columns; subject guides, both print and online; and the resources of the Internet Public Library and other virtual reference desks. (LRW)

  11. Paraprofessionals at the Reference Desk.

    ERIC Educational Resources Information Center

    Murfin, Marjorie E.; Bunge, Charles A.

    1988-01-01

    Describes a study that compared the quality of reference services provided by reference librarians and paraprofessionals, using patron satisfaction as the measure of success. Factors associated with satisfactory service are identified, and suggestions for the effective use of paraprofessional staff are presented. (4 references) (CLB)

  12. Fundamentals of Managing Reference Collections

    ERIC Educational Resources Information Center

    Singer, Carol A.

    2012-01-01

    Whether a library's reference collection is large or small, it needs constant attention. Singer's book offers information and insight on best practices for reference collection management, no matter the size, and shows why managing without a plan is a recipe for clutter and confusion. In this very practical guide, reference librarians will learn:…

  13. Canadian listeriosis reference service.

    PubMed

    Pagotto, Franco; Ng, Lai-King; Clark, Clifford; Farber, Jeff

    2006-01-01

    Listeria monocytogenes, a psychrotrophic organism capable of growing at refrigeration temperatures, is of major concern in extended shelf life, refrigerated foods. Considering that as much as 80-90% of human listeriosis cases are linked to the ingestion of contaminated food, human cases are predominantly seen in high-risk individuals, including organ-transplant recipients, patients with AIDS and HIV-infected individuals, pregnant women, cancer patients, and the elderly. In 2001, the Canadian Listeriosis Reference Service (LRS) was created by the Bureau of Microbial Hazards (Health Canada) and the National Microbiology Laboratory (now part of the Public Health Agency of Canada). Major goals of the LRS include investigation of listeriosis cases and maintenance of a national collection of isolates. The LRS intends to create a comprehensive molecular epidemiological database of all isolates in Canada for use as a resource for outbreak investigations, research and other microbiological investigations. The PFGE profiles are being established and stored for clinical, food, environmental, and possibly animal strains of L. monocytogenes. The LRS pursues research activities for investigation and implementation of other molecular methods for characterizing L. monocytogenes isolates. Ribotyping, Multi-locus Sequence Typing (MLST), Variable Number of Tandem Repeats (VNTR), Multi-locus virulence sequence typing (MLVA), microarray- based technologies and sequence-based typing schemes, are being investigated on selected diversity sets. The LRS has also used PFGE typing for outbreak investigations. The molecular epidemiological data, timely coordination and exchange of information should help to reduce the incidence of listeriosis in Canada. In Canada, listeriosis is not a national notifiable disease, except for the province of Quebec, where it has been since 1999. The LRS, Canadian Public Health Laboratory Network, and federal epidemiologists are currently working on making human

  14. Capillary reference half-cell

    DOEpatents

    Hall, Stephen H.

    1996-01-01

    The present invention is a reference half-cell electrode wherein intermingling of test fluid with reference fluid does not affect the performance of the reference half-cell over a long time. This intermingling reference half-cell may be used as a single or double junction submersible or surface reference electrode. The intermingling reference half-cell relies on a capillary tube having a first end open to reference fluid and a second end open to test fluid wherein the small diameter of the capillary tube limits free motion of fluid within the capillary to diffusion. The electrode is placed near the first end of the capillary in contact with the reference fluid. The method of operation of the present invention begins with filling the capillary tube with a reference solution. After closing the first end of the capillary, the capillary tube may be fully submerged or partially submerged with the second open end inserted into test fluid. Since the electrode is placed near the first end of the capillary, and since the test fluid may intermingle with the reference fluid through the second open end only by diffusion, this intermingling capillary reference half-cell provides a stable voltage potential for long time periods.

  15. Capillary reference half-cell

    DOEpatents

    Hall, S.H.

    1996-02-13

    The present invention is a reference half-cell electrode wherein intermingling of test fluid with reference fluid does not affect the performance of the reference half-cell over a long time. This intermingling reference half-cell may be used as a single or double junction submersible or surface reference electrode. The intermingling reference half-cell relies on a capillary tube having a first end open to reference fluid and a second end open to test fluid wherein the small diameter of the capillary tube limits free motion of fluid within the capillary to diffusion. The electrode is placed near the first end of the capillary in contact with the reference fluid. The method of operation of the present invention begins with filling the capillary tube with a reference solution. After closing the first end of the capillary, the capillary tube may be fully submerged or partially submerged with the second open end inserted into test fluid. Since the electrode is placed near the first end of the capillary, and since the test fluid may intermingle with the reference fluid through the second open end only by diffusion, this intermingling capillary reference half-cell provides a stable voltage potential for long time periods. 11 figs.

  16. Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

    NASA Astrophysics Data System (ADS)

    Salko, Robert K.

    COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for

  17. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    SciTech Connect

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurin; Natesan, Ken

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  18. Optical probe with reference fiber

    DOEpatents

    Da Silva, Luiz B.; Chase, Charles L.

    2006-03-14

    A system for characterizing tissue includes the steps of generating an emission signal, generating a reference signal, directing the emission signal to and from the tissue, directing the reference signal in a predetermined manner relative to the emission signal, and using the reference signal to compensate the emission signal. In one embodiment compensation is provided for fluctuations in light delivery to the tip of the probe due to cable motion.

  19. User Preferences in Reference Services: Virtual Reference and Academic Libraries

    ERIC Educational Resources Information Center

    Cummings, Joel; Cummings, Lara; Frederiksen, Linda

    2007-01-01

    This study examines the use of chat in an academic library's user population and where virtual reference services might fit within the spectrum of public services offered by academic libraries. Using questionnaires, this research demonstrates that many within the academic community are open to the idea of chat-based reference or using chat for…

  20. Reach for Reference: Elementary-Middle School Science Reference Collections

    ERIC Educational Resources Information Center

    Safford, Barbara Ripp

    2005-01-01

    This article presents a brief review of some new school science reference works. Two of the sources are traditional, while one is considered experimental. The two traditional reference works reviewed are "The American Heritage Children's Science Dictionary" for upper elementary grades, and "The American Heritage Student Science Dictionary" for…

  1. Reference Readiness for AV Questions.

    ERIC Educational Resources Information Center

    Drolet, Leon L., Jr.

    1981-01-01

    Reviews 50 reference tools which librarians can use to answer almost any audiovisual question including queries on trivia, equipment selection, biographical information, and motion picture ratings. (LLS)

  2. The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment

    SciTech Connect

    Kammenzind, B.F., Eklund, K.L. and Bajaj, R.

    2001-06-21

    Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pellet contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress

  3. Potential of pin-by-pin SPN calculations as an industrial reference

    SciTech Connect

    Fliscounakis, M.; Girardi, E.; Courau, T.; Couyras, D.

    2012-07-01

    This paper aims at analysing the potential of pin-by-pin SP{sub n} calculations to compute the neutronic flux in PWR cores as an alternative to the diffusion approximation. As far as pin-by-pin calculations are concerned, a SPH equivalence is used to preserve the reactions rates. The use of SPH equivalence is a common practice in core diffusion calculations. In this paper, a methodology to generalize the equivalence procedure in the SP{sub n} equations context is presented. In order to verify and validate the equivalence procedure, SP{sub n} calculations are compared to 2D transport reference results obtained with the APOLL02 code. The validation cases consist in 3x3 analytical assembly color sets involving burn-up heterogeneities, UOX/MOX interfaces, and control rods. Considering various energy discretizations (up to 26 groups) and flux development orders (up to 7) for the SP{sub n} equations, results show that 26-group SP{sub 3} calculations are very close to the transport reference (with pin production rates discrepancies < 1%). This proves the high interest of pin-by-pin SP{sub n} calculations as an industrial reference when relying on 26 energy groups combined with SP{sub 3} flux development order. Additionally, the SP{sub n} results are compared to diffusion pin-by-pin calculations, in order to evaluate the potential benefit of using a SP{sub n} solver as an alternative to diffusion. Discrepancies on pin-production rates are less than 1.6% for 6-group SP{sub 3} calculations against 3.2% for 2-group diffusion calculations. This shows that SP{sub n} solvers may be considered as an alternative to multigroup diffusion. (authors)

  4. Robot at the Reference Desk?

    ERIC Educational Resources Information Center

    Smith, Karen F.

    1986-01-01

    Describes how a librarian, assisted by a knowledge engineer, developed a computerized reference assistance system for a separate government documents department. Rationale for the expert system, problems in selecting reference materials and user questions to computerize, and the formulation of a workable human/computer interface are covered. A…

  5. Dedicated online virtual reference instruction.

    PubMed

    Guillot, Ladonna; Stahr, Beth; Plaisance, Louise

    2005-01-01

    To facilitate nursing students' information literacy skills and enhance traditional library user services, academic librarians have developed synchronous (real-time) online virtual reference instruction in nursing research classes. The authors discuss collaborative efforts of nursing and library faculty in planning, implementing, and evaluating a discipline-specific virtual reference pilot program.

  6. Moving Reference to the Web.

    ERIC Educational Resources Information Center

    McGlamery, Susan; Coffman, Steve

    2000-01-01

    Explores the possibility of using Web contact center software to offer reference assistance to remote users. Discusses a project by the Metropolitan Cooperative Library System/Santiago Library System consortium to test contact center software and to develop a virtual reference network. (Author/LRW)

  7. The Virtual Reference Librarian's Handbook.

    ERIC Educational Resources Information Center

    Lipow, Anne Grodzins

    This book is a practical guide to librarians and their administrators who are thinking about or in the early stages of providing virtual reference service. Part 1, "The Decision to Go Virtual," provides a context for thinking about virtual reference, including the benefits and problems, getting in the virtual frame of mind, and shopping…

  8. Ethics and the Reference Librarian.

    ERIC Educational Resources Information Center

    Bunge, Charles A.

    1999-01-01

    Discussion of ethical reference practice focuses on guidelines for the individual reference librarian's interactions with clients. Topics include the professional-client relationship; competence; diligence; confidentiality; independence of judgment; honesty and candor; and obligations to third parties, including the democratic society at large.…

  9. Reference for radiographic film interpreters

    NASA Technical Reports Server (NTRS)

    Austin, D. L.

    1970-01-01

    Reference of X-ray film images provides examples of weld defects, film quality, stainless steel welded tubing, and acceptable weld conditions. A summary sheet details the discrepancies shown on the film strip. This reference aids in interpreting and evaluating radiographic film of weldments.

  10. Expert Systems for Reference Work.

    ERIC Educational Resources Information Center

    Parrot, James R.

    1986-01-01

    Discussion of library reference work that may be suitable for use of expert systems focuses on (1) information and literature searches, and (2) requests to interpret bibliographic references and locate items listed. Systems and computer-assisted instruction modules designed for information retrieval at the University of Waterloo Library are…

  11. Technostress and the Reference Librarian.

    ERIC Educational Resources Information Center

    Kupersmith, John

    1992-01-01

    Defines "technostress" as the stress experienced by reference librarians who must constantly deal with the demands of new information technology and the changes they produce in the work place. Discussion includes suggested ways in which both organizations and individuals can work to reduce stress. (27 references) (LAE)

  12. Changing Roles for References Librarians.

    ERIC Educational Resources Information Center

    Kelly, Julia; Robbins, Kathryn

    1996-01-01

    Discusses the future outlook for reference librarians, with topics including: "Technology as the Source of Change"; "Impact of the Internet"; "Defining the Virtual Library"; "Rethinking Reference"; "Out of the Library and into the Streets"; "Asking Users About Their Needs"; "Standardization and Artificial Intelligence"; "The Financial Future"; and…

  13. Queuing Theory and Reference Transactions.

    ERIC Educational Resources Information Center

    Terbille, Charles

    1995-01-01

    Examines the implications of applying the queuing theory to three different reference situations: (1) random patron arrivals; (2) random durations of transactions; and (3) use of two librarians. Tables and figures represent results from spreadsheet calculations of queues for each reference situation. (JMV)

  14. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  15. Generalizing indexical-functional reference

    SciTech Connect

    Schoppers, M.; Shu, R.

    1996-12-31

    The goals of situated agents generally do not specify particular objects: they require only that some suitable object should be chosen and manipulated (e.g. any red block). Situated agents engaged in deictic reference grounding, however, may well track a chosen referent object with such fixity of purpose that an unchosen object may be regarded as an obstacle even though it satisfies the agent`s goals. In earlier work this problem was bridged by hand-coding. This paper lifts the problem to the symbol level, endowing agents with perceptual referent selection actions and performing those actions as required to allow or disallow opportunistic re-selection of referents. Our work preserves the ability of situated agents to find and track specific objects, adds an ability to automatically exploit the opportunities allowed by nonspecific references, and provides a starting point for studying how much opportunistic perception is appropriate.

  16. Improved ultrasonic standard reference blocks

    NASA Technical Reports Server (NTRS)

    Eitzen, D. G.; Sushinsky, G. F.; Chwirut, D. J.; Bechtoldt, C. J.; Ruff, A. W.

    1976-01-01

    A program to improve the quality, reproducibility and reliability of nondestructive testing through the development of improved ASTM-type ultrasonic reference standards is described. Reference blocks of aluminum, steel, and titanium alloys are to be considered. Equipment representing the state-of-the-art in laboratory and field ultrasonic equipment was obtained and evaluated. RF and spectral data on ten sets of ultrasonic reference blocks have been taken as part of a task to quantify the variability in response from nominally identical blocks. Techniques for residual stress, preferred orientation, and micro-structural measurements were refined and are applied to a reference block rejected by the manufacturer during fabrication in order to evaluate the effect of metallurgical condition on block response. New fabrication techniques for reference blocks are discussed and ASTM activities are summarized.

  17. Review of the Reference Dose and Reference Concentration Processes Document

    EPA Pesticide Factsheets

    Summarizes the review and deliberations of the Risk Assessment Forum’s RfD/RfC Technical Panel and its recommendations for improvements in oral referencedose/inhalation reference concentration (RfD/RfC) process.

  18. Reference and Standard Atmosphere Models

    NASA Technical Reports Server (NTRS)

    Johnson, Dale L.; Roberts, Barry C.; Vaughan, William W.; Parker, Nelson C. (Technical Monitor)

    2002-01-01

    This paper describes the development of standard and reference atmosphere models along with the history of their origin and use since the mid 19th century. The first "Standard Atmospheres" were established by international agreement in the 1920's. Later some countries, notably the United States, also developed and published "Standard Atmospheres". The term "Reference Atmospheres" is used to identify atmosphere models for specific geographical locations. Range Reference Atmosphere Models developed first during the 1960's are examples of these descriptions of the atmosphere. This paper discusses the various models, scopes, applications and limitations relative to use in aerospace industry activities.

  19. Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics

    NASA Astrophysics Data System (ADS)

    Nuttin, A.; Capellan, N.; David, S.; Doligez, X.; El Mhari, C.; Méplan, O.

    2014-06-01

    Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.

  20. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5. Probabilistic fracture mechanics analysis. Load Combination Program Project I final report

    SciTech Connect

    Harris, D.O.; Lim, E.Y.; Dedhia, D.D.

    1981-06-01

    The primary purpose of the Load Combination Program covered in this report is to estimate the probability of a seismic induced LOCA in the primary piping of a commercial pressurized water reactor (PWR). Best estimates, rather than upper bound results are desired. This was accomplished by use of a fracture mechanics model that employs a random distribution of initial cracks in the piping welds. Estimates of the probability of cracks of various sizes initially existing in the welds are combined with fracture mechanics calculations of how these cracks would grow during service. This then leads to direct estimates of the probability of failure as a function of time and location within the piping system. The influence of varying the stress history to which the piping is subjected is easily determined. Seismic events enter into the analysis through the stresses they impose on the pipes. Hence, the influence of various seismic events on the piping failure probability can be determined, thereby providing the desired information.

  1. Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

    NASA Astrophysics Data System (ADS)

    Riglet-Martial, Ch.; Sercombe, J.; Lamontagne, J.; Noirot, J.; Roure, I.; Blay, T.; Desgranges, L.

    2016-11-01

    The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that irradiated fuels (of nominal stoichiometry close to 2.000) under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U4O9 type in its cold section, of lower temperature. The parameters governing the oxidation states of UO2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

  2. A Study on Structured Simulation Framework for Design and Evaluation of Human-Machine Interface System -Application for On-line Risk Monitoring for PWR Nuclear Power Plant-

    SciTech Connect

    Zhan, J.; Yang, M.; Li, S.C.; Peng, M.J.; Yan, S.Y.; Zhang, Z.J.

    2006-07-01

    The operators in the main control room of Nuclear Power Plant (NPP) need to monitor plant condition through operation panels and understand the system problems by their experiences and skills. It is a very hard work because even a single fault will cause a large number of plant parameters abnormal and operators are required to perform trouble-shooting actions in a short time interval. It will bring potential risks if operators misunderstand the system problems or make a commission error to manipulate an irrelevant switch with their current operation. This study aims at developing an on-line risk monitoring technique based on Multilevel Flow Models (MFM) for monitoring and predicting potential risks in current plant condition by calculating plant reliability. The proposed technique can be also used for navigating operators by estimating the influence of their operations on plant condition before they take an action that will be necessary in plant operation, and therefore, can reduce human errors. This paper describes the risk monitoring technique and illustrates its application by a Steam Generator Tube Rupture (SGTR) accident in a 2-loop Pressurized Water Reactor (PWR) Marine Nuclear Power Plant (MNPP). (authors)

  3. 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.

  4. On the Application of CFD Modeling for the Prediction of the Degree of Mixing in a PWR During a Boron Dilution Transient

    SciTech Connect

    Lycklama, Jan-Aiso; Hoehne, Thomas

    2006-07-01

    In a Pressurized Water Reactor, negative reactivity is present in the core by means of Boric acid as a soluble neutron absorber in the coolant water. During a so-called Boron Dilution Transient (BDT), a de-borated slug of coolant water is transported from the cold leg into the reactor vessel, and the borated coolant water is diluted by mixing with this un-borated water. The resulting decrease in the boron concentration leads to an insertion of positive reactivity in the core, which may lead to a reactivity excursion. The associated power peak may damage the fuel rods. The mixing of borated and un-borated water in downcomer and lower plenum is an important process, because it mitigates the degree of reactivity insertion. In the present study the application of Computational Fluid Dynamics (CFD) for the prediction of this mixing of un-borated with borated water in the RPV has been assessed. The analyses have been compared with the measurement data from the Rossendorf coolant mixing model (ROCOM) experiment. The ROCOM test facility represents the primary cooling system of a KONVOI type of PWR (1300 MW{sub el}). In spite of the complicated spatial, temporal, and geometrical aspects of the flow in the RPV, the agreement between the calculated and the experimental data is good. The CFD model tends to slightly under predict the degree of mixing in the RPV resulting in a slight under-prediction of the boron concentration at the core. (authors)

  5. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    SciTech Connect

    Lafleur, Adrienne M; Charlton, William S; Menlove, Howard O; Swinhoe, Martyn T

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  6. Genetics Home Reference: isolated hyperchlorhidrosis

    MedlinePlus

    ... loss of salt (sodium chloride or NaCl) in sweat. In particular, "hyperchlorhidrosis" refers to the high levels of chloride found in sweat, although both sodium and chloride are released. Because ...

  7. Sports Reference: A Core Collection.

    ERIC Educational Resources Information Center

    Safford, Barbara Ripp

    2003-01-01

    Discusses reasons for including sports books in school library reference collections, explains why they should not be found only in public library collections, and provides six annotated bibliographies of sports books suitable for intermediate or middle school library collections. (LRW)

  8. Computerizing the Reference Desk Schedule.

    ERIC Educational Resources Information Center

    deHaas, Pat

    1983-01-01

    Discussion of the scheduling procedures of librarians' hours at the reference desk at the Rutherford Humanities and Social Sciences Library, University of Alberta, highlights services provided, the preference table system, and manual scheduling versus computer scheduling. (EJS)

  9. Genetics Home Reference: Laron syndrome

    MedlinePlus

    ... AL. Obesity, diabetes and cancer: insight into the relationship from a cohort with growth hormone receptor deficiency. ... healthcare professional . About Genetics Home Reference Site Map Customer Support Selection Criteria for Links USA.gov Copyright ...

  10. Genetics Home Reference: Miyoshi myopathy

    MedlinePlus

    ... Itoyama Y. Dysferlin mutations in Japanese Miyoshi myopathy: relationship to phenotype. Neurology. 2003 Jun 10;60(11): ... healthcare professional . About Genetics Home Reference Site Map Customer Support Selection Criteria for Links USA.gov Copyright ...

  11. Genetics Home Reference: Asperger syndrome

    MedlinePlus

    ... a combination of genetic variations and environmental factors influence the development of this complex condition. Asperger syndrome ... healthcare professional . About Genetics Home Reference Site Map Customer Support Selection Criteria for Links USA.gov Copyright ...

  12. Selected Reference Books of 1999.

    ERIC Educational Resources Information Center

    McIlvaine, Eileen

    2000-01-01

    Presents annotated bibliographies of a selection of recent scholarly and general reference works under the subject headings of publishing, periodical indexes, philosophy and religion, literature, music, art, photography, social sciences, business, history, and new editions. (LRW)

  13. International reference standards in coagulation.

    PubMed

    Raut, Sanj; Hubbard, Anthony R

    2010-07-01

    Measurement of coagulation factor activity using absolute physico-chemical techniques is not possible and estimation therefore relies on comparative bioassay relative to a reference standard with a known or assigned potency. However the inherent variability of locally prepared and calibrated reference standards can give rise to poor agreement between laboratories and methods. Harmonisation of measurement between laboratories at the international level relies on the availability of a common source of calibration for local reference standards and this is provided by the World Health Organization (WHO) International Standards which define the International Unit for the analyte. This article describes the principles, practices and problems of biological standardisation and the development and use of reference standards for assays of coagulation factors, with particular emphasis on WHO International Standards for both concentrates and plasma.

  14. Genetics Home Reference: WAGR syndrome

    MedlinePlus

    ... signs and symptoms of WAGR syndrome can include childhood-onset obesity, inflammation of the pancreas (pancreatitis), and kidney failure. When WAGR syndrome includes childhood-onset obesity, it is often referred to as WAGRO syndrome. ...

  15. Improved ultrasonic standard reference blocks

    NASA Technical Reports Server (NTRS)

    Eitzen, D. G.

    1975-01-01

    A program to improve the quality, reproducibility and reliability of nondestructive testing through the development of improved ASTM-type ultrasonic reference standards is described. Reference blocks of aluminum, steel, and titanium alloys were considered. Equipment representing the state-of-the-art in laboratory and field ultrasonic equipment was obtained and evaluated. Some RF and spectral data on ten sets of ultrasonic reference blocks were taken as part of a task to quantify the variability in response from nominally identical blocks. Techniques for residual stress, preferred orientation, and microstructural measurements were refined and are applied to a reference block rejected by the manufacturer during fabrication in order to evaluate the effect of metallurgical condition on block response.

  16. Virtual reference: chat with us!

    PubMed

    Lapidus, Mariana; Bond, Irena

    2009-01-01

    Virtual chat services represent an exciting way to provide patrons of medical libraries with instant reference help in an academic environment. The purpose of this article is to examine the implementation, marketing process, use, and development of a virtual reference service initiated at the Massachusetts College of Pharmacy and Health Sciences and its three-campus libraries. In addition, this paper will discuss practical recommendations for the future improvement of the service.

  17. Space Station reference configuration description

    NASA Technical Reports Server (NTRS)

    1984-01-01

    The data generated by the Space Station Program Skunk Works over a period of 4 months which supports the definition of a Space Station reference configuration is documented. The data were generated to meet these objectives: (1) provide a focal point for the definition and assessment of program requirements; (2) establish a basis for estimating program cost; and (3) define a reference configuration in sufficient detail to allow its inclusion in the definition phase Request for Proposal (RFP).

  18. Virtual Reference Interferometry: Theory & Experiment

    NASA Astrophysics Data System (ADS)

    Galle, Michael Anthony

    This thesis introduces the idea that a simulated interferogram can be used as a reference for an interferometer. This new concept represents a paradigm shift from the conventional thinking, where a reference is the phase of a wavefront that traverses a known path. The simulated interferogram used as a reference is called a virtual reference. This thesis develops the theory of virtual reference interferometry and uses it for the characterization of chromatic dispersion in short length (<1m) fibers and optical components. Characterization of chromatic dispersion on short length fiber and optical components is a very difficult challenge. Accurate measurement of first and second order dispersion is important for applications from optical component design to nonlinear photonics, sensing and communications. Techniques for short-length dispersion characterization are therefore critical to the development of many photonic systems. The current generation of short-length dispersion measurement techniques are either easy to operate but lack sufficient accuracy, or have sufficient accuracy but are difficult to operate. The use of a virtual reference combines the advantages of these techniques so that it is both accurate and easy to operate. Chromatic dispersion measurements based on virtual reference interferometry have similar accuracy as the best conventional measurement techniques due to the ability to measure first and second order dispersion directly from the interference pattern. Unique capabilities of virtual reference interferometry are demonstrated, followed by a derivation of the operational constraints and system parameters. The technique is also applied to the characterization of few-mode fibers, a hot topic in telecommunications research where mode division multiplexing promises to expand network bandwidth. Also introduced is the theory of dispersive virtual reference interferometry, which can be used to overcome the bandwidth limitations associated with the

  19. Using Virtual Reference Transcripts for Staff Training.

    ERIC Educational Resources Information Center

    Ward, David

    2003-01-01

    Describes a method of library staff training based on chat transcript analysis in which graduate student workers at a university reference desk examined transcripts of actual virtual reference desk transactions to analyze reference interviews. Discusses reference interview standards, reference desk behavior, and reference interview skills in…

  20. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

  1. Sequence Factorization with Multiple References

    PubMed Central

    Wandelt, Sebastian; Leser, Ulf

    2015-01-01

    The success of high-throughput sequencing has lead to an increasing number of projects which sequence large populations of a species. Storage and analysis of sequence data is a key challenge in these projects, because of the sheer size of the datasets. Compression is one simple technology to deal with this challenge. Referential factorization and compression schemes, which store only the differences between input sequence and a reference sequence, gained lots of interest in this field. Highly-similar sequences, e.g., Human genomes, can be compressed with a compression ratio of 1,000:1 and more, up to two orders of magnitude better than with standard compression techniques. Recently, it was shown that the compression against multiple references from the same species can boost the compression ratio up to 4,000:1. However, a detailed analysis of using multiple references is lacking, e.g., for main memory consumption and optimality. In this paper, we describe one key technique for the referential compression against multiple references: The factorization of sequences. Based on the notion of an optimal factorization, we propose optimization heuristics and identify parameter settings which greatly influence 1) the size of the factorization, 2) the time for factorization, and 3) the required amount of main memory. We evaluate a total of 30 setups with a varying number of references on data from three different species. Our results show a wide range of factorization sizes (optimal to an overhead of up to 300%), factorization speed (0.01 MB/s to more than 600 MB/s), and main memory usage (few dozen MB to dozens of GB). Based on our evaluation, we identify the best configurations for common use cases. Our evaluation shows that multi-reference factorization is much better than single-reference factorization. PMID:26422374

  2. Virtual Reference, Real Money: Modeling Costs in Virtual Reference Services

    ERIC Educational Resources Information Center

    Eakin, Lori; Pomerantz, Jeffrey

    2009-01-01

    Libraries nationwide are in yet another phase of belt tightening. Without an understanding of the economic factors that influence library operations, however, controlling costs and performing cost-benefit analyses on services is difficult. This paper describes a project to develop a cost model for collaborative virtual reference services. This…

  3. Reference Anytime Anywhere: Towards Virtual Reference Services at Penn State.

    ERIC Educational Resources Information Center

    Moyo, Lesley M.

    2002-01-01

    Outlines the service rationale, software and technology considerations taken by the Pennsylvania State University library in planning towards online, real-time reference services and provides an overview of the planned pilot project. Discusses recent trends in academic electronic libraries, including providing value-added services to support…

  4. Reference Anywhere, Anytime: Recent Works on Virtual Reference

    ERIC Educational Resources Information Center

    Lindsay, Elizabeth Blakesley

    2008-01-01

    The topic of virtual reference has been a popular one in the past year, and this column highlights a range of those articles. Included are overviews of the topic, articles addressing specific technological solutions and concerns, and pieces that explore the intricacies of delivering and assessing a "traditional" service through a…

  5. The Virtual Reference Librarian: Using Desktop Videoconferencing for Distance Reference.

    ERIC Educational Resources Information Center

    Pagell, Ruth A.

    1996-01-01

    The Center for Business Information and the Goizueta Business School at Emory University (Georgia) tested desktop videoconferencing as a means to deliver reference services, including consultation, documentation, training, and sharing of CD-ROM databases. Provides a brief overview of the technology, describes project beta testing, and discusses…

  6. Reference and Reference Failures. Technical Report No. 398.

    ERIC Educational Resources Information Center

    Goodman, Bradley A.

    In order to build robust natural language processing systems that can detect and recover from miscommunication, the investigation of how people communicate and how they recover from problems in communication described in this artificial intelligence report focused on reference problems which a listener may have in determining what or whom a…

  7. Reference-based phasing using the Haplotype Reference Consortium panel.

    PubMed

    Loh, Po-Ru; Danecek, Petr; Palamara, Pier Francesco; Fuchsberger, Christian; A Reshef, Yakir; K Finucane, Hilary; Schoenherr, Sebastian; Forer, Lukas; McCarthy, Shane; Abecasis, Goncalo R; Durbin, Richard; L Price, Alkes

    2016-11-01

    Haplotype phasing is a fundamental problem in medical and population genetics. Phasing is generally performed via statistical phasing in a genotyped cohort, an approach that can yield high accuracy in very large cohorts but attains lower accuracy in smaller cohorts. Here we instead explore the paradigm of reference-based phasing. We introduce a new phasing algorithm, Eagle2, that attains high accuracy across a broad range of cohort sizes by efficiently leveraging information from large external reference panels (such as the Haplotype Reference Consortium; HRC) using a new data structure based on the positional Burrows-Wheeler transform. We demonstrate that Eagle2 attains a ∼20× speedup and ∼10% increase in accuracy compared to reference-based phasing using SHAPEIT2. On European-ancestry samples, Eagle2 with the HRC panel achieves >2× the accuracy of 1000 Genomes-based phasing. Eagle2 is open source and freely available for HRC-based phasing via the Sanger Imputation Service and the Michigan Imputation Server.

  8. PIRLA DBMS quick reference guide

    SciTech Connect

    Smith, J.

    1991-11-01

    The handbook facilitates the use of the PIRLA (Paleoecological Investigation of Recent Lake Acidification) database retrievals by showing all possible inputs, outputs, ranges, and quick reference information that you would want at your fingertips when accessing the PIRLA data. The handbook assumes no prior knowledge of PIRLA or SIR (a database management system), although a basic familiarity with computers is helpful. The PIRLA Data Base Management System User's Manual is recommended for reference, much additional detail, description of the intrinsic structure of the PIRLA database, and how it is set up under SIR.

  9. 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.

  10. Tractor Transmissions. A Teaching Reference.

    ERIC Educational Resources Information Center

    American Association for Agricultural Engineering and Vocational Agriculture, Athens, GA.

    The manual was developed as a reference for teaching students about transmissions in farm tractors. The manual is divided into five sections: (1) transmission history, (2) gears and bearings in transmission, (3) sliding-gear transmissions, (4) planetary gearing, and (5) glossary. The working principles of the sliding-gear transmission, the most…

  11. When Is Cataphoric Reference Recognised?

    ERIC Educational Resources Information Center

    Filik, Ruth; Sanford, Anthony J.

    2008-01-01

    Pronouns typically have explicit antecedents in the prior discourse otherwise processing difficulty is experienced. However, it has been argued [Gordon, P. C., & Hendrick, R. (1997). "Intuitive knowledge of linguistic co-reference." "Cognition, 62", 325-370; Gordon, P. C., & Hendrick, R. (1998). "The representation and processing of co-reference…

  12. The Rosia Language Reference Manual

    DTIC Science & Technology

    1981-12-01

    Alternate database 16 Anaphora 26, 30, 56 Anaphoric reference 30 Arithmetic expressions 33 Arithmetic operators 8 Bind 54 Binding 54, 57 BNF 94...Descriptions as Class Membership Tests .......... 28 4.2.7. Descriptions and Modification ................... 29 4.2.8. Anaphoric Descriptions...54 4.9.3. Variable Binding ................................ 54 4.9.4. Pattern Matching ................................ 55 5

  13. Newfoundland. Reference Series No. 34.

    ERIC Educational Resources Information Center

    Department of External Affairs, Ottawa (Ontario).

    This booklet, one of a series featuring the Canadian provinces, presents a brief overview of Newfoundland and is suitable for teacher reference or student reading. Separate sections discuss geography and climate, history, economy, population and settlement, arts and culture, leisure and recreation, and heritage. Specific topics include the…

  14. Selected Reference Books of 1998.

    ERIC Educational Resources Information Center

    McIlvaine, Eileen

    1999-01-01

    Reviews a selection of recent scholarly and general reference works under the categories of Periodicals and Newspapers, Philosophy, Literature, Film and Radio, Art and Architecture, Music, Political Science, Women's Studies, and History. A brief summary of new editions of standard works is provided at the end of the articles. (AEF)

  15. Quebec. Reference Series No. 30.

    ERIC Educational Resources Information Center

    Department of External Affairs, Ottawa (Ontario).

    This booklet, one of a series featuring the Canadian provinces, presents a brief overview of Quebec and is suitable for teacher reference or student reading. Separate sections discuss geography, climate, population, history, political history, recent politics, agriculture, forestry, mining, manufacturing and industry, hydroelectric power,…

  16. [Developmental Placement.] Collected Research References.

    ERIC Educational Resources Information Center

    Bjorklund, Gail

    Drawing on information and references in the ERIC system, this literature review describes research related to a child's developmental placement. The issues examined include school entrance age; predictive validity, reliability, and features of Gesell School Readiness Assessment; retention; and the effectiveness of developmental placement. A…

  17. Manitoba. Reference Series No. 28.

    ERIC Educational Resources Information Center

    Department of External Affairs, Ottawa (Ontario).

    This booklet, one of a series featuring the Canadian provinces, presents a brief overview of Manitoba and is suitable for teacher reference or student reading. Separate sections discuss agriculture, mining, energy, transportation and communication, fishing, forestry, fur trapping, health and social services, education, and political life. Specific…

  18. Ontario. Reference Series No. 29.

    ERIC Educational Resources Information Center

    Department of External Affairs, Ottawa (Ontario).

    This booklet, one of a series featuring the Canadian provinces, presents a brief overview of Ontario and is suitable for teacher reference or student reading. Separate sections discuss geography, climate, history, agriculture, forestry, fishing, mining, manufacturing, transportation, energy, arts and culture, sports and recreation, and people and…

  19. Chapter 11: Dietary reference intakes

    Technology Transfer Automated Retrieval System (TEKTRAN)

    The Dietary Reference Intakes (DRI) are a set of recommendations intended to provide guidance in evaluating nutrient intakes and planning meals on the basis of nutrient adequacy. In contrast to their predecessor, Recommended Dietary Allowances last published in 1989, the DRIs differ in two ways: th...

  20. A REFERENCE GRAMMAR OF BENGALI.

    ERIC Educational Resources Information Center

    RAY, PUNYA SLOKA; AND OTHERS

    A REFERENCE GRAMMAR WAS PRODUCED FOR THE BENGALI LANGUAGE. THE WORK CONTAINS CHAPTERS ON--(1) SOCIAL AND HISTORICAL BACKGROUND, (2) HISTORY OF THE LANGUAGE, (3) SOURCES OF LEXICAL ITEMS, (4) ORTHOGRAPHY, (5) PHONOLOGY, (6) NOUN INFLECTIONS, (7) VERBS, (8) POSTPOSITIONS, (9) ENCLITICS, (10) NUMERALS, (11) NEGATION, (12) FORMATIVE AFFIXES IN…

  1. Saskatchewan. Reference Series No. 27.

    ERIC Educational Resources Information Center

    Department of External Affairs, Ottawa (Ontario).

    This booklet, one of a series featuring the Canadian provinces, presents a brief overview of Saskatchewan and is suitable for teacher reference or student reading. Separate sections discuss history, economy, oil, uranium, potash, coal, minerals and metals, agriculture, forestry, tourism and recreation, arts and culture, and people. Specific topics…

  2. Reference Works in Reduced Size

    ERIC Educational Resources Information Center

    Hennessey, David

    1977-01-01

    Lower cost and less shelf space requirements have made major reference works in miniaturized print editions attractive to small libraries. The major disadvantage is the necessity of a magnifying glass for reading entries. A bibliography of reduced-size editions is included. (JAB)

  3. A Reference Grammar of Kashmiri.

    ERIC Educational Resources Information Center

    Kachru, Braj B.

    This study was developed for two pedagogical purposes--first, to provide a skeleton grammar of the Kashmiri language which could be used by teachers of Kashmiri to develop teaching materials for both Indian and non-Indian learners of Kashmiri; and second, to provide an introductory reference manual of Kashmiri for students of the language. The…

  4. Childhood Obesity. Special Reference Briefs.

    ERIC Educational Resources Information Center

    Winick, Myron

    This reference brief deals with the problem of childhood obesity and how it can lead to obesity in the adult. Eighty-four abstracts are presented of studies on the identification, prevention, and treatment of obesity in children, focusing on diet and psychological attitudes. Subjects of the studies were children ranging in age from infancy through…

  5. Guam and Micronesia Reference Sources.

    ERIC Educational Resources Information Center

    Goetzfridt, Nicholas J.; Goniwiecha, Mark C.

    1993-01-01

    This article lists reference sources for studying Guam and Micronesia. The entries are arranged alphabetically by main entry within each section in the categories of: (1) bibliographical works; (2) travel and guide books; (3) handbooks and surveys; (4) dictionaries; (5) yearbooks; (6) periodical and newspaper publications; and (7) audiovisual…

  6. Human Rights: The Essential Reference.

    ERIC Educational Resources Information Center

    Devine, Carol; Hansen, Carol Rae; Wilde, Ralph; Bronkhorst, Daan; Moritz, Frederic A.; Rolle, Baptiste; Sherman, Rebecca; Southard, Jo Lynn; Wilkinson, Robert; Poole, Hilary, Ed.

    This reference work documents the history of human rights theory, explains each article of the Universal Declaration of Human Rights, explores the contemporary human rights movement, and examines the major human rights issues facing the world today. This book is the first to combine historical and contemporary perspectives on these critical…

  7. Alberta. Reference Series No. 26.

    ERIC Educational Resources Information Center

    Department of External Affairs, Ottawa (Ontario).

    This booklet, one of a series featuring the Canadian provinces, presents a brief overview of Alberta and is suitable for teacher reference or student reading. Separate sections discuss the history and population, the provincial government, the economy, transportation, communications, mineral resources, agriculture, manufacturing, forest products,…

  8. Mobile Technologies and Roving Reference

    ERIC Educational Resources Information Center

    Penner, Katherine

    2011-01-01

    As 21st century librarians, we have made apt adjustments for reaching out into the digital world, but we need to consider the students who still use library services within our walls. We can use available handheld, mobile technologies to help patrons too shy to approach the desk and free library staff to bring reference service directly to patrons.

  9. Expert Systems in Reference Services.

    ERIC Educational Resources Information Center

    Roysdon, Christine, Ed.; White, Howard D., Ed.

    1989-01-01

    Eleven articles introduce expert systems applications in library and information science, and present design and implementation issues of system development for reference services. Topics covered include knowledge based systems, prototype development, the use of artificial intelligence to remedy current system inadequacies, and an expert system to…

  10. Selected Reference Books of 1992.

    ERIC Educational Resources Information Center

    McIlvaine, Eileen

    1993-01-01

    Presents an annotated bibliography of 40 recent scholarly and general works of interest to reference workers in university libraries. Topics areas covered include philosophy, religion, language, literature, architecture, economics, law, area studies, Russia and the Soviet Union, women's studies, and Christopher Columbus. New editions and…

  11. 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

  12. Thermocouple, multiple junction reference oven

    NASA Technical Reports Server (NTRS)

    Leblanc, L. P. (Inventor)

    1981-01-01

    An improved oven for maintaining the junctions of a plurality of reference thermocouples at a common and constant temperature is described. The oven is characterized by a cylindrical body defining a heat sink with axially extended-cylindrical cavity a singularized heating element which comprises a unitary cylindrical heating element consisting of a resistance heating coil wound about the surface of metallic spool with an axial bore defined and seated in the cavity. Other features of the oven include an annular array of radially extended bores defined in the cylindrical body and a plurality of reference thermocouple junctions seated in the bores in uniformly spaced relation with the heating element, and a temperature sensing device seated in the axial bore for detecting temperature changes as they occur in the spool and circuit to apply a voltage across the coil in response to detected drops in temperatures of the spool.

  13. ADAM -- Interface Module Reference Manual

    NASA Astrophysics Data System (ADS)

    Chipperfield, A. J.; Kelly, B. D.; Wright, S. L.

    ADAM Interface Modules provide an interface between ADAM application programs and the rest of the system. This document describes in detail the facilities available with ADAM Interface Modules and the rules for using them. It is intended as a reference manual and should shed light on some of the finer points of the ADAM parameter system. Readers requiring an introduction to Interface Modules should read SG/4.

  14. HANFORD WASTE MINEROLOGY REFERENCE REPORT

    SciTech Connect

    DISSELKAMP RS

    2010-06-18

    This report lists the observed mineral phase phases present in the Hanford tanks. This task was accomplished by performing a review of numerous reports using experimental techniques including, but not limited to: x-ray diffraction, polarized light microscopy, scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, electron energy loss spectroscopy, and particle size distribution analyses. This report contains tables that can be used as a quick reference to identify the crystal phases present observed in Hanford waste.

  15. HANFORD WASTE MINERALOGY REFERENCE REPORT

    SciTech Connect

    DISSELKAMP RS

    2010-06-29

    This report lists the observed mineral phases present in the Hanford tanks. This task was accomplished by performing a review of numerous reports that used experimental techniques including, but not limited to: x-ray diffraction, polarized light microscopy, scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, electron energy loss spectroscopy, and particle size distribution analyses. This report contains tables that can be used as a quick reference to identify the crystal phases observed in Hanford waste.

  16. National Software Reference Library (NSRL)

    National Institute of Standards and Technology Data Gateway

    National Software Reference Library (NSRL) (PC database for purchase)   A collaboration of the National Institute of Standards and Technology (NIST), the National Institute of Justice (NIJ), the Federal Bureau of Investigation (FBI), the Defense Computer Forensics Laboratory (DCFL),the U.S. Customs Service, software vendors, and state and local law enforement organizations, the NSRL is a tool to assist in fighting crime involving computers.

  17. Microgrid cyber security reference architecture.

    SciTech Connect

    Veitch, Cynthia K.; Henry, Jordan M.; Richardson, Bryan T.; Hart, Derek H.

    2013-07-01

    This document describes a microgrid cyber security reference architecture. First, we present a high-level concept of operations for a microgrid, including operational modes, necessary power actors, and the communication protocols typically employed. We then describe our motivation for designing a secure microgrid; in particular, we provide general network and industrial control system (ICS)-speci c vulnerabilities, a threat model, information assurance compliance concerns, and design criteria for a microgrid control system network. Our design approach addresses these concerns by segmenting the microgrid control system network into enclaves, grouping enclaves into functional domains, and describing actor communication using data exchange attributes. We describe cyber actors that can help mitigate potential vulnerabilities, in addition to performance bene ts and vulnerability mitigation that may be realized using this reference architecture. To illustrate our design approach, we present a notional a microgrid control system network implementation, including types of communica- tion occurring on that network, example data exchange attributes for actors in the network, an example of how the network can be segmented to create enclaves and functional domains, and how cyber actors can be used to enforce network segmentation and provide the neces- sary level of security. Finally, we describe areas of focus for the further development of the reference architecture.

  18. Reference Inflow Characterization for River Resource Reference Model (RM2)

    SciTech Connect

    Neary, Vincent S

    2011-12-01

    Sandia National Laboratory (SNL) is leading an effort to develop reference models for marine and hydrokinetic technologies and wave and current energy resources. This effort will allow the refinement of technology design tools, accurate estimates of a baseline levelized cost of energy (LCoE), and the identification of the main cost drivers that need to be addressed to achieve a competitive LCoE. As part of this effort, Oak Ridge National Laboratory was charged with examining and reporting reference river inflow characteristics for reference model 2 (RM2). Published turbulent flow data from large rivers, a water supply canal and laboratory flumes, are reviewed to determine the range of velocities, turbulence intensities and turbulent stresses acting on hydrokinetic technologies, and also to evaluate the validity of classical models that describe the depth variation of the time-mean velocity and turbulent normal Reynolds stresses. The classical models are found to generally perform well in describing river inflow characteristics. A potential challenge in river inflow characterization, however, is the high variability of depth and flow over the design life of a hydrokinetic device. This variation can have significant effects on the inflow mean velocity and turbulence intensity experienced by stationary and bottom mounted hydrokinetic energy conversion devices, which requires further investigation, but are expected to have minimal effects on surface mounted devices like the vertical axis turbine device designed for RM2. A simple methodology for obtaining an approximate inflow characterization for surface deployed devices is developed using the relation umax=(7/6)V where V is the bulk velocity and umax is assumed to be the near-surface velocity. The application of this expression is recommended for deriving the local inflow velocity acting on the energy extraction planes of the RM2 vertical axis rotors, where V=Q/A can be calculated given a USGS gage flow time

  19. Sentinel 2 global reference image

    NASA Astrophysics Data System (ADS)

    Dechoz, C.; Poulain, V.; Massera, S.; Languille, F.; Greslou, D.; de Lussy, F.; Gaudel, A.; L'Helguen, C.; Picard, C.; Trémas, T.

    2015-10-01

    Sentinel-2 is a multispectral, high-resolution, optical imaging mission, developed by the European Space Agency (ESA) in the frame of the Copernicus program of the European Commission. In cooperation with ESA, the Centre National d'Etudes Spatiales (CNES) is responsible for the image quality of the project, and will ensure the CAL/VAL commissioning phase. Sentinel-2 mission is devoted the operational monitoring of land and coastal areas, and will provide a continuity of SPOT- and Landsat-type data. Sentinel-2 will also deliver information for emergency services. Launched in 2015 and 2016, there will be a constellation of 2 satellites on a polar sun-synchronous orbit, imaging systematically terrestrial surfaces with a revisit time of 5 days, in 13 spectral bands in visible and shortwave infra-red. Therefore, multi-temporal series of images, taken under the same viewing conditions, will be available. So as to ensure for the multi-temporal registration of the products, specified to be better than 0.3 pixels at 2σ, a Global Reference Image (GRI) will be produced during the CAL/VAL period. This GRI is composed of a set of Sentinel-2 acquisitions, which geometry has been corrected by bundle block adjustment. During L1B processing, Ground Control Points will be taken between this reference image and the sentinel-2 acquisition processed and the geometric model of the image corrected, so as to ensure the good multi-temporal registration. This paper first details the production of the reference during the CALVAL period, and then details the qualification and geolocation performance assessment of the GRI. It finally presents its use in the Level-1 processing chain and gives a first assessment of the multi-temporal registration.

  20. PVWatts Version 1 Technical Reference

    SciTech Connect

    Dobos, A. P.

    2013-10-01

    The NREL PVWatts(TM) calculator is a web application developed by the National Renewable Energy Laboratory (NREL) that estimates the electricity production of a grid-connected photovoltaic system based on a few simple inputs. PVWatts combines a number of sub-models to predict overall system performance, and makes several hidden assumptions about performance parameters. This technical reference details the individual sub-models, documents assumptions and hidden parameters, and explains the sequence of calculations that yield the final system performance estimation.

  1. Concepts in Geodetic Reference Frames.

    DTIC Science & Technology

    1979-10-01

    in the years 1900-1905. The period of C around the origin is about 1.2 years, the Chandler period. The z-axis of the terrestrial coordinate system is...multiplication by the matrix W which expresses the effect of polar motion, also called polar wobble . It has the form 1 0 x w= 0 1 -y (6-4) -x y 1 where...reference. The periods of forced rnotion (of F, H, and I) are about 1 day, the period of free motion of S, C’ and E0 around 0 is the Chandler period of

  2. [Dietary reference intakes of phosphorus].

    PubMed

    Uenishi, Kazuhiro

    2012-10-01

    Phosphorus (P) exists at the all organs and plays important physiological roles in the body. A wide range of food contains P, which is absorbed at a higher level (60-70%) and its insufficiency and deficiency are rarely found. P is used as food additives in many processed food, where risk of overconsumption could be an issue. P has less evidence in terms of nutrition. P has the adequate intake and the tolerable upper intake level, for risk reduction of health disorders associated with excess intake, at the Dietary Reference Intakes for Japanese (2010 edition).

  3. World Reference Center for Arboviruses.

    DTIC Science & Technology

    1984-01-01

    flaviviruses recovered from pooled mosquitoes from Indonesia. Mous Reference Viruses JKT Arbo # Mouse immune ascitic fluids JE ME SEP THU ZIKA 5441 7180 9092...320 ɝ ɝ ɝ 5 Zika ( ZIKA ) 5 5 20 5 10 5 JKT Arbo # 5441 320 20 ɝ 160 ND 20 * ND not done F’, 41 r .• As more workers begin to use mosquito cells for...80 ɠ/4096 ᝺/81920 Zika 16/512 20/160 ɠ/4096 320/81920 Jutiapa 16/512 10/320 8/4096 160/81920 Yakose -- ɠ/4096 20/81920 Aroa -- ɠ/4096 ᝺/81920

  4. Ozone reference models for CIRA

    NASA Astrophysics Data System (ADS)

    Keating, G. M.; Young, D. F.; Pitts, M. C.

    The data bases and computational techniques used in recent models of the O3 distribution in the earth atmosphere are described, summarizing the results of ongoing efforts to define an O3 reference model for incorporation into CIRA. Consideration is given to the analysis of data from satellite instruments (Nimbus 7 LIMS, TOMS, and SBUV; SME UVS and IR; and AE-2 SAGE) to construct models of total column O3 and vertical O3 structure. The satellite-based model predictions are then compared with balloon, rocket, and umkehr measurements in extensive graphs: good agreement is demonstrated both among the satellite data sets and between satellite and nonsatellite data sets.

  5. Empirical Reference Models for COSPAR International Reference Atmosphere (CIRA)

    NASA Astrophysics Data System (ADS)

    Drob, Douglas; Emmert, John; Picone, Michael

    Openly distributed atmospheric reference models are an essential tool for scientific research and operational activities. To meet the needs of all users, such models must utilize rigorous statistical methods and the most comprehensive and reliable data sets in their development. Two such models that meet these requirements are the Naval Research Laboratory, Mass Spectrometer Incoherent Scatter Extended (NRLMSISE-00) and Horizontal Wind Model (HWM-93) empirical reference models. The NRLMSISE-00 model and its predecessors are based on 35 years of empirical modeling experience and over 40 years of research measurements. These global models are well documented and extend from the ground to the exosphere, providing estimates of neutral temperature, density, and major neutral species composition as a function of geographic location, day of year, time of day, and geomagnetic and solar activity conditions. Relative to the most comprehensive span of datasets available these models have the smallest bias and root mean square deviations of any climatological reference model built to date, although there are a few limitations in the 80 to 120 km region. The less advanced HWM-93 model, based on the same statistical methodologies and general mathematical formulation of the NRLMSISE-00 model, provides climatological estimates of the horizontal wind fields over the same variables and range of conditions as the NRLMSISE-00 model. The availability of several new long term data sets, including satellite wind measurements from the WINDII instrument onboard the UARS satellite, as well as ground-based optical Fabery-Perot measurements, provide the opportunity to make significant refinements to the existing model. Initial results from an improved HWM will be shown for altitudes between 100 and 500 km. Improvement in the model's ability to represent the seasonal changes, solar forcing, geomagnetic forcing, diurnal variation, and vertical structure of horizontal winds of the region is

  6. Two-phase flow regimes and carry-over in a large-diameter model of a PWR hot leg. Final report

    SciTech Connect

    Hashemi, A.

    1986-04-01

    This report describes a series of tests investigating two-phase flow characterization and carryover in a transparent model of a Babcock and Wilson (B and W) Pressurized Water Reactor (PWR) hot leg geometry. This work was performed, inpart, to support the interpretation of results from the Once-Through Integral System (OTIS) and Multi-loop Integral Test (MIST) facilities. Test conditions were selected to cover a wide range of gas and liquid superficial velocities (0.01 m/s < j/sub g/ < 2 m/s, 0 < j/sub l/ < 0.5 m/s) expected to occur in a prototypical reactor geometry during a small break loss of coolant accident (SBLOCA). Tests at high gas superficial velocities (j/sub g/ > 2 m/s) were also performed for comparison with semi-analytical predictions. Tests were conducted in two different test rigs, one with 10.2-cm (4-inch) diameter pipe, and the other with 30.5-cm (12-inch) diameter pipe. Results include average void fraction, amount of water carryover through the U-bend, transient flow rates and pressure histories, and video movies of the two-phase flow phenomena. Results of the 10.2-cm (4-inch) pipe tests show generally good agreement with the Taitel and Dukler (1) flow regime map for vertical pipes. For the 30.5-cm pipe tests, slug flow was not observed. Instead, as the air flow rate was increased, the flow regime progressed from bubbly to churn-type flow with the presence of large bubbles (approximately 15-cm diameter). The results also indicate that flow regimes and collapsed liquid level are more strongly dependent on air superficial velocity than the water superficial velocity and that the amount of water carryover for a given air flow rate is a strong function of collapsed water level (void fraction). Furthermore, the results show that similar thresholds for breakdown in natural circulation flow exist between the 10.2-cm and 30.5-cm pipe tests for gas and liquid superficial velocities expected in a SBLOCA. 20 refs., 24 figs.

  7. 40 CFR 1508.24 - Referring agency.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Referring agency. 1508.24 Section 1508.24 Protection of Environment COUNCIL ON ENVIRONMENTAL QUALITY TERMINOLOGY AND INDEX § 1508.24 Referring agency. Referring agency means the federal agency which has referred any matter to the...

  8. Virtual Reference for a Real Public.

    ERIC Educational Resources Information Center

    McDermott, Irene E.

    1998-01-01

    Reviews World Wide Web sites useful as alternative resources for reference librarians. Sites described are: general reference; reference for kids and teens; regional interest for Southern California, including local foreign-language resources and local history sites; and interactive reference. (JAK)

  9. The Art of Collection Development: Reference Style.

    ERIC Educational Resources Information Center

    Schmitt, John P.

    1998-01-01

    Discusses selecting for a reference collection, creative budgeting, cutting a deal, collection awareness (strengths/needs), Web site reviews, R-Net (reviews from diverse areas and institutions), and print vs. electronic reference products. Reference librarian adhere to high standards for reference book and Web sites, teach assessment techniques,…

  10. Instant Messaging Reference: How Does It Compare?

    ERIC Educational Resources Information Center

    Desai, Christina M.

    2003-01-01

    Compares a digital reference service that uses instant messaging with traditional, face-to-face reference based on experiences at the Southern Illinois University library. Addresses differences in reference questions asked, changes in the reference transaction, student expectations, bibliographic instruction, and librarian attitudes and procedures…

  11. Building and Maintaining Digital Reference Services.

    ERIC Educational Resources Information Center

    Wasik, Joann M.

    2000-01-01

    Discusses digital reference services that provide subject expertise and information referral over the Internet to their users. Describes the history of digital reference, how digital reference services work, and explains a six-step process for building and maintaining digital reference services, including training, planning, and evaluating. (LRW)

  12. Reference electrodes for underground storage tanks

    SciTech Connect

    Ansuini, F.J.; Dimond, J.R.

    1995-12-31

    This paper discusses several factors affecting the reference potential established by copper/copper sulfate and silver/silver chloride reference electrodes. Guidelines for using permanent references in underground storage tank applications are presented and some causes of misleading readings with portable references are discussed.

  13. Hypertext Cross-Reference Utilities

    NASA Astrophysics Data System (ADS)

    Warren-Smith, R. F.; Draper, P. W.

    This document describes a set of ``Hypertext Cross-Reference Utilities'' (HTX) which are designed to help maintain large documentation sets whose constituent documents are written using the Hypertext Markup Languagee (HTML). The central part of HTX is a hypertext linker, hlink. This allows hyper-links (or cross-references) to be established between related documents in such a way that it is easy to maintain their integrity as individual documents are updated. Information produced by this linking process is also used by other HTX utilities to provide document search facilities and the ability to randomly access any part of a documentation set. This latter capability forms a basis for constructing hypertext help systems for use by other software. The expected readership of this document includes those who read hypertext documentation, those who write it, and those who maintain it, especially those who write and maintain Starlink documentation. Software developers may also be interested in the possibilities for hypertext help that HTX provides.

  14. Event boundaries and anaphoric reference.

    PubMed

    Thompson, Alexis N; Radvansky, Gabriel A

    2016-06-01

    The current study explored the finding that parsing a narrative into separate events impairs anaphor resolution. According to the Event Horizon Model, when a narrative event boundary is encountered, a new event model is created. Information associated with the prior event model is removed from working memory. So long as the event model containing the anaphor referent is currently being processed, this information should still be available when there is no narrative event boundary, even if reading has been disrupted by a working-memory-clearing distractor task. In those cases, readers may reactivate their prior event model, and anaphor resolution would not be affected. Alternatively, comprehension may not be as event oriented as this account suggests. Instead, any disruption of the contents of working memory during comprehension, event related or not, may be sufficient to disrupt anaphor resolution. In this case, reading comprehension would be more strongly guided by other, more basic language processing mechanisms and the event structure of the described events would play a more minor role. In the current experiments, participants were given stories to read in which we included, between the anaphor and its referent, either the presence of a narrative event boundary (Experiment 1) or a narrative event boundary along with a working-memory-clearing distractor task (Experiment 2). The results showed that anaphor resolution was affected by narrative event boundaries but not by a working-memory-clearing distractor task. This is interpreted as being consistent with the Event Horizon Model of event cognition.

  15. Reference-assisted chromosome assembly.

    PubMed

    Kim, Jaebum; Larkin, Denis M; Cai, Qingle; Asan; Zhang, Yongfen; Ge, Ri-Li; Auvil, Loretta; Capitanu, Boris; Zhang, Guojie; Lewin, Harris A; Ma, Jian

    2013-01-29

    One of the most difficult problems in modern genomics is the assembly of full-length chromosomes using next generation sequencing (NGS) data. To address this problem, we developed "reference-assisted chromosome assembly" (RACA), an algorithm to reliably order and orient sequence scaffolds generated by NGS and assemblers into longer chromosomal fragments using comparative genome information and paired-end reads. Evaluation of results using simulated and real genome assemblies indicates that our approach can substantially improve genomes generated by a wide variety of de novo assemblers if a good reference assembly of a closely related species and outgroup genomes are available. We used RACA to reconstruct 60 Tibetan antelope (Pantholops hodgsonii) chromosome fragments from 1,434 SOAPdenovo sequence scaffolds, of which 16 chromosome fragments were homologous to complete cattle chromosomes. Experimental validation by PCR showed that predictions made by RACA are highly accurate. Our results indicate that RACA will significantly facilitate the study of chromosome evolution and genome rearrangements for the large number of genomes being sequenced by NGS that do not have a genetic or physical map.

  16. Coastwide Reference Monitoring System (CRMS)

    USGS Publications Warehouse

    2010-01-01

    In 1990, the U.S. Congress enacted the Coastal Wetlands Planning, Protection and Restoration Act (CWPPRA) in response to growing awareness of a land loss crisis in Louisiana. Projects funded by CWPPRA require monitoring and evaluation of project effectiveness, and there is also a need to assess the cumulative effects of all projects to achieve a sustainable coastal environment. In 2003, the Louisiana Office of Coastal Protection and Restoration (OCPR) and the U.S. Geological Survey (USGS) received approval from the CWPPRA Task Force to implement the Coastwide Reference Monitoring System (CRMS) as a mechanism to monitor and evaluate the effectiveness of CWPPRA projects at the project, region, and coastwide levels. The CRMS design implements a multiple reference approach by using aspects of hydrogeomorphic functional assessments and probabilistic sampling. The CRMS program is as dynamic as the coastal habitats it monitors. The program is currently funded through CWPPRA and provides data for a variety of user groups, including resource managers, academics, landowners, and researchers.

  17. Speaker Adaptation Using Multiple Reference Speakers

    DTIC Science & Technology

    1989-01-01

    to other methods that use a pooled reference model , this technique normalizes the training speech from multiple reference speakers to a single com...training the reference hidden Markov model (HMM). Our usual prohabilistic spectrum transformation can be applied to the reference HMM to model a new...trained phonetic hidden Markov models of a single reference speaker so that they were appropriate for a new (target) speaker. This method reduced the

  18. NUCLEAR SCIENCE REFERENCES CODING MANUAL

    SciTech Connect

    WINCHELL,D.F.

    2007-04-01

    This manual is intended as a guide for Nuclear Science References (NSR) compilers. The basic conventions followed at the National Nuclear Data Center (NNDC), which are compatible with the maintenance and updating of and retrieval from the Nuclear Science References (NSR) file, are outlined. The NSR database originated at the Nuclear Data Project (NDP) at Oak Ridge National Laboratory as part of a project for systematic evaluation of nuclear structure data.1 Each entry in this computer file corresponds to a bibliographic reference that is uniquely identified by a Keynumber and is describable by a Topic and Keywords. It has been used since 1969 to produce bibliographic citations for evaluations published in Nuclear Data Sheets. Periodic additions to the file were published as the ''Recent References'' issues of Nuclear Data Sheets prior to 2005. In October 1980, the maintenance and updating of the NSR file became the responsibility of the NNDC at Brookhaven National Laboratory. The basic structure and contents of the NSR file remained unchanged during the transfer. In Chapter 2, the elements of the NSR file such as the valid record identifiers, record contents, and text fields are enumerated. Relevant comments regarding a new entry into the NSR file and assignment of a keynumber are also given in Chapter 2. In Chapter 3, the format for keyword abstracts is given followed by specific examples; for each TOPIC, the criteria for inclusion of an article as an entry into the NSR file as well as coding procedures are described. Authors preparing Keyword abstracts either to be published in a Journal (e.g., Nucl. Phys. A) or to be sent directly to NNDC (e.g., Phys. Rev. C) should follow the illustrations in Chapter 3. The scope of 1See W.B.Ewbank, ORNL-5397 (1978). the literature covered at the NNDC, the categorization into Primary and Secondary sources, etc., is discussed in Chapter 4. Useful information regarding permitted character sets, recommended abbreviations, etc., is

  19. Primary Atomic Clock Reference System

    NASA Technical Reports Server (NTRS)

    2001-01-01

    An artist's concept of the Primary Atomic Clock Reference System (PARCS) plarned to fly on the International Space Station (ISS). PARCS will make even more accurate atomic time available to everyone, from physicists testing Einstein's Theory of Relativity, to hikers using the Global Positioning System to find their way. In ground-based atomic clocks, lasers are used to cool and nearly stop atoms of cesium whose vibrations are used as the time base. The microgravity of space will allow the atoms to be suspended in the clock rather than circulated in an atomic fountain, as required on Earth. PARCS is being developed by the Jet Propulsion Laboratory with principal investigators at the National Institutes of Standards and Technology and the University of Colorado, Boulder. See also No. 0103191

  20. Primary Atomic Clock Reference System

    NASA Technical Reports Server (NTRS)

    2001-01-01

    An artist's concept of the Primary Atomic Clock Reference System (PARCS) plarned to fly on the International Space Station (ISS). PARCS will make even more accurate atomic time available to everyone, from physicists testing Einstein's Theory of Relativity, to hikers using the Global Positioning System to find their way. In ground-based atomic clocks, lasers are used to cool and nearly stop atoms of cesium whose vibrations are used as the time base. The microgravity of space will allow the atoms to be suspended in the clock rather than circulated in an atomic fountain, as required on Earth. PARCS is being developed by the Jet Propulsion Laboratory with principal investigators at the National Institutes of Standards and Technology and the University of Colorado, Boulder. See also No. 0100120.

  1. Very preliminary reference Moon model

    NASA Astrophysics Data System (ADS)

    Garcia, Raphaël F.; Gagnepain-Beyneix, Jeannine; Chevrot, Sébastien; Lognonné, Philippe

    2011-09-01

    The deep structure of the Moon is a missing piece to understand the formation and evolution of the Earth-Moon system. Despite the great amount of information brought by the Apollo passive seismic experiment (ALSEP), the lunar structure below deep moonquakes, which occur around 900 km depth, remains largely unknown. We construct a reference Moon model which incorporates physical constraints, and fits both geodesic (lunar mass and polar moment of inertia, and Love numbers) and seismological (body wave arrivals measured by Apollo network) data. In this model, the core radius is constrained by the detection of S waves reflected from the core. In a first step, for each core radius, a radial model of the lunar interior, including P and S wave velocities and density, is inverted from seismic and geodesic data. In a second step, the core radius is determined from the detection of shear waves reflected on the lunar core by waveform stacking of deep moonquake Apollo records. This detection has been made possible by careful data selection and processing, including a correction of the gain of horizontal sensors based on the principle of energy equipartition inside the coda of lunar seismic records, and a precise alignment of SH waveforms by a non-linear inversion method. The Very Preliminary REference MOON model (VPREMOON) obtained here has a core radius of 380 ± 40 km and an average core mass density of 5200 ± 1000 kg/m 3. The large error bars on these estimates are due to the poorly constrained S-wave velocity profile at the base of the mantle and to mislocation errors of deep moonquakes. The detection of horizontally polarized S waves reflected from the core and the absence of detection of vertically polarized S waves favour a liquid state in the outermost part of the core. All these results are consistent, within their error bars, with previous estimates based on lunar rotation dissipation ( Williams et al., 2001) and on lunar induced magnetic moment ( Hood et al., 1999).

  2. Reference materials for cellular therapeutics.

    PubMed

    Bravery, Christopher A; French, Anna

    2014-09-01

    The development of cellular therapeutics (CTP) takes place over many years, and, where successful, the developer will anticipate the product to be in clinical use for decades. Successful demonstration of manufacturing and quality consistency is dependent on the use of complex analytical methods; thus, the risk of process and method drift over time is high. The use of reference materials (RM) is an established scientific principle and as such also a regulatory requirement. The various uses of RM in the context of CTP manufacturing and quality are discussed, along with why they are needed for living cell products and the analytical methods applied to them. Relatively few consensus RM exist that are suitable for even common methods used by CTP developers, such as flow cytometry. Others have also identified this need and made proposals; however, great care will be needed to ensure any consensus RM that result are fit for purpose. Such consensus RM probably will need to be applied to specific standardized methods, and the idea that a single RM can have wide applicability is challenged. Written standards, including standardized methods, together with appropriate measurement RM are probably the most appropriate way to define specific starting cell types. The characteristics of a specific CTP will to some degree deviate from those of the starting cells; consequently, a product RM remains the best solution where feasible. Each CTP developer must consider how and what types of RM should be used to ensure the reliability of their own analytical measurements.

  3. A reference architecture for telemonitoring.

    PubMed

    Clarke, Malcolm

    2004-01-01

    The Telecare Interactive Continuous Monitoring System exploits GPRS to provide an ambulatory device that monitors selected vital signs on a continuous basis. Alarms are sent when parameters fall outside preset limits, and accompanying physiological data may also be transmitted. The always-connected property of GPRS allows continuous interactive control of the device and its sensors, permitting changes to monitoring parameters or even enabling continuous monitoring of a sensor in emergency. A new personal area network (PAN) has been developed to support short-range wireless connection to sensors worn on the body including ECG and finger worn SpO2. Most notable is use of ultra low radio frequency to reduce power to minimum. The system has been designed to use a hierarchical architecture for sensors and "derived" signals, such as HR from ECG, so that each can be independently controlled and managed. Sensors are treated as objects, and functions are defined to control aspects of behaviour. These are refined in order to define a generic set of abstract functions to handle the majority of functions, leaving a minimum of sensor specific commands. The intention is to define a reference architecture in order to research the functionality and system architecture of a telemonitoring system. The Telecare project is funded through a grant from the European Commission (IST programme).

  4. Gender agreement and multiple referents.

    PubMed

    Finocchiaro, Chiara; Mahon, Bradford Z; Caramazza, Alfonso

    2008-01-01

    We report a new pattern of usage in current, spoken Italian that has implications for both psycholinguistic models of language production and linguistic theories of language change. In Italian, gender agreement is mandatory for both singular and plural nouns. However, when two or more nouns of different grammatical gender appear in a conjoined noun phrase (NP), masculine plural agreement is required. In this study, we combined on-line and off-line methodologies in order to assess the mechanisms involved in gender marking in the context of multiple referents. The results of two pronoun production tasks showed that plural feminine agreement was significantly more difficult than plural masculine agreement. In a separate study using offline judgements of acceptability, we found that agreement violations in Italian are tolerated more readily in the case of feminine conjoined noun phrases (e.g., la mela e la banana 'the:fem apple:fem and the: fem banana: fem') than masculine conjoined noun phrases (e.g., il fiore e il libro 'the:mas flower: mas and the:mas book:mas'). Implications of these results are discussed both at the level of functional architecture within the language production system and at the level of changes in language use.

  5. ILC cryogenic systems reference design

    SciTech Connect

    Peterson, T.J.; Geynisman, M.; Klebaner, A.; Theilacker, J.; Parma, V.; Tavian, L.; /CERN

    2008-01-01

    A Global Design Effort (GDE) began in 2005 to study a TeV scale electron-positron linear accelerator based on superconducting radio-frequency (RF) technology, called the International Linear Collider (ILC). In early 2007, the design effort culminated in a reference design for the ILC, closely based on the earlier TESLA design. The ILC will consist of two 250 GeV linacs, which provide positron-electron collisions for high energy physics research. The particle beams will be accelerated to their final energy in superconducting niobium RF cavities operating at 2 kelvin. At a length of about 12 km each, the main linacs will be the largest cryogenic systems in the ILC. Positron and electron sources, damping rings, and beam delivery systems will also have a large number and variety of other superconducting RF cavities and magnets, which require cooling at liquid helium temperatures. Ten large cryogenic plants with 2 kelvin refrigeration are envisioned to cool the main linacs and the electron and positron sources. Three smaller cryogenic plants will cool the damping rings and beam delivery system components predominately at 4.5 K. This paper describes the cryogenic systems concepts for the ILC.

  6. The Reference Forward Model (RFM)

    NASA Astrophysics Data System (ADS)

    Dudhia, Anu

    2017-01-01

    The Reference Forward Model (RFM) is a general purpose line-by-line radiative transfer model, currently supported by the UK National Centre for Earth Observation. This paper outlines the algorithms used by the RFM, focusing on standard calculations of terrestrial atmospheric infrared spectra followed by a brief summary of some additional capabilities and extensions to microwave wavelengths and extraterrestrial atmospheres. At its most basic level - the 'line-by-line' component - it calculates molecular absorption cross-sections by applying the Voigt lineshape to all transitions up to ±25 cm-1 from line-centre. Alternatively, absorptions can be directly interpolated from various forms of tabulated data. These cross-sections are then used to construct infrared radiance or transmittance spectra for ray paths through homogeneous cells, plane-parallel or circular atmospheres. At a higher level, the RFM can apply instrumental convolutions to simulate measurements from Fourier transform spectrometers. It can also calculate Jacobian spectra and so act as a stand-alone forward model within a retrieval scheme. The RFM is designed for robustness, flexibility and ease-of-use (particularly by the non-expert), and no claims are made for superior accuracy, or indeed novelty, compared to other line-by-line codes. Its main limitations at present are a lack of scattering and simplified modelling of surface reflectance and line-mixing.

  7. Gender agreement and multiple referents

    PubMed Central

    Finocchiaro, Chiara; Mahon, Bradford Z.; Caramazza, Alfonso

    2010-01-01

    We report a new pattern of usage in current, spoken Italian that has implications for both psycholinguistic models of language production and linguistic theories of language change. In Italian, gender agreement is mandatory for both singular and plural nouns. However, when two or more nouns of different grammatical gender appear in a conjoined noun phrase (NP), masculine plural agreement is required. In this study, we combined on-line and off-line methodologies in order to assess the mechanisms involved in gender marking in the context of multiple referents. The results of two pronoun production tasks showed that plural feminine agreement was significantly more difficult than plural masculine agreement. In a separate study using offline judgements of acceptability, we found that agreement violations in Italian are tolerated more readily in the case of feminine conjoined noun phrases (e.g., la mela e la banana ‘the:fem apple:fem and the: fem banana: fem’) than masculine conjoined noun phrases (e.g., il fiore e il libro ‘the:mas flower: mas and the:mas book:mas’). Implications of these results are discussed both at the level of functional architecture within the language production system and at the level of changes in language use.* PMID:21037930

  8. Establishment of reference standards in biosimilar studies.

    PubMed

    Zhang, Aijing; Tzeng, Jung-Ying; Chow, Shein-Chung

    2013-07-31

    When an innovative biological product goes off-patent, biopharmaceutical or biotechnological companies may file an application for regulatory approval of biosimilar products. In practice, however, important information on the innovative (reference) product may not be available for assessment. Thus, it is important to first establish a reference standard while assessing biosimilarity between a biosimilar product and the reference product. In this paper, reference standard is established through the biosimilarity index approach based on a reference-replicated study (or R-R study), in which the reference product is compared with itself under various scenarios. The reference standard can then be used for assessing the degree of similarity between the test and reference drugs in biosimilar studies.

  9. PWR FLECHT SEASET 21-rod-bundle flow-blockage task: data and analysis report. NRC/EPRI/Westinghouse report No. 11, main report and appendices A-J

    SciTech Connect

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  10. 33 CFR 242.3 - References.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 3 2014-07-01 2014-07-01 false References. 242.3 Section 242.3 Navigation and Navigable Waters CORPS OF ENGINEERS, DEPARTMENT OF THE ARMY, DEPARTMENT OF DEFENSE FLOOD PLAIN MANAGEMENT SERVICES PROGRAM ESTABLISHMENT OF FEES FOR COST RECOVERY § 242.3 References. The references...

  11. 33 CFR 242.3 - References.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 33 Navigation and Navigable Waters 3 2012-07-01 2012-07-01 false References. 242.3 Section 242.3 Navigation and Navigable Waters CORPS OF ENGINEERS, DEPARTMENT OF THE ARMY, DEPARTMENT OF DEFENSE FLOOD PLAIN MANAGEMENT SERVICES PROGRAM ESTABLISHMENT OF FEES FOR COST RECOVERY § 242.3 References. The references...

  12. Suggested References. PACE I.D. Center.

    ERIC Educational Resources Information Center

    Bergson, Rita

    During the course of the PACE I.D. Center Project, 1966-1969, staff members recorded references that they felt contributed to the general knowledge of the prevention of learning and behavior problems. More specifically, those references that implied concern for the child in his total environment were considered most relevant. The references are…

  13. Virtual Reference Services: Directions and Agendas.

    ERIC Educational Resources Information Center

    Gray, Suzanne M.

    2000-01-01

    Analyzes Web sites of ten large research libraries in order to establish a basic understanding of how some major libraries are currently providing virtual reference services. Discusses centralization; email reference guidelines; creating a service economy; expanding hours of service; virtual users' needs; and real-time virtual reference.…

  14. Rethinking Job References: A Networking Challenge

    ERIC Educational Resources Information Center

    Muir, Clive

    2009-01-01

    Can job references play an active role in shaping one's career plans? Would individuals consider their references as part of their personal and professional network? Although most professionals may respond with a resounding "Yes, of course!" to these questions, the author realized that many of his students were skeptical about job references. To…

  15. Environmental Sciences Reference Sources. An Annotated Bibliography.

    ERIC Educational Resources Information Center

    McMartin, Mary I., Comp.

    This list of Environmental Sciences References Sources is intended to give undergraduate and graduate students a starting point when searching for information in the library. Entries are grouped according to type of reference material and then are listed in alphabetical order. The types of reference material included are guides to dictionaries,…

  16. Frames of Reference in the Classroom

    ERIC Educational Resources Information Center

    Grossman, Joshua

    2012-01-01

    The classic film "Frames of Reference" effectively illustrates concepts involved with inertial and non-inertial reference frames. In it, Donald G. Ivey and Patterson Hume use the cameras perspective to allow the viewer to see motion in reference frames translating with a constant velocity, translating while accelerating, and rotating--all with…

  17. Applying Information Competency to Digital Reference.

    ERIC Educational Resources Information Center

    Ellis, Lisa; Francoeur, Stephen

    This paper presents a case for applying information competency (IC) standards to digital reference services at academic libraries. Practical reasons for applying standards or guidelines to e-mail and online chat reference services are given with some insight to the nature of digital reference interactions. The standards that arose from the…

  18. Quality Standards for Digital Reference Consortia.

    ERIC Educational Resources Information Center

    Kasowitz, Abby; Bennett, Blythe; Lankes, R. David

    2000-01-01

    Identifies a working set of standards by which to assess individual digital reference services (Internet-based human-mediated information services) and to define membership within a collaborative network of digital reference services. The standards are designed for the Virtual Reference Desk AskA Consortium. (Author/LRW)

  19. Reference and the First Person Pronoun.

    ERIC Educational Resources Information Center

    Glock, Hans-Johann; Hacker, P. M. S.

    1996-01-01

    Maintains that deciding whether the first person pronoun is a referring expression requires clarity about the role of "I" and a detailed account of the notion of reference. It is concluded that "I" is a limiting case of reference, in which the possibility of referential failure and misidentification does not apply. (24…

  20. 40 CFR 1043.100 - Reference materials.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 33 2014-07-01 2014-07-01 false Reference materials. 1043.100 Section... § 1043.100 Reference materials. Documents listed in this section have been incorporated by reference into... the availability of this material at NARA, call 202-741-6030, or go to:...

  1. 40 CFR 1042.910 - Reference materials.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 33 2014-07-01 2014-07-01 false Reference materials. 1042.910 Section... Other Reference Information § 1042.910 Reference materials. Documents listed in this section have been... information on the availability of this material at NARA, call 202-741-6030, or go to:...

  2. 40 CFR 1042.910 - Reference materials.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Reference materials. 1042.910 Section... Other Reference Information § 1042.910 Reference materials. Documents listed in this section have been... information on the availability of this material at NARA, call 202-741-6030, or go to:...

  3. 40 CFR 94.5 - Reference materials.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 20 2011-07-01 2011-07-01 false Reference materials. 94.5 Section 94.5... Compression-Ignition Marine Engines § 94.5 Reference materials. We have incorporated by reference the... availability of this material at NARA, call 202-741-6030, or go to:...

  4. 40 CFR 1043.100 - Reference materials.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Reference materials. 1043.100 Section... § 1043.100 Reference materials. Documents listed in this section have been incorporated by reference into... the availability of this material at NARA, call 202-741-6030, or go to:...

  5. 40 CFR 1043.100 - Reference materials.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 34 2012-07-01 2012-07-01 false Reference materials. 1043.100 Section... § 1043.100 Reference materials. Documents listed in this section have been incorporated by reference into... the availability of this material at NARA, call 202-741-6030, or go to:...

  6. 40 CFR 1042.910 - Reference materials.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 34 2012-07-01 2012-07-01 false Reference materials. 1042.910 Section... Other Reference Information § 1042.910 Reference materials. Documents listed in this section have been... information on the availability of this material at NARA, call 202-741-6030, or go to:...

  7. Reference electrodes for aboveground storage tanks

    SciTech Connect

    Ansuini, F.J.; Dimond, J.R.

    1995-12-31

    This paper discusses several factors affecting the reference potential established by copper/copper sulfate and silver/silver chloride reference electrodes. Guidelines for using references in aboveground storage tank applications are presented and some causes of misleading readings are discussed.

  8. Evaluating Reference Services and Reference Personnel: Questions and Answers from the Literature.

    ERIC Educational Resources Information Center

    Bunge, Charles A.

    1994-01-01

    This bibliographic essay discusses a variety of items from the literature on the evaluation of reference services and reference personnel. A model for the evaluation process is presented on which the literature review is based. (Contains 48 references.) (LRW)

  9. Light-water reactors: preliminary safety and environmental information document. Volume I

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  10. Writing More Informative Letters of Reference

    PubMed Central

    Wright, Scott M; Ziegelstein, Roy C

    2004-01-01

    Writing a meaningful and valuable letter of reference is not an easy task. Several factors influence the quality of any letter of reference. First, the accuracy and reliability of the writer's impressions and judgment depend on how well he knows the individual being described. Second, the writer's frame of reference, which is determined by the number of persons at the same level that he has worked with, will impact the context and significance of his beliefs and estimations. Third, the letter-writing skills of the person composing the letter will naturally affect the letter. To support the other components of a candidate's application, a letter of reference should provide specific examples of how an individual's behavior or attitude compares to a reference group and should assess “intangibles” that are hard to glean from a curriculum vitae or from test scores. This report offers suggestions that should help physicians write more informative letters of reference. PMID:15109330

  11. Quantum metrology in coarsened measurement reference

    NASA Astrophysics Data System (ADS)

    Xie, Dong; Xu, Chunling; Wang, An Min

    2017-01-01

    We investigate the role of coarsened measurement reference, which originates from the coarsened reference time and basis, in quantum metrology. When the measurement is based on one common reference basis, the disadvantage of coarsened measurement can be removed by symmetry. Owing to the coarsened reference basis, the entangled state cannot perform better than the product state for a large number of probe particles in estimating the phase. Given a finite uncertainty of the coarsened reference basis, the optimal number of probe particles is obtained. Finally, we prove that the maximally entangled state always achieves better frequency precision in the case of non-Markovian dephasing than that in the case of Markovian dephasing. The product state is more resistant to the interference of the coarsened reference time than the entangled state.

  12. Minimize reference sideband generation in microwave PLLs

    NASA Astrophysics Data System (ADS)

    Goldman, Stan

    1991-02-01

    The processes responsible for producing reference sidebands are outlined, and the sources of coupling to the microwave voltage-controlled oscillator (VCO) tune line including power-supply-generated signals, TTL-controlled interface signals, intermediate programmable-divider signals, and radiated TTL signals are discussed. It is noted that filtering alone is inadequate for reference-sideband suppression, while minimizing the tuning slope and maximizing the reference frequency will result in a reduced reference-sideband level. Minimizing offset currents by using a differential amplifier connection may reduce the reference-sideband level aggravated by an opamp. The selection of a TTL, ECL, or GaAs phase/frequency detector can determine the level of reference sidebands, as well as PCB isolation techniques.

  13. ["Pro Ana": Psychodynamic References for Anorexia Nervosa].

    PubMed

    Siefert, Linda

    2017-02-01

    "Pro Ana": Psychodynamic References for Anorexia Nervosa The internet-based phenomenon "Pro Ana" refers to the eating disorder anorexia nervosa in a positive way. To understand what the phenomenon "Pro Ana" represents, the websites are used as a starting point of the current analysis. Based on these results, similarities and differences between "Pro Ana" and the eating disorder anorexia nervosa are discussed. Furthermore psychodynamic references for anorexia nervosa are derived and finally their importance for treatment motivation will be considered.

  14. Biological and environmental reference materials: Update 1996

    NASA Astrophysics Data System (ADS)

    Roelandts, Iwan

    1997-07-01

    The present column lists additional biological and environmental reference samples. Organs, tissues, body fluids, plant materials, foods, fuels, ashes, dusts, particulate matter, gas mixtures, oils, soils, sediments, sludges and waters have been considered. Three tables are included that provide an easy-to-use survey. The following information is covered: the name of the material, the sample code, the producer, the reference to certification, the names and addresses of the suppliers from whom the reference material may be obtained, and specific remarks.

  15. A standard satellite control reference model

    NASA Technical Reports Server (NTRS)

    Golden, Constance

    1994-01-01

    This paper describes a Satellite Control Reference Model that provides the basis for an approach to identify where standards would be beneficial in supporting space operations functions. The background and context for the development of the model and the approach are described. A process for using this reference model to trace top level interoperability directives to specific sets of engineering interface standards that must be implemented to meet these directives is discussed. Issues in developing a 'universal' reference model are also identified.

  16. An improved hypothetical reference decoder for HEVC

    NASA Astrophysics Data System (ADS)

    Deshpande, Sachin; Hannuksela, Miska M.; Kazui, Kimihiko; Schierl, Thomas

    2013-02-01

    Hypothetical Reference Decoder is a hypothetical decoder model that specifies constraints on the variability of conforming network abstraction layer unit streams or conforming byte streams that an encoding process may produce. High Efficiency Video Coding (HEVC) builds upon and improves the design of the generalized hypothetical reference decoder of H.264/ AVC. This paper describes some of the main improvements of hypothetical reference decoder of HEVC.

  17. Reference materials for new psychoactive substances.

    PubMed

    Archer, Roland P; Treble, Ric; Williams, Keith

    2011-01-01

    Historically, the appearance of new psychoactive materials (and hence the requirement for new reference standards) has been relatively slow. This position has now changed, with 101 new psychoactive substances reported to EMCDDA-Europol since 2006. The newly reported materials, and associated metabolites, require properly certified reference materials to permit reliable identification and quantification. The traditional approach and timescales of reference material production and certification are being increasingly challenged by the appearance of these new substances. Reference material suppliers have to adopt new strategies to meet the needs of laboratories. This situation is particularly challenging for toxicology standards as the metabolism of many of these substances is initially unknown. Reference material production often involves synthesis from first principles. While it is possible to synthesis these materials, there can be significant difficulties, from synthetic complexities through to the need to use controlled materials. These issues are examined through a discussion of the synthesis of cathinones. Use of alternative sources, including pharmaceutical impurity materials or internet sourced products, as starting materials for conversion into appropriately certified reference materials are also discussed. The sudden appearance and sometimes brief lifetime in the market place of many of these novel legal highs or research chemicals present commercial difficulties for reference material producers. The need for collaboration at all levels is highlighted as essential to rapid identification of requirements for new reference materials. National or international commissioning or support may also be required to permit reference material producers to recover their development costs.

  18. Resources and References for Earth Science Teachers

    ERIC Educational Resources Information Center

    Wall, Charles A.; Wall, Janet E.

    1976-01-01

    Listed are resources and references for earth science teachers including doctoral research, new textbooks, and professional literature in astronomy, space science, earth science, geology, meteorology, and oceanography. (SL)

  19. 44 CFR 59.4 - References.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... INSURANCE AND HAZARD MITIGATION National Flood Insurance Program GENERAL PROVISIONS General § 59.4 References. (a) The following are statutory references for the National Flood Insurance Program, under which these regulations are issued: (1) National Flood Insurance Act of 1968 (title XIII of the Housing...

  20. 44 CFR 59.4 - References.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... INSURANCE AND HAZARD MITIGATION National Flood Insurance Program GENERAL PROVISIONS General § 59.4 References. (a) The following are statutory references for the National Flood Insurance Program, under which these regulations are issued: (1) National Flood Insurance Act of 1968 (title XIII of the Housing...

  1. 4 CFR 2.2 - References.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 4 Accounts 1 2012-01-01 2012-01-01 false References. 2.2 Section 2.2 Accounts GOVERNMENT ACCOUNTABILITY OFFICE PERSONNEL SYSTEM PURPOSE AND GENERAL PROVISION § 2.2 References. (a) Subchapters III and IV of Chapter 7 of Title 31 U.S.C. (b) Title 5, United States Code....

  2. 4 CFR 2.2 - References.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 4 Accounts 1 2014-01-01 2013-01-01 true References. 2.2 Section 2.2 Accounts GOVERNMENT ACCOUNTABILITY OFFICE PERSONNEL SYSTEM PURPOSE AND GENERAL PROVISION § 2.2 References. (a) Subchapters III and IV of Chapter 7 of Title 31 U.S.C. (b) Title 5, United States Code....

  3. 4 CFR 2.2 - References.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 4 Accounts 1 2013-01-01 2013-01-01 false References. 2.2 Section 2.2 Accounts GOVERNMENT ACCOUNTABILITY OFFICE PERSONNEL SYSTEM PURPOSE AND GENERAL PROVISION § 2.2 References. (a) Subchapters III and IV of Chapter 7 of Title 31 U.S.C. (b) Title 5, United States Code....

  4. RASD's New Interlibrary Reference Request Form.

    ERIC Educational Resources Information Center

    American Libraries, 1985

    1985-01-01

    Introduces information request form developed by Cooperative Reference Services Committee of American Library Association's Reference and Adult Services Division to be used whenever a library cannot supply information requested by a patron and would like to ask another library for assistance. Front of proposed form with sample question is shown.…

  5. Uncertainty in Reference and Information Service

    ERIC Educational Resources Information Center

    VanScoy, Amy

    2015-01-01

    Introduction: Uncertainty is understood as an important component of the information seeking process, but it has not been explored as a component of reference and information service. Method: Interpretative phenomenological analysis was used to examine the practitioner perspective of reference and information service for eight academic research…

  6. 40 CFR 1066.710 - Reference materials.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... Practice for Measuring Fuel Economy and Emissions of Hybrid-Electric and Conventional Heavy-Duty Vehicles, Issued September 2002, IBR approved for § 1066.501. (c) National Institute of Standards and Technology... VEHICLE-TESTING PROCEDURES Definitions and Other Reference Material § 1066.710 Reference materials....

  7. Economics: A Guide to Reference Sources.

    ERIC Educational Resources Information Center

    Mason, Mary, Comp.

    Approximately 84 reference materials on economics located in the McLennan Library, McGill University (Montreal), are cited in this annotated bibliography. The bibliography serves to provide an overview of the printed bibliographic and reference sources useful for the study of economics. Financial and business sources and statistical compendia and…

  8. Reference. Advisory List of Instructional Media.

    ERIC Educational Resources Information Center

    North Carolina State Dept. of Public Education, Raleigh.

    The reference books featured in this annotated bibliography were selected from those titles that publishers submitted to the North Carolina Department of Public Instruction for review. As such, it is not a comprehensive list of all reference titles in print. This guide organizes the 33 titles into major subject categories: (1) Arts…

  9. Reference Services Planning in the 90s.

    ERIC Educational Resources Information Center

    Eckwright, Gail Z., Ed.; Keenan, Lori M., Ed.

    The focus of this collection of papers about library reference service is on the community outside the library, rather than the special populations served within it. "Conflicts in Value Systems" (Allen B. Veaner) is an overview of the major conflict areas facing the library profession today. "Reference Services for Off-Campus…

  10. Invoking Digital Reference: Creation and Implementation

    ERIC Educational Resources Information Center

    Johnson, Wendell; Burke, Adam

    2003-01-01

    In response to the changing educational landscape of the community college, reference librarians must adapt their service to fit the information needs of their patrons. One such update is synchronous digital reference, which applies instant messaging technology to library service. Todd Library at Waubonsee Community College (Sugar Grove, Illinois)…

  11. 32 CFR 1290.1 - References.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 6 2010-07-01 2010-07-01 false References. 1290.1 Section 1290.1 National Defense Other Regulations Relating to National Defense DEFENSE LOGISTICS AGENCY MISCELLANEOUS PREPARING AND PROCESSING MINOR OFFENSES AND VIOLATION NOTICES REFERRED TO U.S. DISTRICT COURTS §...

  12. 4 CFR 2.2 - References.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 4 Accounts 1 2011-01-01 2011-01-01 false References. 2.2 Section 2.2 Accounts GOVERNMENT ACCOUNTABILITY OFFICE PERSONNEL SYSTEM PURPOSE AND GENERAL PROVISION § 2.2 References. (a) Subchapters III and IV of Chapter 7 of Title 31 U.S.C. (b) Title 5, United States Code....

  13. 4 CFR 2.2 - References.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 4 Accounts 1 2010-01-01 2010-01-01 false References. 2.2 Section 2.2 Accounts GOVERNMENT ACCOUNTABILITY OFFICE PERSONNEL SYSTEM PURPOSE AND GENERAL PROVISION § 2.2 References. (a) Subchapters III and IV of Chapter 7 of Title 31 U.S.C. (b) Title 5, United States Code....

  14. So You Want To Do Virtual Reference?

    ERIC Educational Resources Information Center

    Coffman, Steve

    2001-01-01

    Examines the practical details of setting up live online reference services in libraries. Topics include choosing software; differences between licensed and hosted applications; models used to implement virtual reference services; how libraries are using these services in day to day operations; and marketing strategies. (Author/LRW)

  15. Going Prime Time with Live Chat Reference.

    ERIC Educational Resources Information Center

    Hoag, Tara J.; Cichanowicz, Edana McCaffrey

    2001-01-01

    Describes the development of the Suffolk Cooperative Library System's live, online chat reference service, a pilot project for public libraries in Suffolk County (New York). Topics include chat software selection; a virtual reference collection; marketing; funding; staffing; evaluation; expanded hours of service; email; and extracting data from…

  16. Ensuring Quality in a Virtual Reference Environment

    ERIC Educational Resources Information Center

    Barbier, Pat; Ward, Joyce

    2004-01-01

    Soon after AskALibrarian, Florida's Statewide Virtual Reference Desk, began to offer Chat Reference to the public in 2003, a Quality Assurance Workgroup was established to ensure that the service patrons received would be friendly, accurate, and adequate. To make certain that best practices were used in answering the real time questions, two…

  17. Reference Service in the Information Network.

    ERIC Educational Resources Information Center

    Bunge, Charles A.

    Reference service is defined as the mediation by a librarian between the need structures of users and the structures of information resources. The general process by which reference librarians accomplish this is outlined as including the phases of question clarification, question translation, search strategy formulation, search execution, delivery…

  18. Responsive Reference Service: Breaking Down Age Barriers.

    ERIC Educational Resources Information Center

    Bunge, Charles A.

    1994-01-01

    Examines possible barriers to adequate reference services to children and young adults. Topics discussed include the effects of budget pressures and shifting priorities in public and academic libraries; equity in access to information; reference policies; communication between children and adult services staff and administration; multiple service…

  19. Search Analysts as Successful Reference Librarians.

    ERIC Educational Resources Information Center

    Hammer, Mary M.

    1981-01-01

    Examines five factors critical to the successful functioning of search analysts as reference librarians: professional training, background knowledge and working experience in references services, personality characteristics and interpersonal skills, the marketing of online services, and patron evaluations supplemented by postsearch reference…

  20. 21 CFR 660.52 - Reference preparations.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 7 2012-04-01 2012-04-01 false Reference preparations. 660.52 Section 660.52 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) BIOLOGICS... preparations. Reference Anti-Human Globulin preparations shall be obtained from the Center for...

  1. 21 CFR 660.52 - Reference preparations.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 7 2011-04-01 2010-04-01 true Reference preparations. 660.52 Section 660.52 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) BIOLOGICS... preparations. Reference Anti-Human Globulin preparations shall be obtained from the Center for...

  2. 21 CFR 660.52 - Reference preparations.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 7 2014-04-01 2014-04-01 false Reference preparations. 660.52 Section 660.52 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) BIOLOGICS... preparations. Reference Anti-Human Globulin preparations shall be obtained from the Center for...

  3. 21 CFR 660.52 - Reference preparations.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 7 2010-04-01 2010-04-01 false Reference preparations. 660.52 Section 660.52 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) BIOLOGICS... preparations. Reference Anti-Human Globulin preparations shall be obtained from the Center for...

  4. 21 CFR 660.52 - Reference preparations.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 7 2013-04-01 2013-04-01 false Reference preparations. 660.52 Section 660.52 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) BIOLOGICS... preparations. Reference Anti-Human Globulin preparations shall be obtained from the Center for...

  5. 32 CFR 518.2 - References.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 3 2012-07-01 2009-07-01 true References. 518.2 Section 518.2 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS THE FREEDOM OF INFORMATION ACT PROGRAM General Provisions § 518.2 References. Required and...

  6. 32 CFR 518.2 - References.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 3 2011-07-01 2009-07-01 true References. 518.2 Section 518.2 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS THE FREEDOM OF INFORMATION ACT PROGRAM General Provisions § 518.2 References. Required and...

  7. 32 CFR 518.2 - References.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true References. 518.2 Section 518.2 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS THE FREEDOM OF INFORMATION ACT PROGRAM General Provisions § 518.2 References. Required and...

  8. 33 CFR 214.3 - Reference.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 3 2010-07-01 2010-07-01 false Reference. 214.3 Section 214.3 Navigation and Navigable Waters CORPS OF ENGINEERS, DEPARTMENT OF THE ARMY, DEPARTMENT OF DEFENSE EMERGENCY SUPPLIES OF DRINKING WATER § 214.3 Reference. (a) Pub. L. 84-99, as amended (33 U.S.C. 701n). (b) Pub....

  9. Information Entrepreneurship: Sources for Reference Librarians.

    ERIC Educational Resources Information Center

    Gilton, Donna L.

    1992-01-01

    Discusses information entrepreneurs, or information brokers, and information intrapreneurship, or the establishment of fee-based information services within a library. Annotations are provided for 56 information sources that include general materials, research in these areas, and relationships with traditional reference services. (four references)…

  10. 47 CFR 15.11 - Cross reference.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 47 Telecommunication 1 2011-10-01 2011-10-01 false Cross reference. 15.11 Section 15.11 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL RADIO FREQUENCY DEVICES General § 15.11 Cross reference. The provisions of subparts A, H, I, J and K of part 2 apply to intentional and unintentional...

  11. 47 CFR 15.11 - Cross reference.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 47 Telecommunication 1 2010-10-01 2010-10-01 false Cross reference. 15.11 Section 15.11 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL RADIO FREQUENCY DEVICES General § 15.11 Cross reference. The provisions of subparts A, H, I, J and K of part 2 apply to intentional and unintentional...

  12. 47 CFR 15.11 - Cross reference.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 47 Telecommunication 1 2014-10-01 2014-10-01 false Cross reference. 15.11 Section 15.11 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL RADIO FREQUENCY DEVICES General § 15.11 Cross reference. The provisions of subparts A, H, I, J and K of part 2 apply to intentional and unintentional...

  13. 47 CFR 15.11 - Cross reference.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 47 Telecommunication 1 2013-10-01 2013-10-01 false Cross reference. 15.11 Section 15.11 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL RADIO FREQUENCY DEVICES General § 15.11 Cross reference. The provisions of subparts A, H, I, J and K of part 2 apply to intentional and unintentional...

  14. 47 CFR 15.11 - Cross reference.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 47 Telecommunication 1 2012-10-01 2012-10-01 false Cross reference. 15.11 Section 15.11 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL RADIO FREQUENCY DEVICES General § 15.11 Cross reference. The provisions of subparts A, H, I, J and K of part 2 apply to intentional and unintentional...

  15. Writing references and using citation management software.

    PubMed

    Sungur, Mukadder Orhan; Seyhan, Tülay Özkan

    2013-09-01

    The correct citation of references is obligatory to gain scientific credibility, to honor the original ideas of previous authors and to avoid plagiarism. Currently, researchers can easily find, cite and store references using citation management software. In this review, two popular citation management software programs (EndNote and Mendeley) are summarized.

  16. Referent Salience Affects Second Language Article Use

    ERIC Educational Resources Information Center

    Trenkic, Danijela; Pongpairoj, Nattama

    2013-01-01

    The effect of referent salience on second language (L2) article production in real time was explored. Thai (-articles) and French (+articles) learners of English described dynamic events involving two referents, one visually cued to be more salient at the point of utterance formulation. Definiteness marking was made communicatively redundant with…

  17. 33 CFR 236.3 - References.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 3 2010-07-01 2010-07-01 false References. 236.3 Section 236.3 Navigation and Navigable Waters CORPS OF ENGINEERS, DEPARTMENT OF THE ARMY, DEPARTMENT OF DEFENSE WATER... QUALITY § 236.3 References. (a) PL 89-72 (b) ER 1105-2-10 (c) ER 1105-2-200...

  18. 21 CFR 660.3 - Reference panel.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... ADDITIONAL STANDARDS FOR DIAGNOSTIC SUBSTANCES FOR LABORATORY TESTS Antibody to Hepatitis B Surface Antigen § 660.3 Reference panel. A Reference Hepatitis B Surface Antigen Panel shall be obtained from the Center... shall be used for determining the potency and specificity of Antibody to Hepatitis B Surface Antigen....

  19. 21 CFR 660.3 - Reference panel.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... STANDARDS FOR DIAGNOSTIC SUBSTANCES FOR LABORATORY TESTS Antibody to Hepatitis B Surface Antigen § 660.3 Reference panel. A Reference Hepatitis B Surface Antigen Panel shall be obtained from the Center for... used for determining the potency and specificity of Antibody to Hepatitis B Surface Antigen....

  20. 21 CFR 660.3 - Reference panel.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... ADDITIONAL STANDARDS FOR DIAGNOSTIC SUBSTANCES FOR LABORATORY TESTS Antibody to Hepatitis B Surface Antigen § 660.3 Reference panel. A Reference Hepatitis B Surface Antigen Panel shall be obtained from the Center... shall be used for determining the potency and specificity of Antibody to Hepatitis B Surface Antigen....