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Sample records for pwr otimizacao da

  1. PWR-GALE. PWR Effluent Radioactivity Releases

    SciTech Connect

    Willis, C.A.

    1992-01-13

    PWR-GALE calculates the expected annual releases of radioactive materials in gaseous and liquid effluents from pressurized light water reactors (PWRs). The calculations are based on data generated from operating reactors, field and laboratory tests, and plant-specific considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment during normal operation including anticipated operational occurrences. PWR-GALE consists of two program, PGALEGS and PGALELQ. PGALEGS calculates the releases of radioactive materials (noble gases, radioactive particulates, carbon-14, tritium, argon-41, and iodine) in gaseous effluents from the waste gas processing system, steam generator blowdown system, condenser air ejector exhaust, containment purge exhaust, ventilation exhaust air from the auxiliary and turbine buildings and the spent fuel area, and steam leakage from the secondary system. PGALELQ calculates the releases of radioactive materials in liquid effluents from processed water generated from the boron recovery system to maintain plant water balance or for tritium control; processed liquid waste discharged from the waste systems, steam generator blowdown treatment system, and that discharged from the chemical waste and condensate demineralizer regeneration system; liquid waste discharged from the turbine building floor drain sumps; and detergent waste.

  2. WRAP-PWR verification studies

    SciTech Connect

    Gregory, M V; Ames, P L; Beranek, F; Kuehn, N H; Parks, P B

    1980-01-01

    A modular computational system known as the Water Reactor Analysis Package - Evaluation Model (WRAP-EM) was developed for the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor EM methods and computed results. A subset of the system (WRAP-PWR-EM) provides the computational tools to perform a complete analysis of loss-of-coolant accidents (LOCA's) in pressurized water reactors (PWR's). A set of calculations modeling experimental tests in the Semiscale and LOFT facilities, and calculations of a large break in a typical four-loop Westinghouse PWR plant have verified that the WRAP-PWR-EM system is functioning as intended.

  3. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

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

  4. 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. Analytical description of PWR pressurizer transients. Final report

    SciTech Connect

    Ahl, J.P.

    1985-03-01

    Simulating the complicated physical processes that occur in a PWR pressurizer during a transient presented a considerable challenge to modelers. The computer code developed and validated in this study will help utilities to better understand both the behavior of the pressurizer and the overall performance of a PWR after a loss-of-coolant accident.

  6. 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.

  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. 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.

  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. 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.

  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. 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)

  14. 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.

  15. 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.

  16. PWR plant transient analyses using TRAC-PF1

    SciTech Connect

    Ireland, J.R.; Boyack, B.E.

    1984-01-01

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public.

  17. 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.

  18. Report on the PWR-radiation protection/ALARA Committee

    SciTech Connect

    Malone, D.J.

    1995-03-01

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance.

  19. 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.

  20. 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.

  1. 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.

  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. 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.

  4. 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

  5. Comparison of PWR-IMF and FR fuel cycles

    SciTech Connect

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj |; Necas, Vladimir

    2007-07-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  6. Probabilistic fracture mechanics code for PWR steam generator tube maintenance

    SciTech Connect

    Granger, B. ); Pitner, P. ); Flesch, B. )

    1991-01-01

    This paper presents the COMPROMIS code developed by Electricite de France (EDF) to optimize the maintenance of PWR steam generator (SG) tube bundles. This model, based on probabilistic fracture mechanics, quantifies the impact of in-service inspections and maintenance actions on the risk of failure of an SG tube, with allowance as random variable for all the relevant parameters (distribution of crack sizes, detection and sizing capability, crack initiation and propagation, critical sizes, leak before break risk). The code is SG-specific and is designed to allow realtime evaluation based on manufacturing and inspection data banks.

  7. 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.

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

    SciTech Connect

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

    2006-07-01

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

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

    SciTech Connect

    King, D.B.

    1989-07-01

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

  10. 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.

  11. 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.

  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. 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.

  16. Tritium target performance during an LBLOCA in a PWR

    SciTech Connect

    Reid, B.D.

    1996-12-31

    In December 1995, the U.S. Department of Energy (DOE) announced a preferred strategy for acquiring a new supply of tritium. That strategy is based on pursuing the two most promising production alternatives. These alternatives include either constructing an accelerator-produced tritium system for tritium production or procuring an existing commercial light water reactor or irradiation services from such a reactor to irradiate tritium targets. This paper discusses the safety performance of a tritium target in a commercial pressurized water reactor (PWR). The current conceptual design for the light water tritium targets is quite similar, in terms of external dimensions and materials, to early designs for stainless steel clad discrete burnable absorbers used in PWRs. The tritium targets nominally consist of an annular lithium aluminate pellet wrapped in a Zircaloy-4 getter and clad with Type 316 stainless steel.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  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. 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.

  3. 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.

  4. 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.

  5. 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.

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

    SciTech Connect

    M.Krug, R.Shogan

    2004-09-01

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

  7. 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.

  8. 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

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

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

    1982-01-01

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

  14. The determination of critical nuclides in PWR waste streams

    SciTech Connect

    De Goeyse, A.

    1993-12-31

    The safety studies concerning the final disposal of low- and intermediate-level radioactive waste take into consideration a series of long-lived radionuclides. The problem the producers have to cope with comes from the fact that those nuclides, which are mainly (pure) {beta} emitters or {alpha} emitters, cannot be measured by a direct current method such as gamma scanning. Their determination involves sophisticated radiochemical techniques which are difficult to implement by a producer on a routine basis for normal production waste. A current method for the determination of those nuclides in the waste streams produced by a nuclear power reactor consists in applying correlation factors or scaling factors between those critical nuclides and so called key radionuclides, which can be easily measured and are representative for the occurrence of activation products and fission products in the waste streams. In order to identify and define those correlation factors, ONDRAF/NIRAS, has subcontracted, in agreement with the waste producer (ELECTRABEL), a complete study to the engineering company BELGATOM (BA) for the different waste streams produced by the seven Belgian PWR plants.

  15. PWR (pressurized water reactor) pressurizer transient response: Final report

    SciTech Connect

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model.

  16. 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.

  17. 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.

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

    SciTech Connect

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

    2010-06-22

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

  19. 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.

  20. 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.

  1. 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)

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

    SciTech Connect

    Kimura, C.Y.

    1984-09-01

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

  3. 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

  4. 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)

  5. 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.

  6. 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.

  7. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    NASA Astrophysics Data System (ADS)

    Debarberis, L.; Acosta, B.; Zeman, A.; Sevini, F.; Ballesteros, A.; Kryukov, A.; Gillemot, F.; Brumovsky, M.

    2006-04-01

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  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. 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)

  12. Some thermalhydraulics of closure head adapters in a 3 loops PWR

    SciTech Connect

    Hoffmann, F.; Daubert, O.; Hecker, M.

    1995-09-01

    In 1993 a R&D action, based on numerical simulations and experiments on PWR`s upper head was initiated. This paper presents the test facility TRAVERSIN (a scale model of a 900 MW PWR adapter) and the calculations performed on the geometry of different upper head sections with the Thermalhydraulic Finite Element Code N3S used for 2D and 3D computations. The paper presents the method followed to bring the adapter and upper head study to a successful conclusion. Two complementary approaches are performed to obtain global results on complete fluid flow in the upper head and local results on the flow around the adapters of closure head. A validation test case of these experimental and numerical tools is also presented.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

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

    SciTech Connect

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

    1981-05-01

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

  1. Primary water chemistry improvement for radiation exposure reduction at Japanese PWR Plants

    SciTech Connect

    Nishizawa, Eiichi

    1995-03-01

    Radiation exposure during the refueling outages at Japanese Pressurized Water Reactor (PWR) Plants has been gradually decreased through continuous efforts keeping the radiation dose rates at relatively low level. The improvement of primary water chemistry in respect to reduction of the radiation sources appears as one of the most important contributions to the achieved results and can be classified by the plant operation conditions as follows

  2. 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.

  3. 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.

  4. 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.

  5. 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)

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

    SciTech Connect

    Naegeli, Robert Earl

    2004-06-01

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

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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)

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

    SciTech Connect

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

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

  13. 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.

  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. 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.

  16. Development of emergency operator support system for next Japanese PWR plants

    SciTech Connect

    Ito, K.; Hanada, S.; Yoshida, Y.; Sugino, K.

    2006-07-01

    The purpose of main control room improvement is to reduce operator workload and potential human errors by offering a better working environment where operators can maximize their abilities. Japanese PWR utilities and Mitsubishi have developed an operator support system entitled Emergency Operator Support System (EOSS). The system supports operators in incidental/accidental situations which may be worsened by human errors. In order to confirm the validity of the system, a proto type was built, and was evaluated by operator crews. The consequence showed good result of effectiveness in avoiding potential human errors and decreasing workload of operators. (authors)

  17. 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.

  18. 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.

  19. 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).

  20. Analysis of the return to power scenario following a LBLOCA in a PWR

    SciTech Connect

    Macian, R.; Tyler, T.N.; Mahaffy, J.H.

    1995-09-01

    The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

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

    SciTech Connect

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

    2013-07-01

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

  2. Cartesian Meshing Impacts for PWR Assemblies in Multigroup Monte Carlo and Sn Transport

    NASA Astrophysics Data System (ADS)

    Manalo, K.; Chin, M.; Sjoden, G.

    2014-06-01

    Hybrid methods of neutron transport have increased greatly in use, for example, in applications of using both Monte Carlo and deterministic transport to calculate quantities of interest, such as flux and eigenvalue in a nuclear reactor. Many 3D parallel Sn codes apply a Cartesian mesh, and thus for nuclear reactors the representation of curved fuels (cylinder, sphere, etc.) are impacted in the representation of proper fuel inventory (both in deviation of mass and exact geometry representation). For a PWR assembly eigenvalue problem, we explore the errors associated with this Cartesian discrete mesh representation, and perform an analysis to calculate a slope parameter that relates the pcm to the percent areal/volumetric deviation (areal corresponds to 2D and volumetric to 3D, respectively). Our initial analysis demonstrates a linear relationship between pcm change and areal/volumetric deviation using Multigroup MCNP on a PWR assembly compared to a reference exact combinatorial MCNP geometry calculation. For the same multigroup problems, we also intend to characterize this linear relationship in discrete ordinates (3D PENTRAN) and discuss issues related to transport cross-comparison. In addition, we discuss auto-conversion techniques with our 3D Cartesian mesh generation tools to allow for full generation of MCNP5 inputs (Cartesian mesh and Multigroup XS) from a basis PENTRAN Sn model.

  3. 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.

  4. 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.

  5. Study of a transient identification system using a neural network for a PWR plant

    SciTech Connect

    Ishihara, Yoshinao; Kasai, Masao; Kambara, Masayuki; Mitsuda, Hiromichi; Kurata, Toshikazu; Shirosaki, Hidekazu

    1996-08-01

    This paper presents the procedure and results of a system for identifying PWR plant abnormal events, which uses neural network techniques. The neural network recognizes the abnormal event from the patterns of the transient changes of analog data from plant parameters when they deport from their normal state. For the identification of abnormal events in this study, events that cause a reactor to scram during power operation were selected as the design base events. The test data were prepared by simulating the transients on a compact PWR simulator. The simulation data were analyzed to determine how the plant parameters respond after the occurrence of a transient. A method of converting the pattern of the transient changes into characteristic parameters by fitting the data to pre-determined functions was developed. These characteristic parameters were used as the input data to the neural network. The neural network learning procedure used a generalized delta rule, namely a back-propagation algorithm. The neural network can identify the type of an abnormal event from a limited set of events by using these characteristic parameters obtained from the pattern of the changes in the analog data. From the results of this application of a neural network, it was concluded that it would be possible to use the method to identify abnormal events in a nuclear power plant.

  6. 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.

  7. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    SciTech Connect

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

    SciTech Connect

    Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas

    2015-11-01

    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

  14. 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.

  15. CORMLT code for the analysis of degraded core accidents. Computer code manual. [PWR

    SciTech Connect

    Denny, V.E.

    1984-12-01

    A computer code (CORMLT) has been developed to predict the effects of bouyancy-driven convection on the progression of core-degrading accidents in PWR vessels. Thermal/hydraulics modeling includes the downcomer/bottom-head regions, as well as the upper vessel and adjacent hot-leg portions of the primary coolant system for which gas communication is limited to the intervening discharge nozzles (so-called dead-end volumes). CORMLT requires flow rates and temperatures of any water feed (to the downcomer) versus time. CORMLT provides composition, enthalpy, temperature, and flow rate of steam/hydrogen mixtures within the vessel above the (receding) water surface, as well as estimates of these quantities for interaction between the plenum and the rest of the PCS. CORMLT also provides graphical representations for the morphological behavior of the progression of core meltdown accidents.

  16. 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.

  17. 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)

  18. 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.

  19. 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.

  20. Improvement of PWR liquid radwaste system in Korean Nuclear Power Plants

    SciTech Connect

    Lee, Byung-Sik; Kim, Kil-Jung; Ko, Dae-Hack

    1995-11-01

    Currently in Korea, there are 12 Pressurized Water Reactors (PWR) either operating or under construction. These units encompass several different designs for liquid radwaste systems. These different designs, however, may be categorized into three (3) basic groups based upon waste processing technologies. This paper describes the design concepts and operating experiences for the each. waste processing group. Based upon design and operating experiences, we implement to improve liquid radwaste system by simplification (i.e. elimination of unnecessary equipment) and employing current technologies. These improvements are applied to Yonggwang Unit 5&6, which is now in the basic design stage. This paper also describes some unique features of upgrading the liquid radwaste systems.

  1. 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.

  2. 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.

  3. 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

  4. Results from semiscale MOD-2A upper head injection test series. [PWR

    SciTech Connect

    Shimeck, D.J.; Leonard, M.T.

    1981-01-01

    A series of small break loss-of-coolant (SBLOCA) experiments and associated RELAP5/MOD computer code calculations have been performed by the Semiscale Program at the Idaho National Engineering Laboratory (INEL) to investigate the influence of upper head injection (UHI) on transient response. A UHI system, as designed for pressurized water reactors (PWR's), has an 8.7-MPa accumulator that injects emergency core coolant (ECC) directly into the upper head of the reactor vessel, and loop accumulators nominally pressurized to 2.86 MPa (as opposed to 4.14 MPa in a standard design). Since this configuration was optimized based upon large break LOCA calculations the experiments were requested by the US Nuclear Regulatory Commission (USNRC) to assist in evaluating system performance for SBLOCA's.

  5. 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.

  6. 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)

  7. 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.

  8. Three Dimensional Radiation Transport Analyses in Pwr with Tort and Mcnp

    NASA Astrophysics Data System (ADS)

    Fukuya, Koji; Nakata, Hayato; Kimura, Itsuro; Kitagawa, Hideo; Ohmura, Masaki; Ito, Taku; Shin, Kazuo

    2003-06-01

    Three dimensional (3D) neutron and gamma calculations for structural materials inside the reactor vessel in a commercial PWR were performed using the 3D transport code TORT and the Monte Carlo code MCNP to assess the accuracy of calculations using these codes and libraries. Comparisons with two dimensional DORT calculations with various libraries and surveillance dosimetry measurements indicated that TORT and MCNP calculations give similar agreements with surveillance measurements to DORT calculations. Influences of the cross section data, ENDF/B-IV, ENDF/B-VI and JENDL3.2 on attenuation of the fast flux and dpa rate in the reactor vessel, relative contributions of gamma-rays and thermal neutrons to dpa were discussed.

  9. 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%.

  10. 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.

  11. 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.

  12. Analysis of the NEACRP PWR rod ejection benchmark problems with DIF3D-K

    SciTech Connect

    Kim, M.H.; Taiwo, T.A.; Khalil, H.S.

    1994-03-01

    Analyses of the NEACRP PWR rod ejection transient benchmark problems with the DIF3D-K nodal kinetics code are presented. The DIF3D-K results are shown to be in generally good agreement with results obtained using other codes, in particular reference results previously generated with the PANTHER code. The sensitivity of the transient results to the DIF3D-K input parameters (such as time step size, radial and axial node sizes, and the mesh structure employed for fuel pin heat conduction calculation) are evaluated and discussed. In addition, the potential in reducing computational effort by application of the improved quasistatic scheme (IQS) to these rod ejection transients, which involve very significant flux shape changes and thermal-hydraulic feedback is evaluated.

  13. Assessment of non-backfittable concepts to improve PWR uranium utilization

    SciTech Connect

    LaBelle, D.W.; Sankovich, M.F.; Spetz, S.W.; Uotinen, V.O

    1980-12-01

    Seven non-backfittable improvements to light water reactors were assessed for Batelle/Pacific Northwest Laboratories in support of the Department of Energy's program on Advanced Reactor Studies. The objective was to provide industrial perspective as to which concepts have the best potential for development to improve fuel utilization. The concepts were rated against the assessment criteria while considering the key questions identified for each concept, and recommendations were made for further action on unresolved key questions. The concepts were subjectively ranked against each other in terms of relative investment potential. The ranking considered all criteria but, for example, weighted fuel utilization savings more heavily than development costs. Finally, conclusions and recommendations for future action were determined. The reference design for this study was the NASAP Composite Improved PWR.

  14. 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.

  15. PWR design for low doses in the United Kingdom: The present and the future

    SciTech Connect

    Zodiates, A.M.; Willcock, A.

    1995-03-01

    The Pressurizer Water Reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant System (SNUPPS). This design was developed to meet the United Kingdom (UK) requirements and those improvements are embodied in the Sizewell B plant. Nuclear Electric plc is now looking to the design of the future PWRs to be built in the UK. These PWRs will be based as replicas of the Sizewell B design, but attention will be given to reducing operator doses further. This paper details the approach in operator protection improvements incorporated at Sizewall B, presents the estimated annual collective dose, and identifies the approach being adopted to reduce further operator doses in future plants.

  16. Failure probability of PWR reactor coolant loop piping. [Double-ended guillotine break

    SciTech Connect

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria.

  17. 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.

  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. 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

  20. 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)

  1. 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)

  2. 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.

  3. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    SciTech Connect

    DeHart, M.D.

    1993-01-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

  4. 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.

  5. 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).

  6. 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.

  7. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR (pressurized-water-reactor) plants

    SciTech Connect

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

    1988-01-01

    Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs.

  8. 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.

  9. 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.

  10. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    SciTech Connect

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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)

  17. 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.

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

    SciTech Connect

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

    1995-03-01

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

  19. Feasibility of recycling thorium in a fusion-fission hybrid/PWR symbiotic system

    SciTech Connect

    Josephs, J.M.

    1980-12-31

    A study was made of the economic impact of high levels of radioactivity in the thorium fuel cycle. The sources of this radioactivity and means of calculating the radioactive levels at various stages in the fuel cycle are discussed and estimates of expected levels are given. The feasibility of various methods of recycling thorium is discussed. These methods include direct recycle, recycle after storage for 14 years to allow radioactivity to decrease, shortening irradiation times to limit radioactivity build up, and the use of the window in time immediately after reprocessing where radioactivity levels are diminished. An economic comparison is made for the first two methods together with the throwaway option where thorium is not recycled using a mass energy flow model developed for a CTHR (Commercial Tokamak Hybrid Reactor), a fusion fission hybrid reactor which serves as fuel producer for several PWR reactors. The storage option is found to be most favorable; however, even this option represents a significant economic impact due to radioactivity of 0.074 mills/kW-h which amounts to $4 x 10/sup 9/ over a 30 year period assuming a 200 gigawatt supply of electrical power.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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).

  7. 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.

  8. 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)

  9. 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)

  10. 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.

  11. 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)

  12. 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.

  13. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

    NASA Astrophysics Data System (ADS)

    Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

    2017-09-01

    The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  14. 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.

  15. 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.

  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. 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.

  18. Multicycle PWR in-core fuel management through one-and-a-half-dimensional core modeling

    SciTech Connect

    Petrovic, B.G.; Levine, S.H. )

    1992-01-01

    The one-and-a-half-dimensional (1 and 1/2-D) model of a pressurized water reactor (PWR) core employs one-dimensional (1-D) diffusion calculation followed by a fast, few-step procedure to unfold the 1-D results into the two-dimensional (2-D) results. A computer code was developed based on that model. The initial benchmarking has shown the code to be almost as fast as a plain 1-D code and significantly faster than the analogous 2-D code (10 to 100 times for a typical problem). Yet, it provides results in 2-D form and better accuracy than the 1-D code. The model itself and the initial benchmarking were described in more detail elsewhere. The code has since been enhanced, and a multicycle analysis option has been implemented. This paper presents results of benchmarking the model using the actual data for three successive cycles of the Krsko nuclear power plant and complex low-leakage loading patterns.

  19. Compatibility of PWR gasket and packing materials and resins with organic amines

    SciTech Connect

    Keneshea, F.J.; Hobart, S.A. ); Camenzind, M.J. )

    1992-07-01

    The objectives of this testing program were two-fold: (1) to examine the compatibility of morpholine and five other amines with several synthetic polymeric materials useful for gaskets and seals in pressurized water reactor (PWR) secondary cycles and (2) to examine the potential chemical degradation of ion exchange (IX) resins by morpholine and ethanolamine. The screening of the polymeric materials in the amines was performed by heating small samples of the materials in the amines for one week to one month. Interaction of the amines with the materials was accelerated by testing at elevated temperatures and at high amine concentrations. Two materials (Kalrez and EPDM) that are potentially useful in high-temperature and high-pressure steam systems were tested in morpholine solutions in sealed bombs at 260{degrees}C (500{degrees}). After heating in the aqueous amine solutions, changes in weight were measured and the samples were visually examined for physical changes, such as swelling or cracking. Selected materials underwent testing for hardness, elongation, and tensile strength after heating in morpholine for one month. This document provides the results of this testing program.

  20. Aging mechanisms in the Westinghouse PWR (Pressurized Water Reactor) Control Rod Drive system

    SciTech Connect

    Gunther, W.; Sullivan, K.

    1991-01-01

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs.

  1. 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.

  2. 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)

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. The Application of Modern Nodal Methods to Pwr Reactor Physics Analysis.

    NASA Astrophysics Data System (ADS)

    Knight, M. P.

    Available from UMI in association with The British Library. The objective of this research is to develop efficient computational procedures for PWR reactor calculations, based on modern nodal methods. The analytic nodal method, which is characterised by the use of exact exponential expansions in transverse-integrated equations, is implemented within an existing finite-difference code. This shows considerable accuracy and efficiency on standard benchmark problems, very much in line with existing experience with nodal methods. Assembly powers can be calculated to within 2.0% with just one mesh per assembly. The recovery of fine detail from a nodal solution based on such a coarse mesh requires additional effort. Techniques are develolped in this thesis which allow the basic nodal equations to be used in this reconstruction, and therefore provide a consistent approach. Pin powers can be recovered from assembly-averaged values with little further loss of accuracy. A similar investigation is followed with the transverse leakage distribution. An improvement, which uses known local behaviour, is shown to be very effective in some limited applications, but overall provides little advantage over the much simpler quadratic model. For heterogeneous calculations it is essential that the homogenisation techniques are well matched to the nodal method. The asymmetric design of some assemblies provides a severe test. Techniques are devised that allow some overall representation of this asymmetry to be retained in the reactor calculation, even when using one mesh per assembly. Extensions of this procedure provide an almost exact global representation of a heterogeneous assembly. A complete comparison is performed between reactor calculations at one mesh per pin, and at one mesh per assembly using nodal and homogenisation methods. Homogenisation errors and nodal coarse-mesh errors are shown to be very similar, amounting to about 0.1% on reactor eigenvalue, 2.0% on assembly power and

  9. 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)

  10. 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.

  11. 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

  12. 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

  13. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    NASA Astrophysics Data System (ADS)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  14. 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.

  15. 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.

  16. 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.

  17. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    SciTech Connect

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

  18. 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.

  19. Asymmetric blowdown loads on PWR (pressurized-water-reactor) primary systems: resolution of generic task action plan A-2

    SciTech Connect

    Hosford, S.B.; Mattu, R.; Meyer, R.O.; Throm, E.D.; Tinkler, C.G.

    1981-01-01

    NRC staff, after being informed of newly identified asymmetric loadings resulting from postulated ruptures of primary piping, initiated a generic investigation, Task Action Plan A-2, limited to pressurized-water-reactor (PWR) plants because of their higher primary system pressures. The intent of the investigation was to develop acceptable criteria and guidelines for evaluating plant analyses. The staff concludes that an acceptable basis is provided in this report for performing and reviewing plant analyses. Criteria were developed for evaluating loading transients, structural components, and the fuel assembly. The staff approved computer programs and modeling techniques submitted by each PWR vendor for development of the subcooled blowdown and cavity-pressure loading transients. Audit models were developed to evaluate the structural computer programs and modeling techniques. Methods have been approved for the structural-analysis method submitted by Westinghouse for the Indian Point Unit 3 plant. Criteria and guidelines are provided to perform a detailed evaluation of the fuel assembly. Acceptance criteria are also provided so deformed fuel-assembly spacer grids may be evaluated.

  20. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    NASA Astrophysics Data System (ADS)

    Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen; Zhou, Bangxin; Shoji, Tetsuo

    2016-04-01

    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T-L and L-T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T-L orientation with a higher crack growth rate than that in the specimen L-T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L-T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant.

  1. 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

  2. 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

  3. 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

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

    SciTech Connect

    Lee, Y. K.

    2006-07-01

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

  5. 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.

  6. 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.

  7. 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.

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

    SciTech Connect

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

    1982-10-01

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

  9. 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

  10. 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

  11. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    SciTech Connect

    Weber, P.; Umminger, K.J.; Schoen, B.

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  12. 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.

  13. Computational methods and implementation of the 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction

    SciTech Connect

    Aragones, J.M.; Ahnert, C.

    1995-12-31

    New computational methods have been developed in our 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction. They improve the accuracy and efficiency of the coupled neutronic-thermalhydraulic solution and extend its scope to provide, mainly, the calculation of: the fission reaction rates at the incore mini-detectors; the responses at the excore detectors (power range); the temperatures at the thermocouple locations; and the in-vessel distribution of the loop cold-leg inlet coolant conditions in the reflector and core channels, and to the hot-leg outlets per loop. The functional capabilities implemented in the extended SIMTRAN code for online utilization include: online surveillance, incore-excore calibration, evaluation of peak power factors and thermal margins, nominal update and cycle follow, prediction of maneuvers and diagnosis of fast transients and oscillations. The new code has been installed at the Vandellos-II PWR unit in Spain, since the startup of its cycle 7 in mid-June, 1994. The computational implementation has been performed on HP-700 workstations under the HP-UX Unix system, including the machine-man interfaces for online acquisition of measured data and interactive graphical utilization, in C and X11. The agreement of the simulated results with the measured data, during the startup tests and first months of actual operation, is well within the accuracy requirements. The performance and usefulness shown during the testing and demo phase, to be extended along this cycle, has proved that SIMTRAN and the man-machine graphic user interface have the qualities for a fast, accurate, user friendly, reliable, detailed and comprehensive online core surveillance and prediction.

  14. 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).

  15. 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).

  16. 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.

  17. 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

  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. 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)

  1. OBSERVATIONS AND IMPLICATIONS OF INTERGRANULAR STRESS CORROSION CRACK GROWTH OF ALLOY 152 WELD METALS IN SIMULATED PWR PRIMARY WATER

    SciTech Connect

    Toloczko, Mychailo B.; Olszta, Matthew J.; Overman, Nicole R.; Bruemmer, Stephen M.

    2013-08-15

    Significant intergranular (IG) crack growth during stress corrosion cracking (SCC) tests has been documented during tests in simulated PWR primary water on two alloy 152 specimens cut from a weldment produced by ANL. The cracking morphology was observed to change from transgranular (TG) to mixed mode (up to ~60% IG) during gentle cycling and cycle + hold loading conditions. Measured crack growth rates under these conditions often suggested a moderate degree of environmental enhancement consistent with faster growth on grain boundaries. However, overall SCC propagation rates at constant stress intensity (K) or constant load were very low in all cases. Initial SCC rates up to 6x10-9 mm/s were occasionally measured, but constant K/load growth rates dropped below ~1x10-9 mm/s with time even when significant IG engagement existed. Direct comparisons were made among loading conditions, measured crack growth response and cracking morphology during each test to assess IGSCC susceptibility of the alloy 152 specimens. These results were analyzed with respect to our previous SCC crack growth rate measurements on alloy 152/52 welds.

  2. 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.

  3. 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.

  4. Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

    NASA Astrophysics Data System (ADS)

    Chen, Junjie; Xiao, Qian; Lu, Zhanpeng; Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan

    2017-06-01

    The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters.

  5. 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.

  6. Ligand activation induces different conformational changes in CXCR3 receptor isoforms as evidenced by plasmon waveguide resonance (PWR).

    PubMed

    Boyé, K; Billottet, C; Pujol, N; Alves, I D; Bikfalvi, A

    2017-09-06

    The chemokine receptor CXCR3 plays important roles in angiogenesis, inflammation and cancer. Activation studies and biological functions of CXCR3 are complex due to the presence of spliced isoforms. CXCR3-A is known as a pro-tumor receptor whereas CXCR3-B exhibits anti-tumor properties. Here, we focused on the conformational change of CXCR3-A and CXCR3-B after agonist or antagonist binding using Plasmon Waveguide Resonance (PWR). Agonist stimulation induced an anisotropic response with very distinct conformational changes for the two isoforms. The CXCR3 agonist bound CXCR3-A with higher affinity than CXCR3-B. Using various concentrations of SCH546738, a CXCR3 specific inhibitor, we demonstrated that low SCH546738 concentrations (≤1 nM) efficiently inhibited CXCR3-A but not CXCR3-B's conformational change and activation. This was confirmed by both, biophysical and biological methods. Taken together, our study demonstrates differences in the behavior of CXCR3-A and CXCR3-B upon ligand activation and antagonist inhibition which may be of relevance for further studies aimed at specifically inhibiting the CXCR3A isoform.

  7. Presentation of the MERC work-flow for the computation of a 2D radial reflector in a PWR

    SciTech Connect

    Clerc, T.; Hebert, A.; Leroyer, H.; Argaud, J. P.; Poncot, A.; Bouriquet, B.

    2013-07-01

    This paper presents a work-flow for computing an equivalent 2D radial reflector in a pressurized water reactor (PWR) core, in adequacy with a reference power distribution, computed with the method of characteristics (MOC) of the lattice code APOLLO2. The Multi-modelling Equivalent Reflector Computation (MERC) work-flow is a coherent association of the lattice code APOLLO2 and the core code COCAGNE, structured around the ADAO (Assimilation de Donnees et Aide a l'Optimisation) module of the SALOME platform, based on the data assimilation theory. This study leads to the computation of equivalent few-groups reflectors, that can be spatially heterogeneous, which have been compared to those obtained with the OPTEX similar methodology developed with the core code DONJON, as a first validation step. Subsequently, the MERC work-flow is used to compute the most accurate reflector in consistency with all the R and D choices made at Electricite de France (EDF) for the core modelling, in terms of number of energy groups and simplified transport solvers. We observe important reductions of the power discrepancies distribution over the core when using equivalent reflectors obtained with the MERC work-flow. (authors)

  8. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    NASA Astrophysics Data System (ADS)

    LaFleur, Adrienne M.; Menlove, Howard O.

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies.

  9. Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

    NASA Astrophysics Data System (ADS)

    Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

    2017-02-01

    Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

  10. 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)

  11. Management of thermal peaking factors in CONFU-B PWR assemblies using neutron poisons and tailored enrichment

    SciTech Connect

    Visosky, M.; Hejzlar, P.; Kazimi, M.

    2006-07-01

    CONFU-B assemblies are PWR assemblies containing standard Uranium fuel rods and TRU bearing inert material fuel rods and are designed to achieve net TRU destruction over a 4.5-year irradiation. These highly heterogeneous assemblies tend to exhibit large intra-assembly power peaking factors (IAPPF). Neutronic strategies to reduce IAPPF are developed. The IAPPF are calculated at the assembly level using CASMO4, and these are used to calculate the most restrictive thermal margin (the Minimum Departure from Nucleate Boiling Ratio, MDNBR) using a whole-core VIPRE-01 model. This paper examines two strategies to manage the thermal margin of a CONFU-B assembly while retaining the TRU destruction performance: use of neutron poisons and tailored enrichment schemes. Burnable poisons can be used to suppress BOL reactivity of fresh CONFU-B assemblies with only minor impact on MDNBR and TRU destruction performance. Tailored enrichment, along with the use of soluble boron, can achieve significant improvements in MDNBR, but at some cost to TRU destruction performance. (authors)

  12. Determination of Bandwidths of PWR-UO2 Spent Fuel Radionuclide Inventory Based on Real Operational History Data

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

    An important requirement 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 radionuclide (RN) activities and the associated 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, details of irradiation history are often missing, which complicates the assessment of declared RN inventories. Here, we present a set of burn-up calculations, in which the real operational histories of 339 published or anonymized PWR fuel assemblies (FA) are taken into account. These histories provide information about ranges of values of the associated secondary reactor parameters (SRP), which are useful for the “SRP analysis”. Hence, we can calculate realistic variations in the spectrum of RN inventories. SCALE 6.1 with the ENDF/B-VII.0 library has been employed for the burn-up calculations. The results have been validated using experimental measurements from the online database SFCOMPO.

  13. 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

  14. 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.

  15. 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.

  16. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    SciTech Connect

    Yackle, T R; MacDonald, P E; Broughton, J M

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  17. 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.

  18. 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

  19. 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.

  20. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    SciTech Connect

    Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; Sung, Yixing

    2016-09-07

    VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by a system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.

  1. 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)

  2. 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.

  3. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    DOE PAGES

    Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

    2016-09-07

    VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

  4. 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)

  5. 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

  6. 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

  7. 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.

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

    SciTech Connect

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

    1993-10-01

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

  9. 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.

  10. 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.

  11. 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.

  12. Spent fuel dry storage technology development: thermal evaluation of isolated drywells containing spent fuel (1 kW PWR spent fuel assembly)

    SciTech Connect

    Unterzuber, R; Wright, J B

    1980-09-01

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site. Two PWR spent fuel assemblies having a decay heat level of approximately 1.1 kW were encapsulated inside the E-MAD Hot Bay and placed in instrumented near-surface drywell storage cells. Temperatures from the two isolated drywells and the adjacent soil have been recorded throughout the 19 month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (254{sup 0}F and 203{sup 0}F, respectively) during August 1979. Thereafter, all temperatures decreased and showed a cycling pattern which responded to seasonal atmospheric temperature changes. A computer model was utilized to predict the thermal response of the drywell. Computer predictions of the drywell temperatures and the temperatures of the surrounding soil are presented and show good agreement with the test data.

  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. 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.

  15. 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)

  16. 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

  17. 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.

  18. 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)

  19. Iodine volatility. [PWR; BWR

    SciTech Connect

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

    1984-01-01

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

  20. 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.

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

    SciTech Connect

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

    2015-09-01

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

  2. ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3. 06. 6B - transient film boiling in upflow. [PWR

    SciTech Connect

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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)

  9. Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

    NASA Astrophysics Data System (ADS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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.

  16. 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.

  17. Neutronic Analysis of the Burning of Transuranics in Fully Ceramic Micro-Encapsulated Tri-Isotropic Particle-Fuel in a PWR

    SciTech Connect

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-11-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) – only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO2 and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO2 and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior is dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  18. 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

  19. 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.

  20. The da Vinci robot.

    PubMed

    Moran, Michael E

    2006-12-01

    One might assume from the title of this paper that the nuances of a complex mechanical robot will be discussed, and this would be correct. On the other hand, the date of the design and possible construction of this robot was 1495, a little more than five centuries ago. The key point in the title is the lack of a trademarked name, as Leonardo was the designer of this sophisticated system. His notes from the Codex Altanticus represent the foundation of this report. English translations of da Vinci's notebooks are currently available. Beginning in the 1950s, investigators at the University of California began to ponder the significance of some of da Vinci's markings on what appeared to be technical drawings. Such markings also occur in his Codex Atlanticus (the largest single collection of da Vinci's sheets, consisting of 1119 separate pages and 481 folios) along with a large number of other mechanical devices. Continuing research at the Instituto e Museo di Storia della Scienza in Florence has yielded a great deal of information about Leonardo's intentions with regard to his mechanical knight. It is now known that da Vinci's robot would have had the outer appearance of a Germanic knight. It had a complex core of mechanical devices that probably was human powered. The robot had two independent operating systems. The first had three degree-of-freedom legs, ankles, knees, and hips. The second had four degrees of freedom in the arms with articulated shoulders, elbows, wrists, and hands. A mechanical analog-programmable controller within the chest provided the power and control for the arms. The legs were powered by an external crank arrangement driving the cable, which connected to key locations near each lower extremity's joints. Da Vinci also is known to have devised a programmable front-wheel-drive automobile with rack-and-pinion suspension mechanisms at age 26. He would recall this device again, when, at age 40, he is thought to have built a programmable automated

  1. 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.

  2. Henrique da Rocha Lima.

    PubMed

    Bernardes Filho, Fred; Avelleira, João Carlos Regazzi

    2015-01-01

    Brazilian physician and researcher Henrique da Rocha Lima was born in 1879 in the city of Rio de Janeiro, where he studied medicine and obtained the degree of M.D. in 1901. He specialized in Clinical Medicine in Germany and was the ambassador in European countries of the scientific medicine that emerged from the Oswaldo Cruz Institute in the early twentieth century. Rocha Lima has discovered the causative agent of typhus and had a major contribution to the studies of yellow fever, Chagas disease, Carrión's disease and histoplasmosis. His genius, his research and his discoveries projected his name, and, with it, the image of Brazil in the international scientific scene.

  3. Berengario da Carpi.

    PubMed

    De Santo, N G; Bisaccia, C; De Santo, L S; De Santo, R M; Di Leo, V A; Papalia, T; Cirillo, M; Touwaide, A

    1999-01-01

    Berengario da Carpi was magister of anatomy and surgery at the University of Bologna from 1502 to 1527. Eustachio and Falloppia defined him as 'the restaurator of anatomy'. He was a great surgeon, anatomist and physician of illustrious patients including Lorenzo II dei Medici, Giovanni dalle Bande Nere, Galeazzo Pallavicini, Cardinal Colonna, and Alessandro Soderini. He had strong links to the intellectuals of his time (Forni, Bonamici, Manuzio, Pomponazzi) as well as with the Medici family. He was respected by the Popes Julius II, Leo X and Clement VII. His main contributions are the Isogogae Breves, De Fractura calvae sive cranei, and the illustrated Commentaria on the Anatomy of Mondino de Liucci, a textbook utilized for more than 200 years, which Berengario aimed to restore to its initial text. The Commentaria constitutes the material for the last part of this paper which concludes with a personal translation of some passages on 'The kidney', where the author gives poignant examples of experimental ingenuity.

  4. Henrique da Rocha Lima*

    PubMed Central

    Bernardes Filho, Fred; Avelleira, João Carlos Regazzi

    2015-01-01

    Brazilian physician and researcher Henrique da Rocha Lima was born in 1879 in the city of Rio de Janeiro, where he studied medicine and obtained the degree of M.D. in 1901. He specialized in Clinical Medicine in Germany and was the ambassador in European countries of the scientific medicine that emerged from the Oswaldo Cruz Institute in the early twentieth century. Rocha Lima has discovered the causative agent of typhus and had a major contribution to the studies of yellow fever, Chagas disease, Carrión’s disease and histoplasmosis. His genius, his research and his discoveries projected his name, and, with it, the image of Brazil in the international scientific scene. PMID:26131867

  5. WRS2 UPA DA Removal

    NASA Image and Video Library

    2009-11-23

    ISS021-E-032275 (23 Nov. 2009) --- NASA astronaut Leland Melvin, STS-129 mission specialist, holds the failed Urine Processor Assembly / Distillation Assembly (UPA DA) in the Destiny laboratory of the International Space Station while space shuttle Atlantis remains docked with the station. Melvin and European Space Agency astronaut Frank De Winne (out of frame), Expedition 21 commander, removed and packed the UPA DA, then transferred it from the Water Recovery System 2 (WRS-2) rack to Atlantis for stowage on the middeck.

  6. WRS2 UPA DA Removal

    NASA Image and Video Library

    2009-11-23

    ISS021-E-032273 (23 Nov. 2009) --- European Space Agency astronaut Frank De Winne, Expedition 21 commander, holds the failed Urine Processor Assembly / Distillation Assembly (UPA DA) in the Destiny laboratory of the International Space Station while space shuttle Atlantis remains docked with the station. De Winne and NASA astronaut Leland Melvin (out of frame), STS-129 mission specialist, removed and packed the UPA DA, then transferred it from the Water Recovery System 2 (WRS-2) rack to Atlantis for stowage on the middeck.

  7. da Vinci decoded: does da Vinci stereopsis rely on disparity?

    PubMed

    Tsirlin, Inna; Wilcox, Laurie M; Allison, Robert S

    2012-11-01

    In conventional stereopsis, the depth between two objects is computed based on the retinal disparity in the position of matching points in the two eyes. When an object is occluded by another object in the scene, so that it is visible only in one eye, its retinal disparity cannot be computed. Nakayama and Shimojo (1990) found that a precept of quantitative depth between the two objects could still be established for such stimuli and proposed that this precept is based on the constraints imposed by occlusion geometry. They named this and other occlusion-based depth phenomena "da Vinci stereopsis." Subsequent research found quantitative depth based on occlusion geometry in several other classes of stimuli grouped under the term da Vinci stereopsis. However, Nakayama and Shimojo's findings were later brought into question by Gillam, Cook, and Blackburn (2003), who suggested that quantitative depth in their stimuli was perceived based on conventional disparity. In order to understand whether da Vinci stereopsis relies on one type of mechanism or whether its function is stimulus dependent we examine the nature and source of depth in the class of stimuli used by Nakayama and Shimojo (1990). We use three different psychophysical and computational methods to show that the most likely source for depth in these stimuli is occlusion geometry. Based on these experiments and previous data we discuss the potential mechanisms responsible for processing depth from monocular features in da Vinci stereopsis.

  8. Leonardo da Vinci and the Downburst.

    NASA Astrophysics Data System (ADS)

    Gedzelman, Stanley David

    1990-05-01

    Evidence from the drawings, experiments, and writings of Leonardo da Vinci are presented to demonstrate that da Vinci recognized and, possibly, discovered the downburst and understood its associated airflow. Other early references to vortex flows resembling downbursts are mentioned.

  9. The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

    PubMed

    Park, S D; Kim, J S; Han, S H; Ha, Y K; Song, K S; Jee, K Y

    2009-09-01

    In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

  10. Case study of the propagation of a small flaw under PWR loading conditions and comparison with the ASME code design life. Comparison of ASME Code Sections III and XI

    SciTech Connect

    Yahr, G.T.; Gwaltney, R.C.; Richardson, A.K.; Server, W.L.

    1986-01-01

    A cooperative study was performed by EG and G Idaho, Inc., and Oak Ridge National Laboratory to investigate the degree of conservatism and consistency in the ASME Boiler and Pressure Vessel Code Section III fatigue evaluation procedure and Section XI flaw acceptance standards. A single, realistic, sample problem was analyzed to determine the significance of certain points of criticism made of an earlier parametric study by staff members of the Division of Engineering Standards of the Nuclear Regulatory Commission. The problem was based on a semielliptical flaw located on the inside surface of the hot-leg piping at the reactor vessel safe-end weld for the Zion 1 pressurized-water reactor (PWR). Two main criteria were used in selecting the problem; first, it should be a straight pipe to minimize the computational expense; second, it should exhibit as high a cumulative usage factor as possible. Although the problem selected has one of the highest cumulative usage factors of any straight pipe in the primary system of PWRs, it is still very low. The Code Section III fatigue usage factor was only 0.00046, assuming it was in the as-welded condition, and fatigue crack-growth analyses predicted negligible crack growth during the 40-year design life. When the analyses were extended past the design life, the usage factor was less than 1.0 when the flaw had propagated to failure. The current study shows that the criticism of the earlier report should not detract from the conclusion that if a component experiences a high level of cyclic stress corresponding to a fatigue usage factor near 1.0, very small cracks can propagate to unacceptable sizes.

  11. The major histocompatibility complex genes impact pain response in DA and DA.1U rats.

    PubMed

    Guo, Yuan; Yao, Fan-Rong; Cao, Dong-Yuan; Li, Li; Wang, Hui-Sheng; Xie, Wen; Zhao, Yan

    2015-08-01

    Our recent studies have shown that the difference in basal pain sensitivity to mechanical and thermal stimulation between Dark-Agouti (DA) rats and a novel congenic DA.1U rats is major histocompatibility complex (MHC) genes dependent. In the present study, we further used DA and DA.1U rats to investigate the role of MHC genes in formalin-induced pain model by behavioral, electrophysiological and immunohistochemical methods. Behavioral results showed biphasic nociceptive behaviors increased significantly following the intraplantar injection of formalin in the hindpaw of DA and DA.1U rats. The main nociceptive behaviors were lifting and licking, especially in DA rats (P<0.001 and P<0.01). The composite pain scores (CPS) in DA rats were significantly higher than those in DA.1U rats in both phases of the formalin test (P<0.01). Electrophysiological results also showed the biphasic increase in discharge rates of C and Aδ fibers of L5 dorsal root in the two strains, and the net change of the discharge rate of DA rats was significantly higher than that of DA.1U rats (P<0.05). The mechanical thresholds decreased after formalin injection in both strains (P<0.01), and the net change in the mechanical threshold in DA was greater than that in DA.1U rats (P<0.05). The expression of RT1-B, representation of MHC class II molecule, in laminae I-II of L4/5 spinal cord in DA rats was significantly higher than that in DA.1U rats in the respective experimental group (P<0.05). These results suggested that both DA and DA.1U rats exhibited nociceptive responses in formalin-induced pain model and DA rats were more sensitive to noxious chemical stimulus than DA.1U rats, indicating that MHC genes might contribute to the difference in pain sensitivity.

  12. Autocorrelation descriptor improvements for QSAR: 2DA_Sign and 3DA_Sign

    NASA Astrophysics Data System (ADS)

    Sliwoski, Gregory; Mendenhall, Jeffrey; Meiler, Jens

    2016-03-01

    Quantitative structure-activity relationship (QSAR) is a branch of computer aided drug discovery that relates chemical structures to biological activity. Two well established and related QSAR descriptors are two- and three-dimensional autocorrelation (2DA and 3DA). These descriptors encode the relative position of atoms or atom properties by calculating the separation between atom pairs in terms of number of bonds (2DA) or Euclidean distance (3DA). The sums of all values computed for a given small molecule are collected in a histogram. Atom properties can be added with a coefficient that is the product of atom properties for each pair. This procedure can lead to information loss when signed atom properties are considered such as partial charge. For example, the product of two positive charges is indistinguishable from the product of two equivalent negative charges. In this paper, we present variations of 2DA and 3DA called 2DA_Sign and 3DA_Sign that avoid information loss by splitting unique sign pairs into individual histograms. We evaluate these variations with models trained on nine datasets spanning a range of drug target classes. Both 2DA_Sign and 3DA_Sign significantly increase model performance across all datasets when compared with traditional 2DA and 3DA. Lastly, we find that limiting 3DA_Sign to maximum atom pair distances of 6 Å instead of 12 Å further increases model performance, suggesting that conformational flexibility may hinder performance with longer 3DA descriptors. Consistent with this finding, limiting the number of bonds in 2DA_Sign from 11 to 5 fails to improve performance.

  13. Autocorrelation descriptor improvements for QSAR: 2DA_Sign and 3DA_Sign.

    PubMed

    Sliwoski, Gregory; Mendenhall, Jeffrey; Meiler, Jens

    2016-03-01

    Quantitative structure-activity relationship (QSAR) is a branch of computer aided drug discovery that relates chemical structures to biological activity. Two well established and related QSAR descriptors are two- and three-dimensional autocorrelation (2DA and 3DA). These descriptors encode the relative position of atoms or atom properties by calculating the separation between atom pairs in terms of number of bonds (2DA) or Euclidean distance (3DA). The sums of all values computed for a given small molecule are collected in a histogram. Atom properties can be added with a coefficient that is the product of atom properties for each pair. This procedure can lead to information loss when signed atom properties are considered such as partial charge. For example, the product of two positive charges is indistinguishable from the product of two equivalent negative charges. In this paper, we present variations of 2DA and 3DA called 2DA_Sign and 3DA_Sign that avoid information loss by splitting unique sign pairs into individual histograms. We evaluate these variations with models trained on nine datasets spanning a range of drug target classes. Both 2DA_Sign and 3DA_Sign significantly increase model performance across all datasets when compared with traditional 2DA and 3DA. Lastly, we find that limiting 3DA_Sign to maximum atom pair distances of 6 Å instead of 12 Å further increases model performance, suggesting that conformational flexibility may hinder performance with longer 3DA descriptors. Consistent with this finding, limiting the number of bonds in 2DA_Sign from 11 to 5 fails to improve performance.

  14. The potentially dangerous asteroid 2012 DA14

    NASA Astrophysics Data System (ADS)

    Wlodarczyk, I.

    2012-12-01

    We present computing methods that allow us to study the behaviour of the dynamically interesting potentially dangerous asteroid 2012 DA14. Using the freely available ORBFIT software, we can follow the orbit of the asteroid backward and forward in the future, searching for close approaches to the Earth that might lead to possible impacts. The possible impact orbit for 2026 is computed. We show that it should be possible to recover asteroid 2012 DA14, mainly in 2013 February. It is highly unlikely that asteroid 2012 DA14 will hit any geosynchronous satellites during its close approach on 2013 February 15.

  15. Timing analysis of PWR fuel pin failures

    SciTech Connect

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

    1992-09-01

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

  16. PWR cores with silicon carbide cladding

    SciTech Connect

    Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S.

    2012-07-01

    The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

  17. Consolidation and disposal of PWR fuel inserts

    SciTech Connect

    Wakeman, B.H. )

    1992-08-01

    Design and licensing of the Surry Power Station Independent Spent Fuel Storage Installation was initiated in 1982 by Virginia Power as part of a comprehensive strategy to increase spent fuel storage capacity at the Station. Designed to use large, metal dry storage casks, the Surry Installation will accommodate 84 such casks with a total storage capacity of 811 MTU of spent pressurized water reactor fuel assemblies. Virginia Power provided three storage casks for testing at the Idaho National Engineerinq Laboratory's Test Area North and the testing results have been published by the Electric Power Research Institute. Sixty-nine spent fuel assemblies were transported in truck casks from the Surry Power Station to Test Area North for testing in the three casks. Because of restrictions imposed by the cask testing equipment at Test Area North, the irradiated insert components stored in these fuel assemblies at Surry were removed prior to transport of the fuel assemblies. Retaining these insert components proved to be a problem because of a shortage of spent fuel assemblies in the spent fuel storage pool that did not already contain insert components. In 1987 Virginia Power contracted with Chem-Nuclear Systems, Inc. to process and dispose of 136 irradiated insert components consisting of 125 burnable poison rod assemblies, 10 thimble plugging devices and 1 part-length rod cluster control assembly. This work was completed in August and September 1987, culminating in the disposal at the Barnwell, SC low-level radioactive waste facility of two CNS 3-55 liners containing the consolidated insert components.

  18. Two-phase jet loads. [PWR

    SciTech Connect

    Tomasko, D.

    1980-01-01

    Two-phase jets are currently being studied to improve engineering models for the prediction of loads on pipes and structures during LOCAs. Multi-dimensional computer codes such as BEACON/MOD2, CSQ, and TRAC-P1A are being employed to predict flow characteristics and flow-structure loading. Our ultimate goal is to develop a new approximate engineering model which is superior to the F.J. Moody design model. Computer results are compared with data obtained from foreign sources, and a technique for using the TRAC-P1A vessel component as a containment model is presented. In general, good agreement with the data is obtained for saturated stagnation conditions; however, difficulties are encountered for subcooled stagnation conditions, possibly due to nucleation delay and non-equilibrium effects.

  19. Systems interaction study of a Westinghouse PWR

    SciTech Connect

    Youngblood, R.; Hanan, N.; Fitzpatrick, R.; Xue, D.; Bozoki, G.; Fresco, A.; Papazoglou, I.A.

    1985-01-01

    This paper presents methods and findings of a systems interaction study of Indian Point 3. The study was carried out in support of the resolution of Unresolved Safety Issue A-17 on Systems Interactions. Fault tree methods were employed. Among the study's findings is a single active failure in the low pressure injection function; this discovery led to a plant modification. In addition to providing support to the staff in resolving USI A-17, the project discovered an important new class of failure modes which led the utility to implement a hardware modification. The scope of the project is indicated, key features of the method are highlighted findings are discussed, and comments are offered on the usefulness of this type of, principal study. 9 refs., 1 fig., 1 tab.

  20. Tritium removal and retention device. [PWR; BWR

    SciTech Connect

    Boyle, R.F.; Durigon, D.D.

    1981-07-21

    Apparatus comprising a two layered composite with an internal core of zirconium or zirconium alloy which retains tritium, and an adherent nickel outer layer which acts as a protective and selective window for passage of the tritium.

  1. PWR upper/lower internals shield

    SciTech Connect

    Homyk, W.A.

    1995-03-01

    During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

  2. Phenomenological modelling of steam explosions. [PWR; BWR

    SciTech Connect

    Corradini, M.L.; Drumheller, D.S.

    1980-01-01

    During a hypothetical core meltdown accident, an important safety issue to be addressed is the potential for steam explosions. This paper presents analysis and modelling of experimental results. There are four observations that can be drawn from the analysis: (1) vapor explosions are suppressed by noncondensible gases generated by fuel oxidation, by high ambient pressure, and by high water temperatures; (2) these effects appear to be trigger-related in that an explosion can again be induced in some cases by increasing the trigger magnitude; (3) direct fuel liquid-coolant liquid contact can explain small scale fuel fragmentation; (4) heat transfer during the expansion phase of the explosion can reduce the work potential.

  3. US PWR steam generator management: An overview

    SciTech Connect

    Welty, C.S. Jr.

    1997-02-01

    This paper provides an overview on the status of steam generator management activities in US PWRs, and includes: (1) an overview of the impact of steam generator problems; (2) a brief discussion of historical damage trends and the current damage mechanism of most concern; (3) a discussion of the elements of {open_quotes}steam generator management{close_quotes}; and (4) a description of the approach being followed to implement a degradation-specific protocol for tubing inspection and repair. This paper was prepared in conjunction with another paper presented during the Plenary Session of this Conference, {open_quotes}Steam Generator Degradation: Current Mitigation Strategies for Controlling Corrosion{close_quotes}, and is provided as a supplement to that material.

  4. Origin of the DA and non-DA white dwarf stars

    NASA Technical Reports Server (NTRS)

    Shipman, Harry L.

    1989-01-01

    Various proposals for the bifurcation of the white dwarf cooling sequence are reviewed. 'Primordial' theories, in which the basic bifurcation of the white dwarf sequence is rooted in events predating the white dwarf stage of stellar evolution, are discussed, along with the competing 'mixing' theories in which processes occurring during the white dwarf stage are responsible for the existence of DA or non-DA stars. A new proposal is suggested, representing a two-channel scenario. In the DA channel, some process reduces the hydrogen layer mass to the value of less than 10 to the -7th. The non-DA channel is similar to that in the primordial scenario. These considerations suggest that some mechanism operates in both channels to reduce the thickness of the outermost layer of the white dwarf. It is also noted that accretion from the interstellar medium has little to do with whether a particular white dwarf becomes a DA or a non-DA star.

  5. An Extension of PPLS-DA for Classification and Comparison to Ordinary PLS-DA

    PubMed Central

    Telaar, Anna; Liland, Kristian Hovde; Repsilber, Dirk; Nürnberg, Gerd

    2013-01-01

    Classification studies are widely applied, e.g. in biomedical research to classify objects/patients into predefined groups. The goal is to find a classification function/rule which assigns each object/patient to a unique group with the greatest possible accuracy (classification error). Especially in gene expression experiments often a lot of variables (genes) are measured for only few objects/patients. A suitable approach is the well-known method PLS-DA, which searches for a transformation to a lower dimensional space. Resulting new components are linear combinations of the original variables. An advancement of PLS-DA leads to PPLS-DA, introducing a so called ‘power parameter’, which is maximized towards the correlation between the components and the group-membership. We introduce an extension of PPLS-DA for optimizing this power parameter towards the final aim, namely towards a minimal classification error. We compare this new extension with the original PPLS-DA and also with the ordinary PLS-DA using simulated and experimental datasets. For the investigated data sets with weak linear dependency between features/variables, no improvement is shown for PPLS-DA and for the extensions compared to PLS-DA. A very weak linear dependency, a low proportion of differentially expressed genes for simulated data, does not lead to an improvement of PPLS-DA over PLS-DA, but our extension shows a lower prediction error. On the contrary, for the data set with strong between-feature collinearity and a low proportion of differentially expressed genes and a large total number of genes, the prediction error of PPLS-DA and the extensions is clearly lower than for PLS-DA. Moreover we compare these prediction results with results of support vector machines with linear kernel and linear discriminant analysis. PMID:23408965

  6. An extension of PPLS-DA for classification and comparison to ordinary PLS-DA.

    PubMed

    Telaar, Anna; Liland, Kristian Hovde; Repsilber, Dirk; Nürnberg, Gerd

    2013-01-01

    Classification studies are widely applied, e.g. in biomedical research to classify objects/patients into predefined groups. The goal is to find a classification function/rule which assigns each object/patient to a unique group with the greatest possible accuracy (classification error). Especially in gene expression experiments often a lot of variables (genes) are measured for only few objects/patients. A suitable approach is the well-known method PLS-DA, which searches for a transformation to a lower dimensional space. Resulting new components are linear combinations of the original variables. An advancement of PLS-DA leads to PPLS-DA, introducing a so called 'power parameter', which is maximized towards the correlation between the components and the group-membership. We introduce an extension of PPLS-DA for optimizing this power parameter towards the final aim, namely towards a minimal classification error. We compare this new extension with the original PPLS-DA and also with the ordinary PLS-DA using simulated and experimental datasets. For the investigated data sets with weak linear dependency between features/variables, no improvement is shown for PPLS-DA and for the extensions compared to PLS-DA. A very weak linear dependency, a low proportion of differentially expressed genes for simulated data, does not lead to an improvement of PPLS-DA over PLS-DA, but our extension shows a lower prediction error. On the contrary, for the data set with strong between-feature collinearity and a low proportion of differentially expressed genes and a large total number of genes, the prediction error of PPLS-DA and the extensions is clearly lower than for PLS-DA. Moreover we compare these prediction results with results of support vector machines with linear kernel and linear discriminant analysis.

  7. The Supernova Impostor SN 2010da

    NASA Astrophysics Data System (ADS)

    Binder, Breanna A.; Williams, Benjamin F.; Kong, Albert K. H.; Plucinsky, Paul P.; Gaetz, Terrance J.; Skillman, Evan D.; Dolphin, Andrew E.

    2016-01-01

    Supernova impostors are optical transients that, despite being assigned a supernova designation, do not signal the death of a massive star or accreting white dwarf. Instead, many impostors are thought to be major eruptions from luminous blue variables. Although the physical cause of these eruptions is still debated, tidal interactions from a binary companion has recently gained traction as a possible explanation for observations of some supernova impostors. In this talk, I will discuss the particularly interesting impostor SN 2010da, which exhibits high-luminosity, variable X-ray emission. The X-ray emission is consistent with accretion onto a neutron star, making SN 2010da a likely high mass X-ray binary in addition to a supernova impostor. SN 2010da is a unique laboratory for understanding both binary interactions as drivers of massive star eruptions and the evolutionary processes that create high mass X-ray binaries.

  8. Hidden sketches by Leonardo da Vinci revealed

    NASA Astrophysics Data System (ADS)

    Dumé, Belle

    2009-02-01

    Three drawings on the back of Leonardo da Vinci's The Virgin and Child with St Anne (circa 1508) have been discovered by researchers led by Michel Menu from the Centre de Recherche et de Restauration des Musées de France (C2RMF) and the Louvre Museum in Paris.

  9. DA white dwarfs in the Kepler field

    NASA Astrophysics Data System (ADS)

    Doyle, T. F.; Howell, S. B.; Petit, V.; Lépine, S.

    2017-01-01

    We present 16 new, and confirm 7 previously identified, DA white dwarfs in the Kepler field through ground-based spectroscopy with the Hale 200″, Kitt Peak 4-m, and Bok 2.3-m telescopes. Using atmospheric models, we determine their effective temperatures and surface gravities to constrain their position with respect to the ZZ Ceti (DA pulsator) instability strip, and look for the presence or absence of pulsation with Kepler's unprecedented photometry. Our results are as follows. (i) From our measurements of temperature and surface gravity, 12 of the 23 DA white dwarfs from this work fall well outside of the instability strip. The Kepler photometry available for 11 of these WDs allows us to confirm that none are pulsating. One of these 11 happens to be a presumed binary, KIC 11604781, with a period of ˜5 d. (ii) The remaining 11 DA white dwarfs are instability strip candidates, potentially falling within the current, empirical instability strip, after accounting for uncertainties. These WDs will help constrain the strip's location further, as eight are near the blue edge and three are near the red edge of the instability strip. Four of these WDs do not have Kepler photometry, so ground-based photometry is needed to determine the pulsation nature of these white dwarfs. The remaining seven have Kepler photometry available, but do not show any periodicity on typical WD pulsation time-scales.

  10. How to Think Like Leonardo da Vinci

    ERIC Educational Resources Information Center

    Caouette, Ralph

    2008-01-01

    To be effective and relevant in twenty-first-century learning, art needs to be more inclusive. In this article, the author discusses how teachers can find a good example in Leonardo da Vinci for building an art program. His art, design, and curiosity are the perfect foundation for any art program, at any level. (Contains 3 resources and 3 online…

  11. How to Think Like Leonardo da Vinci

    ERIC Educational Resources Information Center

    Caouette, Ralph

    2008-01-01

    To be effective and relevant in twenty-first-century learning, art needs to be more inclusive. In this article, the author discusses how teachers can find a good example in Leonardo da Vinci for building an art program. His art, design, and curiosity are the perfect foundation for any art program, at any level. (Contains 3 resources and 3 online…

  12. VerSeDa: vertebrate secretome database

    PubMed Central

    Cortazar, Ana R.; Oguiza, José A.

    2017-01-01

    Based on the current tools, de novo secretome (full set of proteins secreted by an organism) prediction is a time consuming bioinformatic task that requires a multifactorial analysis in order to obtain reliable in silico predictions. Hence, to accelerate this process and offer researchers a reliable repository where secretome information can be obtained for vertebrates and model organisms, we have developed VerSeDa (Vertebrate Secretome Database). This freely available database stores information about proteins that are predicted to be secreted through the classical and non-classical mechanisms, for the wide range of vertebrate species deposited at the NCBI, UCSC and ENSEMBL sites. To our knowledge, VerSeDa is the only state-of-the-art database designed to store secretome data from multiple vertebrate genomes, thus, saving an important amount of time spent in the prediction of protein features that can be retrieved from this repository directly. Database URL: VerSeDa is freely available at http://genomics.cicbiogune.es/VerSeDa/index.php PMID:28365718

  13. VerSeDa: vertebrate secretome database.

    PubMed

    Cortazar, Ana R; Oguiza, José A; Aransay, Ana M; Lavín, José L

    2017-01-01

    Based on the current tools, de novo secretome (full set of proteins secreted by an organism) prediction is a time consuming bioinformatic task that requires a multifactorial analysis in order to obtain reliable in silico predictions. Hence, to accelerate this process and offer researchers a reliable repository where secretome information can be obtained for vertebrates and model organisms, we have developed VerSeDa (Vertebrate Secretome Database). This freely available database stores information about proteins that are predicted to be secreted through the classical and non-classical mechanisms, for the wide range of vertebrate species deposited at the NCBI, UCSC and ENSEMBL sites. To our knowledge, VerSeDa is the only state-of-the-art database designed to store secretome data from multiple vertebrate genomes, thus, saving an important amount of time spent in the prediction of protein features that can be retrieved from this repository directly. VerSeDa is freely available at http://genomics.cicbiogune.es/VerSeDa/index.php.

  14. The local effect of octreotide on mechanical pain sensitivity is more sensitive in DA rats than DA.1U rats.

    PubMed

    Yao, Fan-Rong; Wang, Hui-Sheng; Guo, Yuan; Zhao, Yan

    2016-02-01

    A recent study by the authors indicated that major histocompatibility complex (MHC) genes are associated with the differences in basal pain sensitivity and in formalin model between Dark-Agouti (DA) and novel congenic DA.1U rats, which have the same genetic background as DA rats except for the u alleles of MHC. The objective of the present study is to investigate whether there is a difference in the pristane-induced arthritis (PIA) model and local analgesic effect of octreotide (OCT) between DA and DA.1U rats. The hindpaw mechanical withdrawal threshold (MWT) and heat withdrawal latency (HWL) were observed. The C unit firings of the tibial nerve evoked by non-noxious and noxious toe movements were recorded by electrophysiological methods in normal and PIA models in DA and DA.1U rats before and after local OCT administration. The expression of somatostatin receptor 2A (SSTR2A) was observed by immunohistochemistry. The results demonstrate that DA rats have a higher mechanical sensitivity than DA.1U rats after PIA. Local OCT administration significantly elevated MWT in DA rats under normal and PIA sate, but not in DA.1U rats. The electrophysiological experiments showed OCT significantly attenuated the firings of C units evoked by non-noxious and noxious stimulation in DA rats more than those in DA.1U rats both in normal and PIA states. In addition, the expression of SSTR2A in the dorsal horn of the spinal cord was significantly higher in DA than in DA.1U rats. All of the findings suggest a higher local analgesic effect of OCT in DA rats than DA.1U rats, which might be associated with the MHC genes.

  15. A Day in the Life at DaVita Academy

    ERIC Educational Resources Information Center

    Weinstein, Margery

    2010-01-01

    When a company name means "giving life," the bar for learning and development programs is held high. In this article, the author describes what it takes to graduate from DaVita Academy, the soft skills training program dialysis services company DaVita offers all its employees. DaVita's chief executive officer, Kent Thiry, states that the Academy…

  16. A Day in the Life at DaVita Academy

    ERIC Educational Resources Information Center

    Weinstein, Margery

    2010-01-01

    When a company name means "giving life," the bar for learning and development programs is held high. In this article, the author describes what it takes to graduate from DaVita Academy, the soft skills training program dialysis services company DaVita offers all its employees. DaVita's chief executive officer, Kent Thiry, states that the Academy…

  17. 32 CFR 516.25 - DA Form 4.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true DA Form 4. 516.25 Section 516.25 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS LITIGATION Reporting Legal Proceedings to HQDA § 516.25 DA Form 4. (a) General. The DA Form 4 (See figure...

  18. Homogeneous generation of iDA neurons with high similarity to bona fide DA neurons using a drug inducible system.

    PubMed

    Park, Hanseul; Kim, Hongwon; Yoo, Junsang; Lee, Jaekwang; Choi, Hwan; Baek, Soonbong; Lee, C Justin; Kim, Janghwan; Lengner, Christopher J; Sung, Jung-Suk; Kim, Jongpil

    2015-12-01

    Recent work generating induced dopaminergic (iDA) neurons using direct lineage reprogramming potentially provides a novel platform for the study and treatment Parkinson's disease (PD). However, one of the most important issues for iDA-based applications is the degree to which iDA neurons resemble the molecular and functional properties of their endogenous DA neuron counterparts. Here we report that the homogeneity of the reprogramming gene expression system is critical for the generation of iDA neuron cultures that are highly similar to endogenous DA neurons. We employed an inducible system that carries iDA-inducing factors as defined transgenes for direct lineage reprogramming to iDA neurons. This system circumvents the need for viral transduction, enabling a more efficient and reproducible reprogramming process for the generation of genetically homogenous iDA neurons. We showed that this inducible system generates iDA neurons with high similarity to their bona fide in vivo counterparts in comparison to direct infection methods. Thus, our results suggest that homogenous expression of exogenous genes in direct lineage reprogramming is critical for the generation of high quality iDA neuron cultures, making such culture systems a valuable resource for iDA-based drug screening and, ultimately, potential therapeutic intervention in PD. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. PanDA for COMPASS at JINR

    NASA Astrophysics Data System (ADS)

    Petrosyan, A. Sh.

    2016-09-01

    PanDA (Production and Distributed Analysis System) is a workload management system, widely used for data processing at experiments on Large Hadron Collider and others. COMPASS is a high-energy physics experiment at the Super Proton Synchrotron. Data processing for COMPASS runs locally at CERN, on lxbatch, the data itself stored in CASTOR. In 2014 an idea to start running COMPASS production through PanDA arose. Such transformation in experiment's data processing will allow COMPASS community to use not only CERN resources, but also Grid resources worldwide. During the spring and summer of 2015 installation, validation and migration work is being performed at JINR. Details and results of this process are presented in this paper.

  20. [Leonardo da Vinci--a dyslectic genius?].

    PubMed

    Røsstad, Anna

    2002-12-10

    Leonardo da Vinci's texts consist almost exclusively of scientific notes. Working on a book on Leonardo's art, I studied all Leonardo's published texts carefully for any new information. In some prefaces I came to suspect that Leonardo might have suffered from dyslexia. This article considers the question of whether it is possible to find indications of dyslexia in Leonardo's texts and in the accounts of his life.

  1. DA 495: An Aging Pulsar Wind Nebula

    NASA Astrophysics Data System (ADS)

    Kothes, R.; Landecker, T. L.; Reich, W.; Safi-Harb, S.; Arzoumanian, Z.

    2008-11-01

    We present a radio continuum study of the pulsar wind nebula (PWN) DA 495 (G65.7+1.2), including images of total intensity and linear polarization from 408 to 10550 MHz based on the Canadian Galactic Plane Survey and observations with the Effelsberg 100 m Radio Telescope. Removal of flux density contributions from a superimposed H II region and from compact extragalactic sources reveals a break in the spectrum of DA 495 at 1.3 GHz, with a spectral index α = - 0.45 +/- 0.20 below the break and α = - 0.87 +/- 0.10 above it (Sν propto να). The spectral break is more than 3 times lower in frequency than the lowest break detected in any other PWN. The break in the spectrum is likely the result of synchrotron cooling, and DA 495, at an age of ~20,000 yr, may have evolved from an object similar to the Vela X nebula, with a similarly energetic pulsar. We find a magnetic field of ~1.3 mG inside the nebula. After correcting for the resulting high internal rotation measure, the magnetic field structure is quite simple, resembling the inner part of a dipole field projected onto the plane of the sky, although a toroidal component is likely also present. The dipole field axis, which should be parallel to the spin axis of the putative pulsar, lies at an angle of ~50° east of the north celestial pole and is pointing away from us toward the southwest. The upper limit for the radio surface brightness of any shell-type supernova remnant emission around DA 495 is Σ1GHz ~ 5.4 × 10-23 W m-2 Hz-1 sr-1 (assuming a radio spectral index of α = - 0.5), lower than the faintest shell-type remnant known to date.

  2. UBV photometry of hot DA white dwarfs

    NASA Technical Reports Server (NTRS)

    Kidder, K. M.; Holberg, J. B.; Mason, Paul A.

    1991-01-01

    Johnson UBV photometry has been obtained photoelectrically for a set of DA white dwarfs with effective temperatures greater than 20,000 K and for the AM Her type binary HO538 + 608. Most of the white dwarfs lie within existing Einstein IPC or EXOSAT LE soft X-ray fields, therefore they are of interest as potential serendipitous soft X-ray sources. In addition, high dispersion spectroscopy has been used to differentiate seven of these objects to be subdwarfs.

  3. A computational theory of da Vinci stereopsis.

    PubMed

    Tsirlin, Inna; Wilcox, Laurie M; Allison, Robert S

    2014-06-09

    In binocular vision, occlusion of one object by another gives rise to monocular occlusions—regions visible only in one eye. Although binocular disparities cannot be computed for these regions, monocular occlusions can be precisely localized in depth and can induce the perception of illusory occluding surfaces. The phenomenon of depth perception from monocular occlusions, known as da Vinci stereopsis, is intriguing, but its mechanisms are not well understood. We first propose a theory of the mechanisms underlying da Vinci stereopsis that is based on the psychophysical and computational literature on monocular occlusions. It postulates, among other principles, that monocular areas are detected explicitly, and depth from occlusions is calculated based on constraints imposed by occlusion geometry. Next, we describe a biologically inspired computational model based on this theory that successfully reconstructs depth in a large range of stimuli and produces results similar to those described in the psychophysical literature. These results demonstrate that the proposed neural architecture could underpin da Vinci stereopsis and other stereoscopic percepts. © 2014 ARVO.

  4. Leonardo da Vinci's studies of the heart.

    PubMed

    Shoja, Mohammadali M; Agutter, Paul S; Loukas, Marios; Benninger, Brion; Shokouhi, Ghaffar; Namdar, Husain; Ghabili, Kamyar; Khalili, Majid; Tubbs, R Shane

    2013-08-20

    Leonardo da Vinci's detailed drawings are justly celebrated; however, less well known are his accounts of the structures and functions of the organs. In this paper, we focus on his illustrations of the heart, his conjectures about heart and blood vessel function, his experiments on model systems to test those conjectures, and his unprecedented conclusions about the way in which the cardiovascular system operates. In particular, da Vinci seems to have been the first to recognize that the heart is a muscle and that systole is the active phase of the pump. He also seems to have understood the functions of the auricles and pulmonary veins, identified the relationship between the cardiac cycle and the pulse, and explained the hemodynamic mechanism of valve opening and closure. He also described anatomical variations and changes in structure and function that occurred with age. We outline da Vinci's varied career and suggest ways in which his personality, experience, skills and intellectual heritage contributed to these advances in understanding. We also consider his influence on later studies in anatomy and physiology.

  5. The real code of leonardo da vinci.

    PubMed

    Ose, Leiv

    2008-02-01

    Leonardo da Vinci was born in Italy. Among the researchers and scientists, he is favourably known for his remarkable efforts in scientific work. His investigations of atherosclerosis judiciously combine three separate fields of research. In 1506, he finished his masterpiece, painting of Mona Lisa. A careful clinical examination of the famous painting reveals a yellow irregular leather-like spot at the inner end of the left upper eyelid and a soft bumpy well-defined swelling of the dorsum of the right hand beneath the index finger about 3 cm long. This is probably the first case of familial hypercholesterolemia (FH). The FH code of Leonardo da Vinci was given immense consideration by scientists like Carl Muller, who described the xanthomas tuberosum and angina pectoris. On the contrary, Akira Endo searched for microbial metabolites that would inhibit HMG-CoA reductase, the rate-limiting enzyme in the synthesis of cholesterol and finally, Michael Brown and Joseph Goldstein published a remarkable series of elegant and insightful papers in the 70s and 80s. They established that the cellular uptake of low-density lipoprotein (LDL) essentially requires the LDL receptor. this was the real Code of Leonardo da Vinci.

  6. The Real Code of Leonardo da Vinci

    PubMed Central

    Ose, Leiv

    2008-01-01

    Leonardo da Vinci was born in Italy. Among the researchers and scientists, he is favourably known for his remarkable efforts in scientific work. His investigations of atherosclerosis judiciously combine three separate fields of research. In 1506, he finished his masterpiece, painting of Mona Lisa. A careful clinical examination of the famous painting reveals a yellow irregular leather-like spot at the inner end of the left upper eyelid and a soft bumpy well-defined swelling of the dorsum of the right hand beneath the index finger about 3 cm long. This is probably the first case of familial hypercholesterolemia (FH). The FH code of Leonardo da Vinci was given immense consideration by scientists like Carl Muller, who described the xanthomas tuberosum and angina pectoris. On the contrary, Akira Endo searched for microbial metabolites that would inhibit HMG-CoA reductase, the rate-limiting enzyme in the synthesis of cholesterol and finally, Michael Brown and Joseph Goldstein published a remarkable series of elegant and insightful papers in the 70s and 80s. They established that the cellular uptake of low-density lipoprotein (LDL) essentially requires the LDL receptor. In conclusion: this was the real Code of Leonardo da Vinci. PMID:19924278

  7. Rational design of D-A1-D-A2 conjugated polymers with superior spectral coverage.

    PubMed

    Hedström, Svante; Tao, Qiang; Wang, Ergang; Persson, Petter

    2015-10-28

    The spectral coverage of a light-harvesting polymer largely determines the maximum achievable photocurrent in organic photovoltaics, and therefore constitutes a crucial parameter for improving their performance. The D-A1-D-A2 copolymer motif is a new and promising design strategy for extending the absorption range by incorporating two acceptor units with complementary photoresponses. The fundamental factors that promote an extended absorption are here determined for three prototype D-A1-D-A2 systems through a combination of experimental and computational methods. Systematic quantum chemical calculations are then used to reveal the intrinsic optical properties of ten further D-A1-D-A2 polymer candidates. These investigated polymers are all predicted to exhibit intense primary absorption peaks at 615-954 nm, corresponding to charge-transfer (CT) transitions to the stronger acceptor, as well as secondary absorption features at 444-647 nm that originate from CT transitions to the weaker acceptors. Realization of D-A1-D-A2 polymers with superior spectral coverage is thereby found to depend critically on the spatial and energetic separation between the two distinct acceptor LUMOs. Two promising D-A1-D-A2 copolymer candidates were finally selected for further theoretical and experimental study, and demonstrate superior light-harvesting properties in terms of significantly extended spectral coverage. This demonstrates great potential for enhanced light-harvesting in D-A1-D-A2 polymers via multiple absorption features compared to traditional D-A polymers.

  8. DA{phi}NE beam instrumentation

    SciTech Connect

    Ghigo, A.; Biscari, C.; Coiro, O.; Pirro, G. Di; Drago, A.; Gallo, A.; Marcellini, F.; Mazzitelli, G.; Milardi, C.; Sannibale, F.; Serio, M.; Stecchi, A.; Stella, A.; Vignola, G.; Zobov, M.

    1998-12-10

    DA{phi}NE, the Frascati {phi}-Factory, is now under commissioning. The accelerator complex is composed of a linac, an accumulator-damping ring, and two separate main rings, one for electrons and the other for positrons, with two interaction regions in which the experiments will be placed. In order to achieve the luminosity goal, high performance instrumentation and beam diagnostics have been installed. Some of the relevant beam measurements performed are: beam emittance, transverse and longitudinal dimensions, beam positions and tunes, overlap in the interaction points, and luminosity. An overview of the diagnostic instrumentation of the accelerator complex is given together with measurement examples and discussion of operational experiences.

  9. Polarization-switching D/A converter.

    PubMed

    Sun, Shunming; Kalkur, Thottam S

    2005-05-01

    This paper describes a novel digital-to-analog (D/A) conversion technique, which uses the analog quantity polarization as a D/A conversion medium. It can be implemented by CMOS capacitors or by ferroelectric capacitors, which exhibit strong nonlinearity in charge versus voltage behavior. Because a ferroelectric material inherently has spontaneous polarization and generally has a large dielectric constant, the effective capacitance of a ferroelectric capacitor is much larger than that of a CMOS capacitor of the same size. This ensures less influence of bottom-electrode parasitic capacitance on a ferroelectric capacitor. Furthermore, a data converter based on ferroelectric capacitors possesses the potential nonvolatile memory function owing to ferroelectric hysteresis. Along with the architecture proposed for polarization-switching digital-to-analog converter (PDAC), its circuit implementation is introduced. Described is implementation of two 9-bit bipolar PDACs: one is based on CMOS capacitors and the other on off-chip ferroelectric capacitors. Experimental results are presented for the performance of these two prototypes.

  10. 32 CFR 643.121 - Private organizations on DA installations.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 4 2011-07-01 2011-07-01 false Private organizations on DA installations. 643... (CONTINUED) REAL PROPERTY REAL ESTATE Additional Authority of Commanders § 643.121 Private organizations on DA installations. (a) AR 210-1 defines and classifies private organizations, such as thrift shops...

  11. 32 CFR 516.25 - DA Form 4.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 3 2014-07-01 2014-07-01 false DA Form 4. 516.25 Section 516.25 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC... attached to a properly prepared and sealed DA Form 4 are self-authenticating. (See Fed. R. Evid. 902). (b...

  12. 32 CFR 516.25 - DA Form 4.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 3 2012-07-01 2009-07-01 true DA Form 4. 516.25 Section 516.25 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS... properly prepared and sealed DA Form 4 are self-authenticating. (See Fed. R. Evid. 902). (b) Preparation at...

  13. DA-6034 Induces [Ca(2+)]i Increase in Epithelial Cells.

    PubMed

    Yang, Yu-Mi; Park, Soonhong; Ji, Hyewon; Kim, Tae-Im; Kim, Eung Kweon; Kang, Kyung Koo; Shin, Dong Min

    2014-04-01

    DA-6034, a eupatilin derivative of flavonoid, has shown potent effects on the protection of gastric mucosa and induced the increases in fluid and glycoprotein secretion in human and rat corneal and conjunctival cells, suggesting that it might be considered as a drug for the treatment of dry eye. However, whether DA-6034 induces Ca(2+) signaling and its underlying mechanism in epithelial cells are not known. In the present study, we investigated the mechanism for actions of DA-6034 in Ca(2+) signaling pathways of the epithelial cells (conjunctival and corneal cells) from human donor eyes and mouse salivary gland epithelial cells. DA-6034 activated Ca(2+)-activated Cl(-) channels (CaCCs) and increased intracellular calcium concentrations ([Ca(2+)]i) in primary cultured human conjunctival cells. DA-6034 also increased [Ca(2+)]i in mouse salivary gland cells and human corneal epithelial cells. [Ca(2+)]i increase of DA-6034 was dependent on the Ca(2+) entry from extracellular and Ca(2+) release from internal Ca(2+) stores. Interestingly, these effects of DA-6034 were related to ryanodine receptors (RyRs) but not phospholipase C/inositol 1,4,5-triphosphate (IP3) pathway and lysosomal Ca(2+) stores. These results suggest that DA-6034 induces Ca(2+) signaling via extracellular Ca(2+) entry and RyRs-sensitive Ca(2+) release from internal Ca(2+) stores in epithelial cells.

  14. The Case: Bunche-Da Vinci Learning Partnership Academy

    ERIC Educational Resources Information Center

    Eisenberg, Nicole; Winters, Lynn; Alkin, Marvin C.

    2005-01-01

    The Bunche-Da Vinci case described in this article presents a situation at Bunche Elementary School that four theorists were asked to address in their evaluation designs (see EJ791771, EJ719772, EJ791773, and EJ792694). The Bunche-Da Vinci Learning Partnership Academy, an elementary school located between an urban port city and a historically…

  15. The Case: Bunche-Da Vinci Learning Partnership Academy

    ERIC Educational Resources Information Center

    Eisenberg, Nicole; Winters, Lynn; Alkin, Marvin C.

    2005-01-01

    The Bunche-Da Vinci case described in this article presents a situation at Bunche Elementary School that four theorists were asked to address in their evaluation designs (see EJ791771, EJ719772, EJ791773, and EJ792694). The Bunche-Da Vinci Learning Partnership Academy, an elementary school located between an urban port city and a historically…

  16. Leonardo da Vinci's contributions to neuroscience.

    PubMed

    Pevsner, Jonathan

    2002-04-01

    Leonardo da Vinci (1452-1519) made far-reaching contributions to many areas of science, technology and art. Leonardo's pioneering research into the brain led him to discoveries in neuroanatomy (such as those of the frontal sinus and meningeal vessels) and neurophysiology (he was the first to pith a frog). His injection of hot wax into the brain of an ox provided a cast of the ventricles, and represents the first known use of a solidifying medium to define the shape and size of an internal body structure. Leonardo developed an original, mechanistic model of sensory physiology. He undertook his research with the broad goal of providing physical explanations of how the brain processes visual and other sensory input, and integrates that information via the soul.

  17. The DA{phi}NE luminosity monitor

    SciTech Connect

    Di Pirro, G.; Drago, A.; Ghigo, A.; Mazzitelli, G.; Preger, M.; Sannibale, F.; Serio, M.; Vignola, G.; Cervelli, F.; Lomtadze, T.

    1998-12-10

    DA{phi}NE, the Frascati {phi}-factory, is an e{sup +}/e{sup -} collider with 2 interaction points (IPs). The center of mass energy is 1020 MeV and the design luminosity 4.2x10{sup 30} cm{sup -2} s{sup -1} in single bunch mode and 5x10{sup 32} cm{sup -2} s{sup -1} in multibunch mode. Between the possible electromagnetic reactions at the interaction point, single bremsstrahlung (SB) has been selected for the luminosity measurement. The SB high counting rate allows real-time monitoring, which is very useful during machine tune-up and moreover the narrow peak of the SB angular distribution makes the counting rate almost independent from the beam position at the IP. A description of the experimental set-up, calibration results and luminosity measurements is presented.

  18. Da Costa's syndrome or neurocirculatory asthenia.

    PubMed Central

    Paul, O

    1987-01-01

    The syndrome variously called Da Costa's syndrome, effort syndrome, neurocirculatory asthenia, etc has been studied for more than 100 years by many distinguished physicians. Originally identified in men in wartime, it has been widely recognised as a common chronic condition in both sexes in civilian life. Although the symptoms may seem to appear after infections and various physical and psychological stresses, neurocirculatory asthenia is most often encountered as a familial disorder that is unrelated to these factors, although they may aggravate an existing tendency. Respiratory complaints (including breathlessness, with and without effort, and smothering sensations) are almost universal, and palpitation, chest discomfort, dizziness and faintness, and fatigue are common. The physical examination is normal. The aetiology is obscure but patients usually have a normal life span. Reassurance and measures to improve physical fitness are helpful. PMID:3314950

  19. 75 FR 52292 - Airworthiness Directives; Diamond Aircraft Industries GmbH Models DA 40 and DA 40F Airplanes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-25

    ... Directives; Diamond Aircraft Industries GmbH Models DA 40 and DA 40F Airplanes AGENCY: Federal Aviation...: We propose to adopt a new airworthiness directive (AD) for all Diamond Aircraft Industries GmbH... Federal holidays. For service information identified in this proposed AD, contact Diamond...

  20. Current Status of the Hot White Dwarf Luminosity Function and non-DA to DA Ratio from SDSS Data

    NASA Astrophysics Data System (ADS)

    Krzesinski, J.; Stachowski, G.

    2017-03-01

    Recent advances in the determination of the hot white dwarf (WD) luminosity function have reached the point where we have good agreement between the observed and theoretical DA and non-DA LFs for WDs from SDSS DR4. The main progress in calculation of the DA LF was obtained when the WD sample was restricted to only carbon-oxygen core WDs. However, one remaining effect which could influence the LF and the non-DA to DA ratio is the difference in visibility of helium- and hydrogen-atmosphere WDs in a magnitude-limited sample. In this work we explore this effect for the SDSS g filter as a function of effective temperature, as well as make an attempt to evaluate data relevant to the WD sample and LFs from more recent data releases of the SDSS.

  1. Effect of 14-kDa and 47-kDa protein molecules of age garlic extract on peritoneal macrophages.

    PubMed

    Daneshmandi, Saeed; Hajimoradi, Monire; Ahmadabad, Hasan Namdar; Hassan, Zuhair Mohammad; Roudbary, Maryam; Ghazanfari, Tooba

    2011-03-01

    Garlic (Allium sativum), traditionally being used as a spice worldwide, has different applications and is claimed to possess beneficial effects in several health ailments such as tumor and atherosclerosis. Garlic is also an immunomodulator and its different components are responsible for different properties. The present work aimed to assess the effect of protein fractions of garlic on peritoneal macrophages. 14-kDa and 47-kDa protein fractions of garlic were purified. Mice peritoneal macrophages were lavaged and cultured in a microtiter plate and exposed to different concentrations of garlic proteins. MTT assay was performed to evaluate the viability of macrophage. The amount of nitric oxide (NO) was detected in culture supernatants of macrophages by Griess reagent and furthermore, the cytotoxicity study of culture supernatants was carried out on WEHI-164 fibrosarcoma cell line as tumor necrosis factor-α bioassay. MTT assay results for both 14-kDa and 47-kDa protein fractions of stimulated macrophages were not significant (P > 0.05). Both 14-kDa and 47-kDa fractions significantly suppressed production of NO from macrophages (P = 0.007 and P = 0.003, respectively). Cytotoxicity of macrophages' supernatant on WEHI-164 fibrosarcoma cells was not affected by garlic protein fractions (P = 0.066 for 14-kDa and P = 0.085 for 47-kDa fractions). according to our finding, 14-kDa and 47-kDa fractions of aged garlic extract are able to suppress NO production from macrophages, which can be used as a biological advantage. These molecules had no cytotoxic effect on macrophages and do not increase tumoricidal property of macrophages.

  2. Beam position monitor system of DA{phi}NE

    SciTech Connect

    Stella, A.; Drago, A.; Ghigo, A.; Marcellini, F.; Milardi, C.; Sannibale, F.; Serio, M.; Vaccarezza, C.

    1998-12-10

    The DA{phi}NE beam position monitor (BPM) system consists of 150 monitors installed all along the machine. Design issues, calibration procedures, experimental results and performance of the system are described. The closed orbit in the main rings is extracted from the BPM signals through narrowband receivers (realized by Bergoz Precision Beam Instrumentation for DA{phi}NE), then acquired and processed by a real-time task based on four independent processors dealing with different machine areas. The data acquisition system is integrated in the DA{phi}NE control system and measures five complete orbits in a second. Implementation criteria, measurements and results are reported.

  3. Circumstellar features in hot DA white dwarfs

    NASA Astrophysics Data System (ADS)

    Bannister, N. P.; Barstow, M. A.; Holberg, J. B.; Bruhweiler, F. C.

    2003-05-01

    We present a phenomenological study of highly ionized, non-photospheric absorption features in high spectral resolution vacuum ultraviolet spectra of 23 hot DA white dwarfs. Prior to this study, four of the survey objects (Feige 24, REJ 0457-281, G191-B2B and REJ 1614-085) were known to possess these features. We find four new objects with multiple components in one or more of the principal resonance lines: REJ 1738+665, Ton 021, REJ 0558-373 and WD 2218+706. A fifth object, REJ 2156-546, also shows some evidence of multiple components, though further observations are required to confirm the detection. We discuss possible origins for these features including ionization of the local interstellar environment, the presence of material inside the gravitational well of the white dwarf, mass loss in a stellar wind and the existence of material in an ancient planetary nebula around the star. We propose ionization of the local interstellar medium as the origin of these features in G191-B2B and REJ 1738+665, and demonstrate the need for higher-resolution spectroscopy of the sample, to detect multiple interstellar medium velocity components and to identify circumstellar features that may lie close to the photospheric velocity.

  4. Phantom surfaces in da Vinci stereopsis.

    PubMed

    Wardle, Susan G; Gillam, Barbara J

    2013-02-08

    In binocular viewing of natural three-dimensional scenes, occlusion relationships between objects at different depths create regions of the background that are visible to only one eye. These monocular regions can support depth perception. There are two viewing conditions in which a monocular region can be on the nasal side of a binocular surface--(a) when a background surface is viewed through an aperture and (b) when a region is camouflaged against the background in one eye's view. We created stimuli with a monocular region using complex textures in which camouflage was not possible, and for which there was no physical aperture. For these stimuli, observers perceived a strong phantom contour in near depth at the edge of the monocular region, with the monocular texture perceived behind at the depth of the binocular surface. Depth-matching with a probe showed that the depth of the phantom occluding surface was as precise as for stimuli with regular binocular disparity. Monocular regions of texture on the opposite (temporal) side of the binocular surface were perceived behind, as predicted by occlusion geometry, and there was no phantom surface. We discuss the implications for models of da Vinci stereopsis and stereoscopic edge processing, and consider the involvement of a form of Panum's limiting case. We conclude that the visual system uses a combination of occlusion geometry and complex matching to precisely locate edges in depth that lack a luminance contour.

  5. Der Telemanipulator daVinci als mechanisches Trackingsystem

    NASA Astrophysics Data System (ADS)

    Käst, Johannes; Neuhaus, Jochen; Nickel, Felix; Kenngott, Hannes; Engel, Markus; Short, Elaine; Reiter, Michael; Meinzer, Hans-Peter; Maier-Hein, Lena

    Der Telemanipulator daVinci (Intuitive Surgical, Sunnyvale, Kalifornien) ist ein M aster-Slave System für roboterassistierte minimalinvasive Chirurgie. Da er über integrierte Gelenksensoren verfügt, kann er unter Verwendung der daVinci-API als mechanisches Trackingsystem verwendet werden. In dieser Arbeit evaluieren wir die Präzision und Genauigkeit eines daVinci mit Hilfe eines Genauigkeitsphantoms mit bekannten Maßen. Der ermittelte Positionierungsfehler liegt in der Größenordnung von 6 mm und ist somit für einen Großteil der medizinischen Fragestellungen zu hoch. Zur Reduktion des Fehlers schlagen wir daher eine Kalibrierung der Gelenksensoren vor.

  6. 40 CFR 60.48Da - Compliance provisions.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... successive boiler operating days is completed within 60 days after achieving the maximum production rate at... determine compliance by using the CEMS specified under § 60.49Da for measuring NOX and oxygen (O2) (or...

  7. 40 CFR 60.48Da - Compliance provisions.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... after achieving the maximum production rate at which the affected facility will be operated, but not... elect to determine compliance by using the CEMS specified under § 60.49Da for measuring NOX and oxygen...

  8. 40 CFR 60.51Da - Reporting requirements.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... Generating Units § 60.51Da Reporting requirements. (a) For SO2, NOX, PM, and NOX plus CO emissions, the... emission rates (so) and inlet emission rates (si) as applicable. (3) The lower confidence limit for the...

  9. 40 CFR 60.51Da - Reporting requirements.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... Generating Units § 60.51Da Reporting requirements. (a) For SO2, NOX, PM, and NOX plus CO emissions, the... emission rates (so) and inlet emission rates (si) as applicable. (3) The lower confidence limit for the...

  10. ECTA/DaSy Framework Self-Assessment Comparison Tool

    ERIC Educational Resources Information Center

    Center for IDEA Early Childhood Data Systems (DaSy), 2016

    2016-01-01

    The Self-Assessment Comparison (SAC) Tool is for state Part C and Section 619/Preschool programs to use to assess changes in the implementation of one or more components of the ECTA System Framework and/or subcomponenets of the DaSy Data System Framework. It is a companion to the ECTA/DaSy Framework Self-Assessment. Key features of the SAC are…

  11. Tree branching: Leonardo da Vinci's rule versus biomechanical models.

    PubMed

    Minamino, Ryoko; Tateno, Masaki

    2014-01-01

    This study examined Leonardo da Vinci's rule (i.e., the sum of the cross-sectional area of all tree branches above a branching point at any height is equal to the cross-sectional area of the trunk or the branch immediately below the branching point) using simulations based on two biomechanical models: the uniform stress and elastic similarity models. Model calculations of the daughter/mother ratio (i.e., the ratio of the total cross-sectional area of the daughter branches to the cross-sectional area of the mother branch at the branching point) showed that both biomechanical models agreed with da Vinci's rule when the branching angles of daughter branches and the weights of lateral daughter branches were small; however, the models deviated from da Vinci's rule as the weights and/or the branching angles of lateral daughter branches increased. The calculated values of the two models were largely similar but differed in some ways. Field measurements of Fagus crenata and Abies homolepis also fit this trend, wherein models deviated from da Vinci's rule with increasing relative weights of lateral daughter branches. However, this deviation was small for a branching pattern in nature, where empirical measurements were taken under realistic measurement conditions; thus, da Vinci's rule did not critically contradict the biomechanical models in the case of real branching patterns, though the model calculations described the contradiction between da Vinci's rule and the biomechanical models. The field data for Fagus crenata fit the uniform stress model best, indicating that stress uniformity is the key constraint of branch morphology in Fagus crenata rather than elastic similarity or da Vinci's rule. On the other hand, mechanical constraints are not necessarily significant in the morphology of Abies homolepis branches, depending on the number of daughter branches. Rather, these branches were often in agreement with da Vinci's rule.

  12. ECTA/DaSy Framework Self-Assessment Comparison Tool

    ERIC Educational Resources Information Center

    Center for IDEA Early Childhood Data Systems (DaSy), 2016

    2016-01-01

    The Self-Assessment Comparison (SAC) Tool is for state Part C and Section 619/Preschool programs to use to assess changes in the implementation of one or more components of the ECTA System Framework and/or subcomponenets of the DaSy Data System Framework. It is a companion to the ECTA/DaSy Framework Self-Assessment. Key features of the SAC are…

  13. Migration of ATLAS PanDA to CERN

    NASA Astrophysics Data System (ADS)

    Stewart, Graeme Andrew; Klimentov, Alexei; Koblitz, Birger; Lamanna, Massimo; Maeno, Tadashi; Nevski, Pavel; Nowak, Marcin; Emanuel De Castro Faria Salgado, Pedro; Wenaus, Torre

    2010-04-01

    The ATLAS Production and Distributed Analysis System (PanDA) is a key component of the ATLAS distributed computing infrastructure. All ATLAS production jobs, and a substantial amount of user and group analysis jobs, pass through the PanDA system, which manages their execution on the grid. PanDA also plays a key role in production task definition and the data set replication request system. PanDA has recently been migrated from Brookhaven National Laboratory (BNL) to the European Organization for Nuclear Research (CERN), a process we describe here. We discuss how the new infrastructure for PanDA, which relies heavily on services provided by CERN IT, was introduced in order to make the service as reliable as possible and to allow it to be scaled to ATLAS's increasing need for distributed computing. The migration involved changing the backend database for PanDA from MySQL to Oracle, which impacted upon the database schemas. The process by which the client code was optimised for the new database backend is discussed. We describe the procedure by which the new database infrastructure was tested and commissioned for production use. Operations during the migration had to be planned carefully to minimise disruption to ongoing ATLAS offline computing. All parts of the migration were fully tested before commissioning the new infrastructure and the gradual migration of computing resources to the new system allowed any problems of scaling to be addressed.

  14. DaTo: an atlas of biological databases and tools.

    PubMed

    Li, Qilin; Zhou, Yincong; Jiao, Yingmin; Zhang, Zhao; Bai, Lin; Tong, Li; Yang, Xiong; Sommer, Björn; Hofestädt, Ralf; Chen, Ming

    2016-12-18

    This work presents DaTo, a semi-automatically generated world atlas of biological databases and tools. It extracts raw information from all PubMed articles which contain exact URLs in their abstract section, followed by a manual curation of the abstract and the URL accessibility. DaTo features a user-friendly query interface, providing extensible URL-related annotations, such as the status, the location and the country of the URL. A graphical interaction network browser has also been integrated into the DaTo web interface to facilitate exploration of the relationship between different tools and databases with respect to their ontology-based semantic similarity. Using DaTo, the geographical locations, the health statuses, as well as the journal associations were evaluated with respect to the historical development of bioinformatics tools and databases over the last 20 years. We hope it will inspire the biological community to gain a systematic insight into bioinformatics resources. DaTo is accessible via http://bis.zju.edu.cn/DaTo/.

  15. Mycobacterium tuberculosis 19-kDa lipoprotein promotes neutrophil activation.

    PubMed

    Neufert, C; Pai, R K; Noss, E H; Berger, M; Boom, W H; Harding, C V

    2001-08-01

    Certain microbial substances, e.g., LPS, can activate neutrophils or prime them to enhance their response to other activating agents, e.g., fMLP. We investigated the role of the Mycobacterium tuberculosis (MTB) 19-kDa lipoprotein in activation of human neutrophils. MTB 19-kDa lipoprotein initiated phenotypic changes characteristic of neutrophil activation, including down-regulation of CD62 ligand (L-selectin) and up-regulation of CD35 (CR1) and CD11b/CD18 (CR3, Mac-1). In addition, exposure of neutrophils to MTB 19-kDa lipoprotein enhanced the subsequent oxidative burst in response to fMLP as assessed by oxidation of dihydrorhodamine 123 (determined by flow cytometry). LPS also produced these effects with similar kinetics, but an oligodeoxynucleotide containing a CpG motif failed to induce any priming or activation response. Although the effects of LPS required the presence of serum, neutrophil activation by MTB 19-kDa lipoprotein occurred independently of serum factors, suggesting the involvement of different receptors and signaling mechanisms for LPS and MTB 19-kDa lipoprotein. Thus, MTB 19-kDa lipoprotein serves as a pathogen-associated molecular pattern that promotes neutrophil priming and activation.

  16. The Ubiquitin Receptor DA1 Interacts with the E3 Ubiquitin Ligase DA2 to Regulate Seed and Organ Size in Arabidopsis[C][W

    PubMed Central

    Xia, Tian; Li, Na; Dumenil, Jack; Li, Jie; Kamenski, Andrei; Bevan, Michael W.; Gao, Fan; Li, Yunhai

    2013-01-01

    Seed size in higher plants is determined by the coordinated growth of the embryo, endosperm, and maternal tissue. Several factors that act maternally to regulate seed size have been identified, such as AUXIN RESPONSE FACTOR2, APETALA2, KLUH, and DA1, but the genetic and molecular mechanisms of these factors in seed size control are almost totally unknown. We previously demonstrated that the ubiquitin receptor DA1 acts synergistically with the E3 ubiquitin ligase ENHANCER1 OF DA1 (EOD1)/BIG BROTHER to regulate the final size of seeds in Arabidopsis thaliana. Here, we describe another RING-type protein with E3 ubiquitin ligase activity, encoded by DA2, which regulates seed size by restricting cell proliferation in the maternal integuments of developing seeds. The da2-1 mutant forms large seeds, while overexpression of DA2 decreases seed size of wild-type plants. Overexpression of rice (Oryza sativa) GRAIN WIDTH AND WEIGHT2, a homolog of DA2, restricts seed growth in Arabidopsis. Genetic analyses show that DA2 functions synergistically with DA1 to regulate seed size, but does so independently of EOD1. Further results reveal that DA2 interacts physically with DA1 in vitro and in vivo. Therefore, our findings define the genetic and molecular mechanisms of three ubiquitin-related proteins DA1, DA2, and EOD1 in seed size control and indicate that they are promising targets for crop improvement. PMID:24045020

  17. The ubiquitin receptor DA1 interacts with the E3 ubiquitin ligase DA2 to regulate seed and organ size in Arabidopsis.

    PubMed

    Xia, Tian; Li, Na; Dumenil, Jack; Li, Jie; Kamenski, Andrei; Bevan, Michael W; Gao, Fan; Li, Yunhai

    2013-09-01

    Seed size in higher plants is determined by the coordinated growth of the embryo, endosperm, and maternal tissue. Several factors that act maternally to regulate seed size have been identified, such as auxin response factor2, apetala2, KLUH, and DA1, but the genetic and molecular mechanisms of these factors in seed size control are almost totally unknown. We previously demonstrated that the ubiquitin receptor DA1 acts synergistically with the E3 ubiquitin ligase enhancer1 OF DA1 (EOD1)/big brother to regulate the final size of seeds in Arabidopsis thaliana. Here, we describe another RING-type protein with E3 ubiquitin ligase activity, encoded by DA2, which regulates seed size by restricting cell proliferation in the maternal integuments of developing seeds. The da2-1 mutant forms large seeds, while overexpression of DA2 decreases seed size of wild-type plants. Overexpression of rice (Oryza sativa) grain width and weight2, a homolog of DA2, restricts seed growth in Arabidopsis. Genetic analyses show that DA2 functions synergistically with DA1 to regulate seed size, but does so independently of EOD1. Further results reveal that DA2 interacts physically with DA1 in vitro and in vivo. Therefore, our findings define the genetic and molecular mechanisms of three ubiquitin-related proteins DA1, DA2, and EOD1 in seed size control and indicate that they are promising targets for crop improvement.

  18. Mevalocidin: a novel, phloem mobile phytotoxin from Fusarium DA056446 and Rosellinia DA092917.

    PubMed

    Gerwick, B Clifford; Brewster, William K; Deboer, Gerrit J; Fields, Steve C; Graupner, Paul R; Hahn, Donald R; Pearce, Cedric J; Schmitzer, Paul R; Webster, Jeffery D

    2013-02-01

    A multiyear effort to identify new natural products was built on a hypothesis that both phytotoxins from plant pathogens and antimicrobial compounds might demonstrate herbicidal activity. The discovery of one such compound, mevalocidin, is described in the current report. Mevalocidin was discovered from static cultures of two unrelated fungal isolates designated Rosellinia DA092917 and Fusarium DA056446. The chemical structure was confirmed by independent synthesis. Mevalocidin demonstrated broad spectrum post-emergence activity on grasses and broadleaves and produced a unique set of visual symptoms on treated plants suggesting a novel mode of action. Mevalocidin was rapidly absorbed in a representative grass and broadleaf plant. Translocation occurred from the treated leaf to other plant parts including roots confirming phloem as well as xylem mobility. By 24 hr after application, over 20 % had been redistributed through-out the plant. Mevalocidin is a unique phytotoxin based on its chemistry, with the uncommon attribute of demonstrating both xylem and phloem mobility in grass and broadleaf plants.

  19. Secretion of 10-kDa and 12-kDa thioredoxin species from blood monocytes and transformed leukocytes.

    PubMed

    Sahaf, B; Rosén, A

    2000-01-01

    Thioredoxins (TRX) are ubiquitous, small redox-active proteins with multiple functions, including antioxidant, cytoprotective, and chemoattractant activities. In addition to a 12-kDa intracellular form, extracellular 10-kDa and 12-kDa TRX have been defined. The biological activities of the 10-kDa TRX were previously measured as eosinophil cytotoxicity enhancing activity or B-cell stimulatory activity. Cytotrophoblastic cell lines also release a 10-kDa TRX form. To study the biological role of 10-kDa TRX, we established two highly sensitive enzyme-linked immuno-spot assays (ELISPOT), which detect secreted truncated 10-kDa and full-length 12-kDa TRX at the single cell level. TRX secretion was investigated in several cell lines including the T-helper cell hybridoma MP6, the Jurkat T-cell leukemia, the U-937 myelomonocytic leukemia, and the 3B6, EBV-transformed, lymphoblastoid B-cell line. The highest number of secreting cells was found in 3B6 cultures, median = 34 (quartiles, 27-39) per well (10(5) cells). Peripheral blood monocytes isolated from healthy donors secreted significantly more TRX after stimulation with ionomycin, phorbol myristate acetate (PMA), fMLP, and lipopolysaccharide (LPS), compared to unstimulated cells. Oxidative stress induced by thioloxidant diamide also induced the secretion of both truncated and full-length TRX measured in ELISPOT (p = 0.047 and p = 0.031, respectively). The biological activity of the truncated and full-length forms was tested in a cell migration assay. Truncated TRX was devoid of protein disulfide reductase activity, but retained strong chemoattractant activity for human monocytes, in the same range as full-length TRX, as previously reported (Bertini et al., 1999).

  20. PKC phosphorylates residues in the N-terminal of the DA transporter to regulate amphetamine-induced DA efflux.

    PubMed

    Wang, Qiang; Bubula, Nancy; Brown, Jason; Wang, Yunliang; Kondev, Veronika; Vezina, Paul

    2016-05-27

    The DA transporter (DAT), a phosphoprotein, controls extracellular dopamine (DA) levels in the central nervous system through transport or reverse transport (efflux). Multiple lines of evidence support the claim that PKC significantly contributes to amphetamine-induced DA efflux. Other signaling pathways, involving CaMKII and ERK, have also been shown to regulate DAT mediated efflux. Here we assessed the contribution of putative PKC residues (S4, S7, S13) in the N-terminal of the DAT to amphetamine-induced DA efflux by transfecting DATs containing different serine to alanine (S-A) point mutations into DA pre-loaded HEK-293 cells and incubating these cells in amphetamine (2μM). The effects of a S-A mutation at the non-PKC residue S12 and a threonine to alanine (T-A) mutation at the ERK T53 residue were also assessed for comparison. WT-DATs were used as controls. In an initial experiment, we confirmed that inhibiting PKC with Go6976 (130nM) significantly reduced amphetamine-induced DA efflux. In subsequent experiments, cells transfected with the S4A, S12A, S13A, T53A and S4,7,13A mutants showed a reduction in amphetamine-induced DA efflux similar to that observed with Go6976. Interestingly, cells transfected with the S7A mutant, identified by some as a PKC-PKA residue, showed unperturbed WT-DAT levels of amphetamine-induced DA efflux. These results indicate that phosphorylation by PKC of select residues in the DAT N-terminal can regulate amphetamine-induced efflux. PKC can act either independently or in concert with other kinases such as ERK to produce this effect.

  1. Purification and characterization of 94kDa and 80kDa forms of the muscarinic cholinergic receptor

    SciTech Connect

    Fracek, S.P. Jr.; Venter, J.C.; Kerlavage, A.R.

    1986-05-01

    Two molecular forms of the muscarinic cholinergic receptor have been consistently observed in a variety of species, albeit in variable amounts. Proteins which are specifically labeled by (/sup 3/H)propylbenzilylcholine mustard ((/sup 3/H)PrBCM) were observed at 94kDa and 80kDa upon SDS-PAGE of membrane proteins prepared from brains and hearts of trout, frog, turtle, chicken, rat, and pig. They have developed a purification procedure which yields each of these proteins in a homogeneous form suitable for structural analysis. The four step procedure involves affinity chromatography on 3-(2'-aminobenzhydryloxy)tropane-sepharose, concentration on hydroxylapatite, preparative SDS-PAGE and extraction of individual bands from the gel. Limited tryptic digestion of purified (/sup 3/H)PrBCM-labeled porcine atrial muscarinic receptor yields (/sup 3/H)-labeled fragments of 75, 65, 52, 40, 35, 30, 25, and 20kDa, in close agreement with results of analogous digestions of muscarinic receptor from other species and tissues. Complete tryptic digestion and subsequent mapping by reverse-phase HPLC yields very similar profiles for (/sup 125/I)-labeled 94kDa and 80kDA receptor forms. Most peaks which elute in the hydrophobic region of the profile overlap for the two proteins while the 94kDa protein contains several additional peaks of apparent low hydrophobicity.

  2. 76 FR 12627 - Airworthiness Directives; Diamond Aircraft Industries GmbH Models DA 42, DA 42 NG, and DA 42 M-NG...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-08

    ... Federal Aviation Administration 14 CFR Part 39 RIN 2120-AA64 Airworthiness Directives; Diamond Aircraft... on Diamond aeroplanes, the majority of which were DA 40. In additional, at least 18 doors have been replaced because of damage found on the hinge. Diamond Aircraft Industries conducted analyses...

  3. A cosmologia no ensino da geografia

    NASA Astrophysics Data System (ADS)

    Santos, S. C.; Chiaradia, A. P. M.

    2003-08-01

    O principal objetivo deste trabalho é auxiliar o professor de Geografia em sala de aula no ensino de tópicos relacionados com a Cosmologia. A idéia deste trabalho surgiu quando foi constatado que o professor de Geografia tem dificuldades de ensinar este tópico. Esta constatação foi feita por uma das autoras ao lecionar este tópico no ensino fundamental e em discussões com outros professores de Geografia. Da mesma maneira que ocorria desde os tempos mais antigos, os alunos têm muito interesse em conhecer os fenômenos que ocorrem no Cosmo, porém os livros didáticos de Geografia utilizados em sala de aula não são ricos em informações sobre este assunto. Assim, o professor de Geografia tem poucas informações para discutir este assunto em sala de aula e não dá a devida importância para este tópico. Então, foi desenvolvido um material de apoio para professores de Geografia sobre a origem do Universo, sua evolução e seu possível futuro evolutivo segundo as mais recentes teorias, com base em perguntas feitas pelos alunos de ensino fundamental e as informações trazidas nos livros didáticos Não cabe a este material inovar e tão pouco trazer uma metodologia de ensino de Cosmologia. Neste material o professor de Geografia pode encontrará um banco de informações, que constitui no estabelecimento de conceitos, teorias e hipóteses, sobre a Cosmologia, em linguagem simples e de fácil entendimento. Para desenvolvê-lo, foram feitas pesquisas não exaustivas em livros e revistas científicas, compilação e discussão em forma cronológica das teorias aceitas sobre modelos cosmológicos. Portanto, este material será apresentado neste trabalho.

  4. Overview of ATLAS PanDA Workload Management

    SciTech Connect

    Maeno T.; De K.; Wenaus T.; Nilsson P.; Stewart G. A.; Walker R.; Stradling A.; Caballero J.; Potekhin M.; Smith D.

    2011-01-01

    The Production and Distributed Analysis System (PanDA) plays a key role in the ATLAS distributed computing infrastructure. All ATLAS Monte-Carlo simulation and data reprocessing jobs pass through the PanDA system. We will describe how PanDA manages job execution on the grid using dynamic resource estimation and data replication together with intelligent brokerage in order to meet the scaling and automation requirements of ATLAS distributed computing. PanDA is also the primary ATLAS system for processing user and group analysis jobs, bringing further requirements for quick, flexible adaptation to the rapidly evolving analysis use cases of the early datataking phase, in addition to the high reliability, robustness and usability needed to provide efficient and transparent utilization of the grid for analysis users. We will describe how PanDA meets ATLAS requirements, the evolution of the system in light of operational experience, how the system has performed during the first LHC data-taking phase and plans for the future.

  5. Overview of ATLAS PanDA Workload Management

    NASA Astrophysics Data System (ADS)

    Maeno, T.; De, K.; Wenaus, T.; Nilsson, P.; Stewart, G. A.; Walker, R.; Stradling, A.; Caballero, J.; Potekhin, M.; Smith, D.; ATLAS Collaboration

    2011-12-01

    The Production and Distributed Analysis System (PanDA) plays a key role in the ATLAS distributed computing infrastructure. All ATLAS Monte-Carlo simulation and data reprocessing jobs pass through the PanDA system. We will describe how PanDA manages job execution on the grid using dynamic resource estimation and data replication together with intelligent brokerage in order to meet the scaling and automation requirements of ATLAS distributed computing. PanDA is also the primary ATLAS system for processing user and group analysis jobs, bringing further requirements for quick, flexible adaptation to the rapidly evolving analysis use cases of the early datataking phase, in addition to the high reliability, robustness and usability needed to provide efficient and transparent utilization of the grid for analysis users. We will describe how PanDA meets ATLAS requirements, the evolution of the system in light of operational experience, how the system has performed during the first LHC data-taking phase and plans for the future.

  6. The ATLAS PanDA Monitoring System and its Evolution

    NASA Astrophysics Data System (ADS)

    Klimentov, A.; Nevski, P.; Potekhin, M.; Wenaus, T.

    2011-12-01

    The PanDA (Production and Distributed Analysis) Workload Management System is used for ATLAS distributed production and analysis worldwide. The needs of ATLAS global computing imposed challenging requirements on the design of PanDA in areas such as scalability, robustness, automation, diagnostics, and usability for both production shifters and analysis users. Through a system-wide job database, the PanDA monitor provides a comprehensive and coherent view of the system and job execution, from high level summaries to detailed drill-down job diagnostics. It is (like the rest of PanDA) an Apache-based Python application backed by Oracle. The presentation layer is HTML code generated on the fly in the Python application which is also responsible for managing database queries. However, this approach is lacking in user interface flexibility, simplicity of communication with external systems, and ease of maintenance. A decision was therefore made to migrate the PanDA monitor server to Django Web Application Framework and apply JSON/AJAX technology in the browser front end. This allows us to greatly reduce the amount of application code, separate data preparation from presentation, leverage open source for tools such as authentication and authorization mechanisms, and provide a richer and more dynamic user experience. We describe our approach, design and initial experience with the migration process.

  7. Geomagnetic observations on tristan da cunha, south atlantic ocean

    USGS Publications Warehouse

    Matzka, J.; Olsen, N.; Maule, C.F.; Pedersen, L.W.; Berarducci, A.M.; Macmillan, S.

    2009-01-01

    Few geomagnetic ground observations exist of the Earth's strongest core field anomaly, the South Atlantic Anomaly (SAA). The geomagnetic repeat station on the island Tristan da Cunha, located half-way between South Africa and South America at 37?? 05' S, 12?? 18' W, is therefore of crucial importance. We have conducted several sets of repeat station measurements during magnetically quiet conditions (Kp 2o or less) in 2004. The procedures are described and the results are compared to those from earlier campaigns and to the predictions of various global field models. Features of the local crustal bias field and the solar quiet daily variation are discussed. We also evaluate the benefit of continuous magnetic field recordings from Tristan da Cunha, and argue that such a data set is a very valuable addition to geomagnetic satellite data. Recently, funds were set up to establish and operate a magnetometer station on Tristan da Cunha during the Swarm magnetic satellite mission (2011-2014).

  8. 1991 DA: An asteroid in a bizarre orbit

    NASA Technical Reports Server (NTRS)

    Steel, Duncan; Mcnaught, Robert H.; Asher, David

    1992-01-01

    Asteroidal object 1991 DA has an orbit of high inclination, crossing the planets from Mars to Uranus. This is unique for an asteroid, but not unusual for a comet of the Halley-type: it therefore seems likely that 1991 DA is an extinct or dormant comet. Previous CCD imaging has shown no indication of a coma; spectroscopic observations of 1991 DA which lack any evidence of strong comet-like emissions are reported. Numerical integrations of the orbit of this object were performed which show that is has been remarkably stable for the past approximately 20,000 yr, but chaotic before that. This may allow a new estimate to be made of the physical lifetimes of comets.

  9. Situação da Mulher na Astronomia Brasileira

    NASA Astrophysics Data System (ADS)

    Silva, Adriana V. R.

    2007-07-01

    O conteúdo desse texto surgiu de uma apresentação de mesmo título que fiz na XXXI Reunião Anual da Sociedade Astronômica Brasileira (SAB) em 2005. Esse tema foi inspirado originalmente pela minha participação no "2nd UIPAP International Conference on Women in Physics" realizado entre 23 e 25 de maio de 2005 no Rio de Janeiro. Essa é uma conferência internacional que acontece de três em três anos, sendo que a primeira ocorreu em 2002 na cidade de Paris, França. Participei dessa conferência como membro da delegação da Sociedade Brasileira de Física e um dos trabalhos que apresentei versava sobre a situação das mulheres na Astronomia brasileira, cujos resultados principais discorro a seguir. A situação das astrônomas, baseada nos dados dos sócios da SAB coletados no final de 2004, é comparada com a das físicas brasileiras e também com as nossas colegas americanas. Os dados identificam ainda uma maior evasão da carreira por parte das mulheres do que os homens. Alguns dos possíveis motivos da evasão são discutidos, como o desejo de constituir família e/ou isolamento. Resultados um tanto preocupantes com relação à distribuição de bolsas de produtividade do CNPq também são apresentados. As principais discussões e estratégias recomendadas nesse congresso são mencionadas de forma resumida ao final.

  10. The future of PanDA in ATLAS distributed computing

    NASA Astrophysics Data System (ADS)

    De, K.; Klimentov, A.; Maeno, T.; Nilsson, P.; Oleynik, D.; Panitkin, S.; Petrosyan, A.; Schovancova, J.; Vaniachine, A.; Wenaus, T.

    2015-12-01

    Experiments at the Large Hadron Collider (LHC) face unprecedented computing challenges. Heterogeneous resources are distributed worldwide at hundreds of sites, thousands of physicists analyse the data remotely, the volume of processed data is beyond the exabyte scale, while data processing requires more than a few billion hours of computing usage per year. The PanDA (Production and Distributed Analysis) system was developed to meet the scale and complexity of LHC distributed computing for the ATLAS experiment. In the process, the old batch job paradigm of locally managed computing in HEP was discarded in favour of a far more automated, flexible and scalable model. The success of PanDA in ATLAS is leading to widespread adoption and testing by other experiments. PanDA is the first exascale workload management system in HEP, already operating at more than a million computing jobs per day, and processing over an exabyte of data in 2013. There are many new challenges that PanDA will face in the near future, in addition to new challenges of scale, heterogeneity and increasing user base. PanDA will need to handle rapidly changing computing infrastructure, will require factorization of code for easier deployment, will need to incorporate additional information sources including network metrics in decision making, be able to control network circuits, handle dynamically sized workload processing, provide improved visualization, and face many other challenges. In this talk we will focus on the new features, planned or recently implemented, that are relevant to the next decade of distributed computing workload management using PanDA.

  11. The role of transparency in da Vinci stereopsis.

    PubMed

    Zannoli, Marina; Mamassian, Pascal

    2011-10-15

    The majority of natural scenes contains zones that are visible to one eye only. Past studies have shown that these monocular regions can be seen at a precise depth even though there are no binocular disparities that uniquely constrain their locations in depth. In the so-called da Vinci stereopsis configuration, the monocular region is a vertical line placed next to a binocular rectangular occluder. The opacity of the occluder has been mentioned to be a necessary condition to obtain da Vinci stereopsis. However, this opacity constraint has never been empirically tested. In the present study, we tested whether da Vinci stereopsis and perceptual transparency can interact using a classical da Vinci configuration in which the opacity of the occluder varied. We used two different monocular objects: a line and a disk. We found no effect of the opacity of the occluder on the perceived depth of the monocular object. A careful analysis of the distribution of perceived depth revealed that the monocular object was perceived at a depth that increased with the distance between the object and the occluder. The analysis of the skewness of the distributions was not consistent with a double fusion explanation, favoring an implication of occlusion geometry in da Vinci stereopsis. A simple model that includes the geometry of the scene could account for the results. In summary, the mechanism responsible to locate monocular regions in depth is not sensitive to the material properties of objects, suggesting that da Vinci stereopsis is solved at relatively early stages of disparity processing. Copyright © 2011 Elsevier Ltd. All rights reserved.

  12. The DaCHS Multi-protocol VO Server

    NASA Astrophysics Data System (ADS)

    Demleitner, M.

    2014-05-01

    GAVO's Data Center Helper Suite (DaCHS) is a suite of tools for publishing data to the Virtual Observatory. It implements all major VO protocols (SCS, SIAP, SSAP, TAP, OAI-PMH). The integrated management and ingestion component allow defining metadata, structure, and services once and re-use the definition throughout the publication cycle from initial metadata aquisition to registry record generation. It has been driving GAVO's data center since 2008 and is now deployed in multiple locations around the globe. This poster briefly describes the design of the system as well as a bird's-eye view of data publishing with DaCHS.

  13. Using DA White Dwarfs to Calibrate Synthetic Photometry

    NASA Astrophysics Data System (ADS)

    Holberg, J. B.

    2007-04-01

    Four widely used photometric systems, namely the Johnson-Kron-Cousins UBVRI, the Strömgren uvby, the 2MASS JHKs and the Sloan Digital Sky Survey ugriz systems have been directly compared with the HST absolute photometric scale of Bohlin & Gilliland (2004). These comparisons are subsequently used to construct a large grid of accurate synthetic magnitudes for DA white dwarfs. This grid is, in turn, critically evaluated with respect to the observed photometry from substantial samples of actual white dwarfs. The advantages of DA white dwarfs as photometric stars are emphasized, and the prospects for extending the use of these stars into the near infrared are highlighted.

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

    DOEpatents

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

    1980-06-06

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

  15. Support arrangements for core modules of nuclear reactors. [PWR

    DOEpatents

    Bollinger, L.R.

    1983-11-03

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

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

    SciTech Connect

    Bailey, W.J.; Tokar, M.

    1982-12-01

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

  17. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    DTIC Science & Technology

    2013-06-01

    the probability of cladding failure. Radiation damage to Zr causes degradation of its mechanical properties . Thermal stress, vibration, fatigue...damaging to Zr’s mechanical properties even in the absence of radiation, however radiation increases the severity of these effects —and the effect of...tends to lead to degradation of mechanical properties via formation of elements not present in the as manufactured alloy , helium production, and

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

    SciTech Connect

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

    1985-01-01

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

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

    SciTech Connect

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

    1982-07-01

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

  20. Coolant monitoring apparatus for nuclear reactors. [PWR; BWR

    SciTech Connect

    Tokarz, R.D.

    1981-08-06

    A system for monitoring coolant conditions within a pressurized vessel is described. A length of tubing extends outward from the vessel from an open end containing a first line restriction at the location to be monitored. The flowing fluid is cooled and condensed before passing through a second line restriction. Measurement of pressure drop at the second line restriction gives an indication of fluid condition at the first line restriction. Multiple lengths of tubing with open ends at incremental elevations can measure coolant level within the vessel.

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

    SciTech Connect

    Berry, D. L.

    1980-05-01

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

  2. 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

  3. Sensitivity studies of seismic risk models. Final report. [PWR

    SciTech Connect

    Ravindra, M.K.; Banon, H.; Sues, R.H.; Thrasher, R.D.

    1984-06-01

    Recent PRA studies have used the Zion Method for estimating the seismic risks of nuclear power plants. During the course of these studies and in the subsequent regulatory and peer reviews, several questions were raised regarding the sensitivity of risk estimates. The present report has addressed these concerns with the objective of deriving generic conclusions. Sensitivity of seismically-induced severe core damage frequencies to different modeling assumptions was investigated using the Zion and Indian Point Unit 2 probabilistic safety studies as base cases. These included the effects of peak acceleration truncation, fragility modeling, dependence between component failures, and the significance of gross design and construction errors.

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

    SciTech Connect

    Amos, C N; Schrock, V E

    1983-09-01

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

  5. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  6. Artificial neural network prediction of PWR critical boron concentration

    SciTech Connect

    Wendt, S.E.; Maldonado, G.I.; Bartlett, E.B.

    1995-12-31

    The direct calculation of core parameters such as k{sub eff} and pin power peaks for light water reactors is ordinarily accomplished by numerically solving the neutron diffusion equation. Despite the rapid advances in computer architecture and algorithm development, further calculational speedups are always in great demand. One example of such an application is nuclear fuel management optimization, where the core attributes of tens of thousands of loading pattern candidates must typically be evaluated over the fuel cycle. If an artificial neural network (ANN) could be trained to accurately model the neutronic behavior of a core, a substantial time savings could be realized in the prediction of core parameters. Such an ANN could be exploited in at least two ways: 1. The a priori training of an ANN model could be tailored to address a specific plant and its corresponding licensing core neutronics software. 2. Once trained to within acceptable accuracy guidelines, an ANN model could provide the luxury of nearly instantaneous evaluations of core parameters. Recent publications by Kim et al. on core parameter prediction via ANNs have revealed a variety of promising results, which, in part, motivated our studies. Kim proved that a solution was possible; however, the large size and complexity of such a model can lead to memorization instead of generalization of the problem`s solution. Thus, the purpose of this work was to show that a much smaller ANN could predict a global core parameter such as the critical boron concentration over a wide range of training and validation data. The successful modeling of this problem with a much smaller ANN is considered to be a significant highlight of this study. This work employed Studsvik of America`s SOA1 Database, which proved to be useful for ANN training and validation.

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

    SciTech Connect

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

    1984-01-01

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

  8. PWR steam generator chemical cleaning. Phase II. Final report

    SciTech Connect

    Not Available

    1980-01-01

    Two techniques believed capable of chemically dissolving the corrosion products in the annuli between tubes and support plates were developed in laboratory work in Phase I of this project and were pilot tested in Indian Point Unit No. 1 steam generators. In Phase II, one of the techniques was shown to be inadequate on an actual sample taken from an Indian Point Unit No. 2 steam generator. The other technique was modified slightly, and it was demonstrated that the tube/support plate annulus could be chemically cleaned effectively.

  9. Electropolishing qualification program for PWR steam generator divider plates

    SciTech Connect

    Spalaris, C.N. )

    1990-09-01

    A program was conducted to establish electropolishing parameters for Inconel 600 plate and Inco 182 weld metal. Test parameters were chosen so as to define margins in the principal process variables, as a prerequisite for applying electropolishing to reactor components. The test program and evaluation of the results obtained are included in this report. 12 refs., 35 figs., 5 tabs.

  10. Need and procedure for calibration of fuel rod simulators. [PWR

    SciTech Connect

    Dabbs, R.D.; Ott, L.J.

    1980-01-01

    An experimental thermocouple calibration procedure and four-part calibration program, ORTCAL (ORNL Thermocouple Calibration), have been developed to supply FRS performance information to the inverse heat conduction model and program ORINC. Case studies have shown that failure to fully classify FRSs with regard to component physical properties, gaps, etc., can result in severe errors during inverse calculations of the driving potential at the surface of the FRS (..delta..T), the spatial and temporal history of the heat flow within the FRS, and the surface heat flux.

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

    SciTech Connect

    Klamerus, L.J.

    1980-01-01

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

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

    SciTech Connect

    Berry, D.L.

    1983-01-01

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

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

    SciTech Connect

    Ishii, M.; Kataoka, I.

    1982-01-01

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

  14. 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.

  15. 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.

  16. PWR internal flow modeling with fuel assemblies details

    SciTech Connect

    Popov, E.; Yan, J.; Karoutas, Z.; Gehin, J.; Brewster, R.; Baglietto, E.

    2012-07-01

    This study is an example of a massive parallel computing of the coolant flow in a nuclear reactor. It resolves the flow velocities in each assembly on pin level and predicts the flow distribution in complex geometries such as the lower and upper reactor plenums. The size of the developed model (1.035 billion cells) required the runs to be executed on the NCCS clusters (www.nccs.gov). STAR-CCM+ code (www.ed-adapco.com) was installed on two clusters: JAGUARXT5 and FROST, both of which were capable of executing this model. (authors)

  17. ENEL overall PWR plant models and neutronic integrated computing systems

    SciTech Connect

    Pedroni, G.; Pollachini, L.; Vimercati, G.; Cori, R.; Pretolani, F.; Spelta, S.

    1987-01-01

    To support the design activity of the Italian nuclear energy program for the construction of pressurized water reactors, the Italian Electricity Board (ENEL) needs to verify the design as a whole (that is, the nuclear steam supply system and balance of plant) both in steady-state operation and in transient. The ENEL has therefore developed two computer models to analyze both operational and incidental transients. The models, named STRIP and SFINCS, perform the analysis of the nuclear as well as the conventional part of the plant (the control system being properly taken into account). The STRIP model has been developed by means of the French (Electricite de France) modular code SICLE, while SFINCS is based on the Italian (ENEL) modular code LEGO. STRIP validation was performed with respect to Fessenheim French power plant experimental data. Two significant transients were chosen: load step and total load rejection. SFINCS validation was performed with respect to Saint-Laurent French power plant experimental data and also by comparing the SFINCS-STRIP responses.

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

    SciTech Connect

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

    1980-01-01

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

  19. Stress-corrosion cracking in BWR and PWR piping

    SciTech Connect

    Weeks, R.W.

    1983-07-01

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

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

    SciTech Connect

    Ishii, M.; Kataoka, I.

    1982-01-01

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

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

    SciTech Connect

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

    1983-01-01

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

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

    SciTech Connect

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

    1982-01-01

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

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

    SciTech Connect

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

    1983-06-01

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

  4. Method and apparatus for monitoring two-phase flow. [PWR

    DOEpatents

    Sheppard, J.D.; Tong, L.S.

    1975-12-19

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods.

  5. Amorphous and Nanocrystalline High Temperature Magnetic Material for PWR

    DTIC Science & Technology

    2006-03-01

    losses. Finite element software packages including FEMLAB©, FEMME © and FLEXPDE© were investigated to determine the fabrication parameters for the...analysis was based on a combination of NiZn ferrite as a core material with a spiral Cu coil. The geometry used in FEMME to simulate the effects of...various parameters and frequencies is shown in Figure III.4.5. FEMME © was chosen because it was simple to use and a planar inductor can be simulated

  6. Probability and consequences of misloading fuel in a PWR

    SciTech Connect

    Diamond, D.J.; Hsu, C.J.; Mubayi, V. )

    1991-08-01

    This report documents the results of a study into the frequency and consequences of misloading fresh fuel assemblies during the reloading of a pressurized water reactor. The consequences that were considered included (1) loss of required shutdown margin, (2) inadvertent criticality, and (3) worker exposure within the plant given inadvertent criticality. Neutronic calculations were performed for different patterns of fresh fuel clustered together in a Combustion Engineering rector. The fresh fuel considered had a high U-235 content and was assumed to be loaded without control element assemblies. The frequencies of misloading fresh fuel assemblies into these clustered patterns were calculated taking into account operator errors an equipment malfunctions that could occur during an offload/reload sequence. The study has improved our understanding of how difficult it is to misload fuel and has quantified the loss of shutdown margin and the frequency of occurrence for specific misloadings as well as the doses that might result from an inadvertent criticality. 15 refs., 18 figs., 13 tabs.

  7. Training and Health. Leonardo da Vinci Series: Good Practices.

    ERIC Educational Resources Information Center

    Commission of the European Communities, Brussels (Belgium). Directorate-General for Education and Culture.

    This document profiles programs in the fields of health and medicine that are offered through the European Commission's Leonardo da Vinci program. The following programs are profiled: (1) CYTOTRAIN (a transnational vocational training program in cervical cancer screening); (2) Apollo (a program of open and distance learning for paramedical…

  8. 40 CFR 60.49Da - Emission monitoring.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... (except a wet scrubber) is used for reducing PM, SO2, or carbon monoxide (CO) emissions; (B) Only natural..., calibrate, maintain, and operate a CEMS, and record the output of the system, for measuring the O2 or carbon... for which the data capture requirement of § 60.49Da(p)(4)(i) was not met. (iv) Notwithstanding...

  9. Leonardo da Vinci's foot: historical evidence of concept.

    PubMed

    Jastifer, James R; Toledo-Pereyra, Luis H

    2012-10-01

    Leonardo da Vinci (1452-1519), world-renowned Italian renaissance master, is known for his contributions to, and broad interests in science and art. The objective of this work is to demonstrate the extent of his science by applying the use of his concepts to current models of foot and ankle mechanics. The art and science of Leonardo Da Vinci were extensively analyzed by reviewing his original drawings and hand written notebooks as well as their English translation. Current medical journals including the topics of foot, ankle, and biomechanics were reviewed for modern evidence and application of his concepts. The library of Michigan State University and the electronic library of the Royal Library at Windsor Castle were extensively utilized. From the depths of Santa Maria Nuova Hospital in Florence and Santo Spirito Hospital in Rome, through his commentary and anatomical drawings of around 30 cadaver dissections he performed, Leonardo da Vinci expressed his concept of foot and ankle anatomy and mechanics. He laid forth concepts, which vary little from current theories including those of proportion, statics and joint stability, sesamoid biomechanics, and structural support of the foot. Leonardo da Vinci, by combining an interest in anatomy and a gift of genius and artistic ability laid a foundation of foot and ankle anatomy and mechanics that have been applied in modern clinical sciences. Leonardo in this way made important contributions to the practice of foot and ankle orthopedics.

  10. The Potential da Vinci in All of Us

    ERIC Educational Resources Information Center

    Petto, Sarah; Petto, Andrew

    2009-01-01

    The study of the human form is fundamental to both science and art curricula. For vertebrates, perhaps no feature is more important than the skeleton to determine observable form and function. As Leonard da Vinci's famous Proportions of the Human Figure (Virtruvian Man) illustrates, the size, shape, and proportions of the human body are defined by…

  11. The DaVinci Project: Multimedia in Art and Chemistry.

    ERIC Educational Resources Information Center

    Simonson, Michael; Schlosser, Charles

    1998-01-01

    Provides an overview of the DaVinci Project, a collaboration of students, teachers, and researchers in chemistry and art to develop multimedia materials for grades 3-12 visualizing basic concepts in chemistry and visual art. Topics addressed include standards in art and science; the conceptual framework for the project; and project goals,…

  12. 40 CFR 60.47Da - Commercial demonstration permit.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 6 2011-07-01 2011-07-01 false Commercial demonstration permit. 60... Steam Generating Units for Which Construction is Commenced After September 18, 1978 § 60.47Da Commercial... technology may apply to the Administrator for a commercial demonstration permit. The Administrator will...

  13. Studying and Working Abroad. Leonardo da Vinci Series: Good Practices.

    ERIC Educational Resources Information Center

    Commission of the European Communities, Brussels (Belgium). Directorate-General for Education and Culture.

    This document profiles recent successful examples of students studying and working abroad as part of the European Commission's Leonardo da Vinci program, which is designed to give students across the European Union the opportunity to experience vocational training in a foreign country. The following examples are presented: (1) 3 Finnish students…

  14. The Potential da Vinci in All of Us

    ERIC Educational Resources Information Center

    Petto, Sarah; Petto, Andrew

    2009-01-01

    The study of the human form is fundamental to both science and art curricula. For vertebrates, perhaps no feature is more important than the skeleton to determine observable form and function. As Leonard da Vinci's famous Proportions of the Human Figure (Virtruvian Man) illustrates, the size, shape, and proportions of the human body are defined by…

  15. The DaVinci Project: Multimedia in Art and Chemistry.

    ERIC Educational Resources Information Center

    Simonson, Michael; Schlosser, Charles

    1998-01-01

    Provides an overview of the DaVinci Project, a collaboration of students, teachers, and researchers in chemistry and art to develop multimedia materials for grades 3-12 visualizing basic concepts in chemistry and visual art. Topics addressed include standards in art and science; the conceptual framework for the project; and project goals,…

  16. Da Que Hablar (Something To Talk About), 1991-1993.

    ERIC Educational Resources Information Center

    Da Que Hablar, 1993

    1993-01-01

    This document consists of all 14 issues of a bimonthly serial, from its inception in May 1991 through November 1993. "Da Que Hablar" provides numerous authentic materials from magazines and newspapers to stimulate discussion in Spanish in the foreign language classroom. The articles cover topics such as current events, cultural issues,…

  17. Linear theory radial and nonradial pulsations of DA dwarf stars

    SciTech Connect

    Starrfield, S.; Cox, A.N.; Hodson, S.; Pesnell, W.D.

    1982-07-28

    The Los Alamos stellar envelope and radial linear non-adiabatic computer code, along with a new Los Alamos non-radial code are used to investigate the total hydrogen mass necessary to produce the non-radial instability of DA dwarfs. (GHT)

  18. Women and Technical Professions. Leonardo da Vinci Series: Good Practices.

    ERIC Educational Resources Information Center

    Commission of the European Communities, Brussels (Belgium). Directorate-General for Education and Culture.

    This document profiles programs for women in technical professions that are offered through the European Commission's Leonardo da Vinci program. The following programs are profiled: (1) Artemis and Diana (vocational guidance programs to help direct girls toward technology-related careers); (2) CEEWIT (an Internet-based information and…

  19. 40 CFR 60.45Da - Standard for mercury (Hg).

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 6 2010-07-01 2010-07-01 false Standard for mercury (Hg). 60.45Da... for mercury (Hg). (a) For each coal-fired electric utility steam generating unit other than an IGCC... gases that contain mercury (Hg) emissions in excess of each Hg emissions limit in paragraphs...

  20. Modelagem do vento e da fotosfera de AG Carinae

    NASA Astrophysics Data System (ADS)

    Groh, J. H.; Damineli, A.

    2003-08-01

    A trajetória evolutiva das estrelas de alta massa depende fortemente de suas taxas de perda de massa. Apesar do rápido progresso no estudo destas estrelas, a taxa de perda de massa e outros parâmetros físicos básicos, como a temperatura superficial e a velocidade terminal do vento ainda não estão bem determinados. Isto ocorre devido à presença de ventos irregulares, rápidos e fortes ao redor destas estrelas, tornando a interpretação dos seus espectros uma tarefa difícil. Assim, a modelagem do vento e da fotosfera dessas estrelas está sendo cada vez mais usada para obter tais parâmetros a partir dos espectros. O aumento da taxa de perda de massa durante a fase LBV (Variáveis Luminosas Azuis), comparado com outros tipos de estrelas, tem sido atribuído a instabilidades do tipo S Doradus. Dispomos de uma base de dados espectroscópicos cobrindo 22 anos de observações de AG Carinae, incluindo um ciclo S Doradus completo, com espectros CCD em alta resolução na faixa óptica e infravermelha. Utilizamos o programa desenvolvido por Schmutz (1997) para uma análise preliminar desse ciclo, obtendo a taxa de perda de massa a partir da linha do Ha. Não existe uma correlação clara da taxa de perda de massa com mudanças da temperatura efetiva, do raio da estrela e do fluxo na banda V. A estrela atingiu seu mínimo fotométrico (raio mínimo) em 1990 e o máximo fotométrico (raio máximo) em 1995, enquanto que o fluxo máximo da linha do Ha ocorreu em 1996. Além disso a taxa de perda de massa não segue esse ciclo, contrariamente às idéias correntes. Para fazer um modelo mais realista estamos usando o programa CMFGEN (Hillier & Miller), que trata a fotosfera e o vento estelar de forma consistente, considerando a radiação fora do equilíbrio termodinâmico (NLTE) e com blanketting total de linhas. Simulamos o espectro de AG Carinae em duas épocas extremas do ciclo S Dor para testar os resultados obtidos com o modelo mais simplificado.

  1. 49 CFR 178.58 - Specification 4DA welded steel cylinders for aircraft use.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 3 2014-10-01 2014-10-01 false Specification 4DA welded steel cylinders for...) SPECIFICATIONS FOR PACKAGINGS Specifications for Cylinders § 178.58 Specification 4DA welded steel cylinders for aircraft use. (a) Type, size, and service pressure. A DOT 4DA is a welded steel sphere (two...

  2. 49 CFR 178.58 - Specification 4DA welded steel cylinders for aircraft use.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 3 2012-10-01 2012-10-01 false Specification 4DA welded steel cylinders for...) SPECIFICATIONS FOR PACKAGINGS Specifications for Cylinders § 178.58 Specification 4DA welded steel cylinders for aircraft use. (a) Type, size, and service pressure. A DOT 4DA is a welded steel sphere (two...

  3. 49 CFR 178.58 - Specification 4DA welded steel cylinders for aircraft use.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 3 2013-10-01 2013-10-01 false Specification 4DA welded steel cylinders for...) SPECIFICATIONS FOR PACKAGINGS Specifications for Cylinders § 178.58 Specification 4DA welded steel cylinders for aircraft use. (a) Type, size, and service pressure. A DOT 4DA is a welded steel sphere (two...

  4. 49 CFR 178.58 - Specification 4DA welded steel cylinders for aircraft use.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 3 2011-10-01 2011-10-01 false Specification 4DA welded steel cylinders for...) SPECIFICATIONS FOR PACKAGINGS Specifications for Cylinders § 178.58 Specification 4DA welded steel cylinders for aircraft use. (a) Type, size, and service pressure. A DOT 4DA is a welded steel sphere (two...

  5. Adapter for contact force sensing of the da Vinci robot.

    PubMed

    Shimachi, Shigeyuki; Hirunyanitiwatna, Surakij; Fujiwara, Yasunori; Hashimoto, Akira; Hakozaki, Yoshinori

    2008-06-01

    At present, the da Vinci surgical robot system does not provide haptic feedback. One of the authors has proposed a contact-force sensing method called the 'overcoat method', in which the instrument/driver is supported by force sensors. In the da Vinci robot, the instrument jaws are powered by a wire-pulley mechanism; thus, in order to apply the overcoat method to the da Vinci system, we must transfer the power through a frame that is supported by force sensors. The authors have attempted to add a force-sensor function to the Sterile Adapter of the da Vinci system. In developing a sensorized adapter, a new configuration of force sensors and a new axial-force-free (AFF) joint have been devised in order to obtain an independent 'axial force effect' from the drive torque fed from the da Vinci robot arm. The force-sensing errors of the present system have been measured to have a maximum value of approximately 0.2 N while driving the jaws, and a maximum value of approximately 0.2 N when the robot arm is inclined with some excitation. Some impact reference forces applied on to the ends of the jaws agree with the outputs of the sensorized adapter to within <0.05 N. It is shown that the new adapter can be sterilized. One apprehension is that the total weight of the new adapter-approximately 1.2 kg-might unbalance the robot arm. In the case of the new adapter, the centre-line of the instrument shaft is shifted externally through approximately 3.5 mm from its original position. However, a new cannula for the da Vinci robot might solve this problem. The new configuration of force sensors and the new AFF joint work well in their basic functions. The total force-sensing error is estimated as approximately 0.5 N. One of the main reasons for the error appears to be the deformation of the adapter frame. (c) 2008 John Wiley & Sons, Ltd.

  6. JCoDA: a tool for detecting evolutionary selection

    PubMed Central

    2010-01-01

    Background The incorporation of annotated sequence information from multiple related species in commonly used databases (Ensembl, Flybase, Saccharomyces Genome Database, Wormbase, etc.) has increased dramatically over the last few years. This influx of information has provided a considerable amount of raw material for evaluation of evolutionary relationships. To aid in the process, we have developed JCoDA (Java Codon Delimited Alignment) as a simple-to-use visualization tool for the detection of site specific and regional positive/negative evolutionary selection amongst homologous coding sequences. Results JCoDA accepts user-inputted unaligned or pre-aligned coding sequences, performs a codon-delimited alignment using ClustalW, and determines the dN/dS calculations using PAML (Phylogenetic Analysis Using Maximum Likelihood, yn00 and codeml) in order to identify regions and sites under evolutionary selection. The JCoDA package includes a graphical interface for Phylip (Phylogeny Inference Package) to generate phylogenetic trees, manages formatting of all required file types, and streamlines passage of information between underlying programs. The raw data are output to user configurable graphs with sliding window options for straightforward visualization of pairwise or gene family comparisons. Additionally, codon-delimited alignments are output in a variety of common formats and all dN/dS calculations can be output in comma-separated value (CSV) format for downstream analysis. To illustrate the types of analyses that are facilitated by JCoDA, we have taken advantage of the well studied sex determination pathway in nematodes as well as the extensive sequence information available to identify genes under positive selection, examples of regional positive selection, and differences in selection based on the role of genes in the sex determination pathway. Conclusions JCoDA is a configurable, open source, user-friendly visualization tool for performing evolutionary

  7. PanDaTox: A tool for accelerated metabolic engineering

    SciTech Connect

    Amitai, Gil; Sorek, Rotem

    2012-07-18

    Metabolic engineering is often facilitated by cloning of genes encoding enzymes from various heterologous organisms into E. coli. Such engineering efforts are frequently hampered by foreign genes that are toxic to the E. coli host. We have developed PanDaTox (www.weizmann.ac.il/pandatox), a web-based resource that provides experimental toxicity information for more than 1.5 million genes from hundreds of different microbial genomes. The toxicity predictions, which were extensively experimentally verified, are based on serial cloning of genes into E. coli as part of the Sanger whole genome shotgun sequencing process. PanDaTox can accelerate metabolic engineering projects by allowing researchers to exclude toxic genes from the engineering plan and verify the clonability of selected genes before the actual metabolic engineering experiments are conducted.

  8. [Giovanbattista da Monte (Montanus): father of modern clinical medicine].

    PubMed

    Franceschetti, Diego; Agazia, Bruno; Zanchin, Giorgio

    2005-01-01

    The figure of Giovanbattista da Monte (1489-1551) is associated with the introduction of clinical teaching at the patient's beside, in 1543, at the San Francesco Hospital of Padua. In the XVI century, teaching was still based on the explanation and comment of the ancient authors and the educational programme was founded on theoretical aspects. The "practical" approach consisted of the treatment "ex cathedra" of diseases according to the various parts of the body, without observing the course of the pathological events with a direct confirmation at the patient's beside.To his merit, Da Monte established the practise of training students to gather the case history, to carry out an objective examination, and to closely examine disease phenomena with lessons at the bedside of the patient. Practical clinical training was thus introduced as the crucial moment in the formation of the physician.

  9. TA-DA: A TOOL FOR ASTROPHYSICAL DATA ANALYSIS

    SciTech Connect

    Da Rio, Nicola; Robberto, Massimo

    2012-12-01

    We present the Tool for Astrophysical Data Analysis (TA-DA), a new software aimed to greatly simplify and improve the analysis of stellar photometric data in comparison with theoretical models, and allow the derivation of stellar parameters from multi-band photometry. Its flexibility allows one to address a number of such problems: from the interpolation of stellar models, or sets of stellar physical parameters in general, to the computation of synthetic photometry in arbitrary filters or units; from the analysis of observed color-magnitude diagrams to a Bayesian derivation of stellar parameters (and extinction) based on multi-band data. TA-DA is available as a pre-compiled Interactive Data Language widget-based application; its graphical user interface makes it considerably user-friendly. In this paper, we describe the software and its functionalities.

  10. A novel 29-kDa chicken heat shock protein.

    PubMed

    Einat, M F; Haberfeld, A; Shamay, A; Horev, G; Hurwitz, S; Yahav, S

    1996-12-01

    The family of small heat shock proteins is the more variable among the highly conserved superfamily of heat shock proteins (HSP). Using a metabolic labeling procedure with tissue explants, we have detected in chickens a new member of the small HSP family with an apparent molecular weight of 29-kDa. This protein was induced in broiler chickens' heart muscle and lungs following an in vivo heat stress. The 29-kDa band appears after 3 h of heat stress, much later than the induction of HSP 90, HSP 70, and HSP 27. The late onset of induction suggests that HSP 29 plays a more specific role of a "second stage defense protein".

  11. DaVinci's Mona Lisa entering the next dimension.

    PubMed

    Carbon, Claus-Christian; Hesslinger, Vera M

    2013-01-01

    For several of Leonardo da Vinci's paintings, such as The Virgin and Child with St Anne or the Mona Lisa, there exist copies produced by his own studio. In case of the Mona Lisa, a quite exceptional, rediscovered studio copy was presented to the public in 2012 by the Prado Museum in Madrid. Not only does it mirror its famous counterpart superficially; it also features the very same corrections to the lower layers, which indicates that da Vinci and the 'copyist' must have elaborated their panels simultaneously. On the basis of subjective (thirty-two participants estimated painter-model constellations) as well as objective data (analysis of trajectories between landmarks of both paintings), we revealed that both versions differ slightly in perspective. We reconstructed the original studio setting and found evidence that the disparity between both paintings mimics human binocular disparity. This points to the possibility that the two Giocondas together might represent the first stereoscopic image in world history.

  12. Calibration of Synthetic Photometry Using DA White Dwarfs

    NASA Astrophysics Data System (ADS)

    Holberg, J. B.; Bergeron, P.

    2005-12-01

    We have calibrated four major ground-based photometric systems with respect to the Hubble Space Telescope absolute flux scale, which is defined by Vega and four fundamental DA white dwarfs. These photometric systems include the Johnson-Kron-Cousins UBVRI, the Stromgren uvby filters, the 2MASS JHKs and the Sloan Digital Sky Survey ugriz filters. Synthetic magnitudes are calculated from model white dwarf spectra folded through the published filter response functions, these magnitudes in turn are absolutely calibrated with respect to the HST flux scale. Effective zero magnitude fluxes and zero point offsets of each system are determined. In order to verify the external observational consistency as well as to demonstrate the applicability of these definitions, the synthetic magnitudes are compared with the respective observed magnitudes of larger sets of DA white dwarfs that have well determined effective temperatures and surface gravities and which span a wide range in both of these parameters.

  13. Da Vinci Xi Robot-Assisted Penetrating Keratoplasty.

    PubMed

    Chammas, Jimmy; Sauer, Arnaud; Pizzuto, Joëlle; Pouthier, Fabienne; Gaucher, David; Marescaux, Jacques; Mutter, Didier; Bourcier, Tristan

    2017-06-01

    This study aims (1) to investigate the feasibility of robot-assisted penetrating keratoplasty (PK) using the new Da Vinci Xi Surgical System and (2) to report what we believe to be the first use of this system in experimental eye surgery. Robot-assisted PK procedures were performed on human corneal transplants using the Da Vinci Xi Surgical System. After an 8-mm corneal trephination, four interrupted sutures and one 10.0 monofilament running suture were made. For each procedure, duration and successful completion of the surgery as well as any unexpected events were assessed. The depth of the corneal sutures was checked postoperatively using spectral-domain optical coherence tomography (SD-OCT). Robot-assisted PK was successfully performed on 12 corneas. The Da Vinci Xi Surgical System provided the necessary dexterity to perform the different steps of surgery. The mean duration of the procedures was 43.4 ± 8.9 minutes (range: 28.5-61.1 minutes). There were no unexpected intraoperative events. SD-OCT confirmed that the sutures were placed at the appropriate depth. We confirm the feasibility of robot-assisted PK with the new Da Vinci Surgical System and report the first use of the Xi model in experimental eye surgery. Operative time of robot-assisted PK surgery is now close to that of conventional manual surgery due to both improvement of the optical system and the presence of microsurgical instruments. Experimentations will allow the advantages of robot-assisted microsurgery to be identified while underlining the improvements and innovations necessary for clinical use.

  14. Improving Security in the ATLAS PanDA System

    NASA Astrophysics Data System (ADS)

    Caballero, J.; Maeno, T.; Nilsson, P.; Stewart, G.; Potekhin, M.; Wenaus, T.

    2011-12-01

    The security challenges faced by users of the grid are considerably different to those faced in previous environments. The adoption of pilot jobs systems by LHC experiments has mitigated many of the problems associated with the inhomogeneities found on the grid and has greatly improved job reliability; however, pilot jobs systems themselves must then address many security issues, including the execution of multiple users' code under a common 'grid' identity. In this paper we describe the improvements and evolution of the security model in the ATLAS PanDA (Production and Distributed Analysis) system. We describe the security in the PanDA server which is in place to ensure that only authorized members of the VO are allowed to submit work into the system and that jobs are properly audited and monitored. We discuss the security in place between the pilot code itself and the PanDA server, ensuring that only properly authenticated workload is delivered to the pilot for execution. When the code to be executed is from a 'normal' ATLAS user, as opposed to the production system or other privileged actor, then the pilot may use an EGEE developed identity switching tool called gLExec. This changes the grid proxy available to the job and also switches the UNIX user identity to protect the privileges of the pilot code proxy. We describe the problems in using this system and how they are overcome. Finally, we discuss security drills which have been run using PanDA and show how these improved our operational security procedures.

  15. GEODSS Tracking Results on Asteroid 2012 DA14

    DTIC Science & Technology

    2013-06-12

    meteor until stony or metallic material remnants make it to the surface of the planet in which case they become meteorites . Theoretically, the...passing in close proximity to the planet are a daily manifestation. In fact within days of the 2012 DA14 transit of the planet a large meteorite entered...although meteorites approximately basketball size hit the Earth about once per day, and objects the size of an automobile reach us roughly on a weekly

  16. Da Vinci's codex and the anatomy of healthcare.

    PubMed

    Stephens-Borg, Keith

    2012-08-01

    We usually display a laid-back approach to medical jargon throughout our theatre work. The word 'perioperative' is built from the Greek word 'peri' (around) and the Latin 'operari' (to work). Latin and Greek became the prefixed language of choice for Leonardo da Vinci, and his research was pivotal in determining the way in which surgical procedures are documented. Ancient manuscripts aided the unfolding of the secrets of anatomy, and Leonardo revealed that art was the key in expressive detailed explanation.

  17. Is there any mantle plume beneath Tristan da Cunha?

    NASA Astrophysics Data System (ADS)

    Schloemer, A.; Geissler, W. H.; Jegen, M. D.; Jokat, W.

    2015-12-01

    Tristan da Cunha is a volcanic island in the South Atlantic located very close to the Mid-Atlantic Ridge. Generally, it is accepted to be the location of a mantle plume, which has been active at least since the breakup of Gondwana at 130 Ma, the time when the Paraná/Etendeka flood basalts were emplaced. Furthermore, it is associated with the formation of the Walvis Ridge and the Rio Grande Rise, and therefore it's one of the key examples of a hot spot track linking a flood basalt province to an active ocean island volcano. However, global tomography models are contradicting about the origin of Tristan da Cunha: Whether it is a deep mantle plume or caused by shallow plate tectonics. To gain a better understanding, we deployed 24 broadband ocean-bottom seismometers, 26 ocean-bottom electromagnetic stations and 2 seismological land stations in January 2012 with the German research vessel Maria S. Merian. We acquired continuous seismological data for one year and recovered the instruments in January 2013.We use cross-correlated travel time residuals of teleseismic earthquakes to perform a finite-frequency tomography to resolve the P wave velocity upper mantle structure beneath the island. Here we show our preliminary results of the 3-D velocity perturbations in the upper mantle: We do not image a plume-like structure directly beneath the island. Instead we observe a low velocity region in the southwest of our array that might be related to a local mantle upwelling (mantle plume). Additionally we show the local seismicity in the Tristan da Cunha region.Chen et al. and Baba et al. will present the first results on the magnetotelluric experiment and Ryberg et al. will present the crustal structure around the Tristan da Cunha hotspot.

  18. DA-9801 promotes neurite outgrowth via ERK1/2-CREB pathway in PC12 cells.

    PubMed

    Won, Jong Hoon; Ahn, Kyong Hoon; Back, Moon Jung; Ha, Hae Chan; Jang, Ji Min; Kim, Ha Hyung; Choi, Sang-Zin; Son, Miwon; Kim, Dae Kyong

    2015-01-01

    In the present study, we examined the mechanisms underlying the effect of DA-9801 on neurite outgrowth. We found that DA-9801 elicits its effects via the mitogen-activated protein kinase (MEK) extracellular signal-regulated kinase (ERK)1/2-cAMP response element-binding protein (CREB) pathway. DA-9801, an extract from a mixture of Dioscorea japonica and Dioscorea nipponica, was reported to promote neurite outgrowth in PC12 cells. The effects of DA-9801 on cell viability and expression of neuronal markers were evaluated in PC12 cells. To investigate DA-9801 action, specific inhibitors targeting the ERK signaling cascade were used. No cytotoxicity was observed in PC12 cells at DA-9801 concentrations of less than 30 µg/mL. In the presence of nerve growth factor (NGF, 2 ng/mL), DA-9801 promoted neurite outgrowth and increased the relative mRNA levels of neurofilament-L (NF-L), a marker of neuronal differentiation. The Raf-1 inhibitor GW5074 and MEK inhibitor PD98059 significantly attenuated DA-9801-induced neurite outgrowth. Additionally, the MEK1 and MEK2 inhibitor SL327 significantly attenuated the increase in the percentage of neurite-bearing PC12 cells induced by DA-9801 treatment. Conversely, the selective p38 mitogen-activated protein kinase inhibitor SB203580 did not attenuate the DA-9801 treatment-induced increase in the percentage of neurite-bearing PC12 cells. DA-9801 enhanced the phosphorylation of ERK1/2 and CREB in PC12 cells incubated with and without NGF. Pretreatment with PD98059 blocked the DA-9801-induced phosphorylation of ERK1/2 and CREB. In conclusion, DA-9801 induces neurite outgrowth by affecting the ERK1/2-CREB signaling pathway. Insights into the mechanism underlying this effect of DA-9801 may suggest novel potential strategies for the treatment of peripheral neuropathy.

  19. Domain annotation of trimeric autotransporter adhesins—daTAA

    PubMed Central

    Szczesny, Pawel; Lupas, Andrei

    2008-01-01

    Motivation: Trimeric autotransporter adhesins (TAAs), such as Yersinia YadA, Neisseria NadA, Moraxella UspAs, Haemophilus Hia and Bartonella BadA, are important pathogenicity factors of proteobacteria. Their high sequence diversity and distinct mosaic-like structure lead to difficulties in the annotation of their sequences. These stem from the large number of short repeats, the presence of compositionally unusual coiled-coils, fuzzy domain boundaries and regions of seemingly low sequence complexity. Results: We have developed a workflow, named daTAA, for the accurate domain annotation of TAAs. Its core consists of manually curated alignments and of knowledge-based rules that enhance assignments made by sequence similarity. Compared to general domain annotation servers such as PFAM, daTAA captures more domains and provides more sensitive domain detection, as well as integrated and detailed coiled-coil assignments. Availability: The daTAA server is freely accessible at http://toolkit.tuebingen.mpg.de/dataa Contact: andrei.lupas@tuebingen.mpg.de Supplementary information: Supplementary data are available at Bioinformatics online. PMID:18397894

  20. Common commercial cosmetic products induce arthritis in the DA rat.

    PubMed Central

    Sverdrup, B; Klareskog, L; Kleinau, S

    1998-01-01

    Many different agents, including mineral oil and silicone, have the capacity to act as immunological adjuvants, i.e., they can contribute to the activation of the immune system. Some adjuvants, including mineral oil, are known to induce arthritis in certain strains of rats after intradermal injection or percutaneous application. The aim of this study was to determine if common commercial cosmetic products containing mineral oil could induce arthritis in the highly susceptible DA (Dark Agouti) rat. Intradermal injection of five out of eight assayed cosmetic products without further additives resulted in arthritis with synovitis. One of the products induced a very aggressive arthritis, which had declined after 5-9 weeks. When this product was also assayed for arthritogenicity upon percutaneous administration, it induced a mild and transient arthritis in 5 out of 10 DA rats, whereas control animals showed no clinical signs of joint involvement. No arthritic reaction was seen in rats after peroral feeding with the most arthritogenic product or by intravaginal application of Freund's adjuvants. Silicone gel implants in DA rats did not cause arthritis. We conclude that mineral oils included in common commercially available products retain their adjuvant properties and are arthritogenic in the presently investigated arthritis-prone rat strain. There is yet no evidence that mineral oils present in cosmetics may contribute to arthritis in humans, but we suggest that this question should be subject to further investigation. Images Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Figure 6 PMID:9417771

  1. The DaVinci Group: a second modern Ophthalmotrope.

    PubMed

    Pruehsner, William R; Enderle, John D

    2006-01-01

    A group of undergraduate students at the University of Connecticut Biomedical Engineering Program has formed a "club" in order to more fully understand and educate themselves in modeling anatomical processes. This group is called the DaVinci Robot or DaVinci Group. Experiments to mechanically model the six extraocular muscles of the eye have been performed, each meeting little success. While researching methods that would lead to better success, the concept of the Ophthalmotrope was discovered. The Ophthalmotrope is a mechanical visual aide used in teaching the function of the extraocular muscles, prevalent in the mid 1800's. The Group decided to study this device and ultimately decided to build one. The paper presented here discusses our third experiment, currently under investigation, that is, to build an Opthalmotrope. Difficulties with this task are lack of any information with regard to how to construct this device. Presented are descriptions of the Group's initial experiments and research conducted into the construction of the Ophthalmotrpe. In the main body of the presented paper is a description of how the DaVinci Group Ophthalmotrope is constructed. Concluding is a discussion of the progress of the construction of the Ophthalmotrope along with a brief listing of research conducted in order to build the device.

  2. The ATLAS PanDA Pilot in Operation

    NASA Astrophysics Data System (ADS)

    Nilsson, P.; Caballero, J.; De, K.; Maeno, T.; Stradling, A.; Wenaus, T.; ATLAS Collaboration

    2011-12-01

    The Production and Distributed Analysis system (PanDA) [1-2] was designed to meet ATLAS [3] requirements for a data-driven workload management system capable of operating at LHC data processing scale. Submitted jobs are executed on worker nodes by pilot jobs sent to the grid sites by pilot factories. This paper provides an overview of the PanDA pilot [4] system and presents major features added in light of recent operational experience, including multi-job processing, advanced job recovery for jobs with output storage failures, gLExec [5-6] based identity switching from the generic pilot to the actual user, and other security measures. The PanDA system serves all ATLAS distributed processing and is the primary system for distributed analysis; it is currently used at over 100 sites worldwide. We analyze the performance of the pilot system in processing real LHC data on the OSG [7], EGI [8] and Nordugrid [9-10] infrastructures used by ATLAS, and describe plans for its evolution.

  3. Pictorial cues constrain depth in da Vinci stereopsis.

    PubMed

    Makino, Yoshinari; Yano, Masafumi

    2006-01-01

    "da Vinci stereopsis" is defined as depth seen in a monocular object occluded by a binocular one, and the visual system must solve its depth ambiguity [Nakayama, K., & Shimojo, S. (1990). da Vinci stereopsis: Depth and subjective occluding contours from unpaired image points. Vision Research, 30, 1811-1825]. Although fused images include various pictorial features, effects of pictorial depth cues have never been systematically investigated in da Vinci stereopsis. To examine this, we created stereograms consisting of a monocular bar flanked by binocular bars with a fixed large horizontal separation, in which the monocular bar induced a subjective occluding edge. Manipulating vertical size or contrast of the bars could affect the depth of the monocular bar. Conflicting these cues revealed that the effect of vertical size was stronger than that of contrast in all our subjects. Measurements of the depth indicated that the relative vertical size of the bars quantitatively determined the perceived depth, of which levels had large inter-subject differences. All these experiments indicate that the visual system can use the pictorial depth cues as a constraint to determine the depth of monocular elements.

  4. OpenDA-NEMO framework for ocean data assimilation

    NASA Astrophysics Data System (ADS)

    van Velzen, Nils; Altaf, Muhammad Umer; Verlaan, Martin

    2016-05-01

    Data assimilation methods provide a means to handle the modeling errors and uncertainties in sophisticated ocean models. In this study, we have created an OpenDA-NEMO framework unlocking the data assimilation tools available in OpenDA for use with NEMO models. This includes data assimilation methods, automatic parallelization, and a recently implemented automatic localization algorithm that removes spurious correlations in the model based on uncertainties in the computed Kalman gain matrix. We have set up a twin experiment where we assimilate sea surface height (SSH) satellite measurements. From the experiments, we can conclude that the OpenDA-NEMO framework performs as expected and that the automatic localization significantly improves the performance of the data assimilation algorithm by successfully removing spurious correlations. Based on these results, it looks promising to extend the framework with new kinds of observations and work on improving the computational speed of the automatic localization technique such that it becomes feasible to include large number of observations.

  5. Identification of an abundant 56 kDa protein implicated in food allergy as granule-bound starch synthase

    USDA-ARS?s Scientific Manuscript database

    Rice, the staple food of South and East Asian counties, is considered to be hypoallergenic. However, several clinical studies have documented rice-induced allergy in sensitive patients. Rice proteins with molecular weights of 14-16 kDa, 26 kDa, 33 kDa and 56 kDa have been identified as allergens. Re...

  6. [The Arabic influence in the "Colóquios dos simples e drogas da India" of Garcia da Orta].

    PubMed

    Ricordel, Joëlle

    2015-09-01

    The "Colóquios dos simples e drogas he cousas medicinais de Índia" (Conversations on the simples, drugs and medicinal substances of India) (1563) of Garcia da Orta is a botanical and pharmacognosy book. The author is a Portuguese physician who studied in the Spanish universities and practiced medicine mainly in India. He studies in short chapters presented in the form of dialogues about sixty simples. Sources to which he refers are indicative of a "classical" training, but also the mark of a curious and open mind to different cultures. The Arabic sources are numerous and mainly concern the identification of substances by abundant synonyms of their names in foreign languages and different medicinal uses that may have been done by the ancient physicians. However, Da Orta is critical with respect to these sources, seeking contradictions and differences of opinion among authors. He confronts them with the oral information collected thanks to a wide network of contacts.

  7. Seeking 'energy' vs. pain relief in spas in Brazil (Caldas da Imperatriz) and Portugal (Termas da Sulfurea).

    PubMed

    Quintela, Maria Manuel

    2011-04-01

    This paper is a comparative ethnography of the therapeutic practices at two different spa locations: Caldas da Imperatriz, SC, Brazil, and Termas da Sulfurea in Cabeco de Vide, Portugal. The comparison reveals the existence of contrasting 'explanatory models' held by the spa-goers as well as by the official medical systems. In the Portuguese context this model is highly medicalized; in the Brazilian case, spa treatments are viewed as 'alternative' or 'complementary' therapy and are also related to religious philosophies. Each model corresponds to a different idiom expressing certain experiences and world views, one focusing on 'pains' (dores) and the other on 'energy' (energia), the former leading to the rationale of 'curing', the latter to the notion of 'energizing'. In this paper the author intends to analyze and contrast the categories found in these models, which originate from different conceptions of health, illness and healing for Brazilian and Portuguese spa-goers.

  8. Characterization of a dopamine receptor (DA sub 2K ) in the kidney inner medulla

    SciTech Connect

    Huo, T.; Ye, M.Q.; Healy, D.P. )

    1991-04-15

    Dopamine (DA) produces a natriuretic/diuretic response in the kidney by mechanisms that are still not well understood. There is some indication that DA{sub 2} receptors may be involved in mediating the effects of DA, but little is known regarding the nature of this receptor in the kidney. Autoradiographic localization of ({sup 3}H)spiperone, a DA{sub 2} antagonist, indicated that high-density binding was restricted to inner medullary collecting ducts (IMCDs). ({sup 3}H)Spiperone binding was saturable, high affinity and high density. Functionally, DA stimulated prostaglandin E{sub 2} production by IMCD cells, an effect that could be blocked by the DA{sub 2} antagonist domperidone. These results indicate that the kidney inner medulla expresses a functional DA receptor that may represent a newly identified DA receptor subtype (here designated DA{sub 2K}). Moreover, these results suggest that the kidney inner medulla may be a significant site at which DA, either directly or indirectly, influences water and electrolyte excretion.

  9. The primary structure of skeletal muscle myosin heavy chain: III. Sequence of the 22 kDa fragment and the alignment of the 23 kDa, 50 kDa, and 22 kDa fragments.

    PubMed

    Maita, T; Miyanishi, T; Matsuzono, K; Tanioka, Y; Matsuda, G

    1991-07-01

    The amino acid sequence of the 197-residue 22 kDa fragment from chicken pectoralis muscle was determined to be as follows: K-K-G-S-S-F-Q-T-V-S-A-L-F-R-E-N-L-N-K-L- M-A-N-L-R-S-T-H-P-H-F-V-R-C-I-I-P-N-E-T-K-T-P-G-A-M-E-H-E-L-V-L-H-Q-L-R- C-N-G-V- L-E-G-I-R-I-C-R-K-G-F-P-S-R-V-L-Y-A-D-F-K-Q-R-Y-R-V-L-N-A-S-A-I-P-E-G-Q- F-M-D-S- K-K-A-S-E-K-L-L-G-S-I-D-V-D-h-T-Q-Y-R-F-G-H-T-K-V-F-F-K-A-G-L-L-G-L-L-E- E-M-R-D- D-K-L-A-E-I-I-T-R-T-Q-A-R-C-R-G-F-L-M-R-V-E-Y-R-R-M-V-E-R-R-E-S-I-F-C-I- Q-Y-N-V-R-S-F-M-N-V-K-H-W-P-W-M-K-L-F-F-K, where h stands for 3-N-methylhistidine. The amino acid sequences of the 22 kDa fragment and its equivalent fragment from chicken ventricle and gizzard muscle myosins were also determined by our group. Predicted secondary structures of these 22 kDa fragment regions and of the reported chicken embryo myosin revealed some possible structural differences.(ABSTRACT TRUNCATED AT 250 WORDS)

  10. Leonardo da Vinci and Kethem-Kiveris vena.

    PubMed

    Dolezal, Antonín; Skorepova-Honzlova, Zita; Jelen, Karel

    2012-01-01

    In the drawing of coitus by Leonardo da Vinci are pictured the contemporary hypotheses regarding this act. The authors analyze the mamillaruteral connection depicted by the artist and grow up to believe that this is a hypothetical kiveris vena, female vein described by Anatomist Master Nicolai Physicus from the Salerno School. The Hebrew roots were found in the name. The connection is described also by Mondino in The Anathomia. The same connection can be found in the picture of the pregnant woman in Fasciculus Medicinæ by Johannes De Ketham.

  11. LEONARDO DA VINCI AND THE ORIGIN OF SEMEN.

    PubMed

    Noble, Denis; DiFrancesco, Dario; Zancani, Diego

    2014-12-20

    It is well known that Leonardo da Vinci made several drawings of the human male anatomy. The early drawings (before 1500) were incorrect in identifying the origin of semen, where he followed accepted teaching of his time. It is widely thought that he did not correct this mistake, a view that is reflected in several biographies. In fact, he made a later drawing (after 1500) in which the description of the anatomy is remarkably accurate and must have been based on careful dissection. In addition to highlighting this fact, acknowledged previously in only one other source, this article reviews the background to Leonardo's knowledge of the relevant anatomy.

  12. A quasi-passive CMOS pipeline D/A converter

    NASA Technical Reports Server (NTRS)

    Wang, Fong-Jim; Temes, Gabor C.; Law, Simon

    1989-01-01

    A novel pipeline digital-to-analog converter configuration, based on switched-capacitor techniques, is described. An n-bit D/A conversion can be implemented by cascading n + 1 unit cells. The device count of the circuit increases linearly, not exponentially, with the conversion accuracy. The new configuration can be pipelined. Hence, the conversion rate can be increased without requiring a higher clock rate. An experimental 10-bit DAC prototype has been fabricated using a 3-micron CMOS process. The results show that high-speed, high-accuracy, and low-power operation can be achieved without special process or postprocess trimming.

  13. Ultratight crystal packing of a 10 kDa protein

    SciTech Connect

    Trillo-Muyo, Sergio; Chruszcz, Maksymilian; Minor, Wladek; Kuisiene, Nomeda

    2013-03-01

    The crystal structure of the C-terminal domain of a putative U32 peptidase from G. thermoleovorans is reported; it is one of the most tightly packed protein structures reported to date. While small organic molecules generally crystallize forming tightly packed lattices with little solvent content, proteins form air-sensitive high-solvent-content crystals. Here, the crystallization and full structure analysis of a novel recombinant 10 kDa protein corresponding to the C-terminal domain of a putative U32 peptidase are reported. The orthorhombic crystal contained only 24.5% solvent and is therefore among the most tightly packed protein lattices ever reported.

  14. Leonardo da Vinci and the origin of semen

    PubMed Central

    Noble, Denis; DiFrancesco, Dario; Zancani, Diego

    2014-01-01

    It is well known that Leonardo da Vinci made several drawings of the human male anatomy. The early drawings (before 1500) were incorrect in identifying the origin of semen, where he followed accepted teaching of his time. It is widely thought that he did not correct this mistake, a view that is reflected in several biographies. In fact, he made a later drawing (after 1500) in which the description of the anatomy is remarkably accurate and must have been based on careful dissection. In addition to highlighting this fact, acknowledged previously in only one other source, this article reviews the background to Leonardo's knowledge of the relevant anatomy. PMID:27494016

  15. High Resolution EUV & FUV Spectroscopy of DA White Dwarfs

    NASA Astrophysics Data System (ADS)

    Barstow, M. A.; Good, S. A.; Bannister, N. P.; Burleigh, M. R.; Holberg, J. B.; Bruhweiler, F. C.; Napiwotzki, R.; Cruddace, R. G.; Kowalski, M. P.

    We report on recent results from a high-resolution spectroscopic survey of hot DA white dwarfs, based on IUE, FUSE and HST observations. For the first time, we address the measurement of element abundances in a completely objective manner with a spectroscopic model fitting technique, which allows us to consider formally the limits that can be placed on abundances in stars where no heavy elements are detected. We also include our latest analysis of the high resolution EUV spectrum of G191-B2B recorded by J-PEX.

  16. Mammary artery harvesting using the Da Vinci Si robotic system

    PubMed Central

    Canale, Leonardo Secchin; Bonatti, Johannes

    2014-01-01

    Internal mammary artery harvesting is an essential part of any coronary artery bypass operation. Totally endoscopic coronary artery bypass graft surgery has become reality in many centers as a safe and effective alternative to conventional surgery in selected patients. Internal mammary artery harvesting is the initial part of the procedure and should be performed equally safely if one wants to achieve excellence in patency rates for the bypass. We here describe the technique for mammary harvesting with the Da Vinci Si robotic system. PMID:24896171

  17. Visual tracking of da Vinci instruments for laparoscopic surgery

    NASA Astrophysics Data System (ADS)

    Speidel, S.; Kuhn, E.; Bodenstedt, S.; Röhl, S.; Kenngott, H.; Müller-Stich, B.; Dillmann, R.

    2014-03-01

    Intraoperative tracking of laparoscopic instruments is a prerequisite to realize further assistance functions. Since endoscopic images are always available, this sensor input can be used to localize the instruments without special devices or robot kinematics. In this paper, we present an image-based markerless 3D tracking of different da Vinci instruments in near real-time without an explicit model. The method is based on different visual cues to segment the instrument tip, calculates a tip point and uses a multiple object particle filter for tracking. The accuracy and robustness is evaluated with in vivo data.

  18. An improvement of LLNA:DA to assess the skin sensitization potential of chemicals.

    PubMed

    Zhang, Hongwei; Shi, Ying; Wang, Chao; Zhao, Kangfeng; Zhang, Shaoping; Wei, Lan; Dong, Li; Gu, Wen; Xu, Yongjun; Ruan, Hongjie; Zhi, Hong; Yang, Xiaoyan

    2017-01-01

    We developed a modified local lymph node assay based on ATP (LLNA:DA), termed the Two-Stage LLNA:DA, to further reduce the animal numbers in the identification of sensitizers. In the Two-Stage LLNA:DA procedure, 13 chemicals ranging from non-sensitizers to extreme sensitizers were selected. The first stage used reduced LLNA:DA (rLLNA:DA) to screen out sensitive chemicals. The second stage used LLNA:DA based on OECD 442 (A) to classify those potential sensitizers screened out in the first stage. In the first stage, the SIs of the methyl methacrylate, salicylic acid, methyl salicylate, ethyl salicylate, isopropanol and propanediol were below 1.8 and need not to be tested in the second step. Others continued to be tested by LLNA:DA. In the second stage, sodium lauryl sulphate and xylene were classified as weak sensitizers. a-hexyl cinnamic aldehyde and eugenol were moderate sensitizers. Benzalkonium chloride and glyoxal were strong sensitizers, and phthalic anhydride was an extreme sensitizer. The 9/9, 11/12, 10/11, and 8/13 (positive or negative only) categories of the Two-Stage LLNA:DA were consistent with those from the other methods (LLNA, LLNA:DA, GPMT/BT and HMT/HPTA), suggesting that Two-Stage LLNA:DA have a high coincidence rate with reported data. In conclusion, The Two-Stage LLNA:DA is in line with the "3R" rules, and can be a modification of LLNA:DA but needs more study.

  19. Purification, characterization and gene cloning of Da-36, a novel serine protease from Deinagkistrodon acutus venom.

    PubMed

    Zheng, Ying; Ye, Feng-Ping; Wang, Jie; Liao, Guo-Yang; Zhang, Yun; Fan, Quan-Shui; Lee, Wen-Hui

    2013-06-01

    A serine protease termed Da-36 was isolated from crude venom of Deinagkistrodon acutus. The enzyme was a single chain protein with an apparent molecular weight of 36,000 on SDS-PAGE with an isoelectric point of 6.59. Da-36 could clot human plasma by cleaving the Aα, Bβ and γ chains of fibrinogen and also exhibited arginine esterase activity. The proteolytic activity of Da-36 toward TAME was strongly inhibited by PMSF and moderately affected by benzamidine and aprotinin, indicating that it was a serine protease. Meanwhile, Da-36 showed stability with wide temperature (20-50 °C) and pH value ranges (pH 6-10). Divalent metal ions of Ca(2+), Mg(2+), and Mn(2+) had no effects but Zn(2+) and Cu(2+) inhibited the arginine esterase activity of Da-36. Total DNA was extracted directly from the lyophilized crude venom and the gene (5.5 kbp) coding for Da-36 had been successfully cloned. Sequence analysis revealed that the Da-36 gene contained five exons and four introns. The mature Da-36 was encoded by four separate exons. The deduced mature amino acid sequence of Da-36 was in good agreement with the determined N-terminal sequence of the purified protein and shared high homology with other serine proteases isolated from different snake venoms. Blast search using amino acid sequence of Da-36 against public database revealed that Da-36 showed a maximal identity of 90% with both Dav-X (Swiss-Prot: Q9I8W9.1) and thrombin-like protein 1 (GenBank: AAW56608.1) from the same snake species, indicating that Da-36 is a novel serine protease.

  20. Identification of Methylated Deoxyadenosines in Genomic DNA by dA6m DNA Immunoprecipitation

    PubMed Central

    Koziol, Magdalena J.; Bradshaw, Charles R.; Allen, George E.; Costa, Ana S.H.; Frezza, Christian

    2017-01-01

    dA6m DNA immunoprecipitation followed by deep sequencing (DIP-Seq) is a key tool in identifying and studying the genome-wide distribution of N6-methyldeoxyadenosine (dA6m). The precise function of this novel DNA modification remains to be fully elucidated, but it is known to be absent from transcriptional start sites and excluded from exons, suggesting a role in transcriptional regulation (Koziol et al., 2015). Importantly, its existence suggests that DNA might be more diverse than previously believed, as further DNA modifications might exist in eukaryotic DNA (Koziol et al., 2015). This protocol describes the method to perform dA6m DNA immunoprecipitation (DIP), as was applied to characterize the first dA6m methylome analysis in higher eukaryotes (Koziol et al., 2015). In this protocol, we describe how genomic DNA is isolated, fragmented and then DNA containing dA6m is pulled down with an antibody that recognizes dA6m in genomic DNA. After subsequent washes, DNA fragments that do not contain dA6m are eliminated, and the dA6m containing fragments are eluted from the antibody in order to be processed further for subsequent analyses. Background This protocol was developed in order to identify regions in the genome that contain dA6m. It can be used to detect dA6m in different genomes. As a guideline, this protocol was established from existing approaches used to detect adenosine methylation in RNA (Dominissini et al., 2013). We developed this protocol and adapted it for the detection of dA6m in DNA, rather than detecting adenosine methylation RNA. This was required, as no protocol was available at that time to allow the genome-wide identification of dA6m in eukaryotic DNA. PMID:28180135

  1. A 45-kDa acetylcholinesterase protoxin of Aeromonas hydrophila: purification and immunogenicity in fish.

    PubMed

    Pérez, M J; Rodríguez, L A; Fernández-Briera, A; Nieto, T P

    2002-05-21

    A rabbit antiserum to the 15-kDa acetylcholinesterase toxin neutralised the lethal effect of the 15-kDa toxin of Aeromonas hydrophila when injected into trout. However, immunisation of fish with the 15-kDa toxoid failed to induce an antibody response, and a higher molecular mass form of this toxin was purified from the extracellular products with the aim of inducing an immune response in fish. The optimal conditions for production of extracellular products by A. hydrophila strain B32 were studied to increase the concentration of this protoxin. The extracellular products were fractionated by molecular exclusion chromatography to yield a purified protoxin with an estimated molecular mass of 45 kDa by SDS-PAGE and which gave a positive reaction in Western blotting with the rabbit anti-15-kDa toxin serum. Since the 45-kDa protoxin showed lower specific acetylcholinesterase activity than the active 15-kDa toxin, the behaviour of the active site was studied using specific inhibitors. This 45-kDa protoxin was 13.3-fold less toxic than the 15-kDa toxin and induced antibody production in fish.

  2. Evolution of the ATLAS PanDA workload management system for exascale computational science

    NASA Astrophysics Data System (ADS)

    Maeno, T.; De, K.; Klimentov, A.; Nilsson, P.; Oleynik, D.; Panitkin, S.; Petrosyan, A.; Schovancova, J.; Vaniachine, A.; Wenaus, T.; Yu, D.; Atlas Collaboration

    2014-06-01

    An important foundation underlying the impressive success of data processing and analysis in the ATLAS experiment [1] at the LHC [2] is the Production and Distributed Analysis (PanDA) workload management system [3]. PanDA was designed specifically for ATLAS and proved to be highly successful in meeting all the distributed computing needs of the experiment. However, the core design of PanDA is not experiment specific. The PanDA workload management system is capable of meeting the needs of other data intensive scientific applications. Alpha-Magnetic Spectrometer [4], an astro-particle experiment on the International Space Station, and the Compact Muon Solenoid [5], an LHC experiment, have successfully evaluated PanDA and are pursuing its adoption. In this paper, a description of the new program of work to develop a generic version of PanDA will be given, as well as the progress in extending PanDA's capabilities to support supercomputers and clouds and to leverage intelligent networking. PanDA has demonstrated at a very large scale the value of automated dynamic brokering of diverse workloads across distributed computing resources. The next generation of PanDA will allow other data-intensive sciences and a wider exascale community employing a variety of computing platforms to benefit from ATLAS' experience and proven tools.

  3. Effects of DA-Phen, a dopamine-aminoacidic conjugate, on alcohol intake and forced abstinence.

    PubMed

    Sutera, Flavia Maria; De Caro, Viviana; Cannizzaro, Carla; Giannola, Libero Italo; Lavanco, Gianluca; Plescia, Fulvio

    2016-09-01

    The mesolimbic dopamine (DA) system plays a key role in drug reinforcement and is involved in the development of alcohol addiction. Manipulation of the DAergic system represents a promising strategy to control drug-seeking behavior. Previous studies on 2-amino-N-[2-(3,4-dihydroxy-phenyl)-ethyl]-3-phenyl-propionamide (DA-Phen) showed in vivo effects as a DA-ergic modulator. This study was aimed at investigate DA-Phen effects on operant behavior for alcohol seeking behavior, during reinstatement following subsequent periods of alcohol deprivation. For this purpose, male Wistar rats were tested in an operant paradigm of self-administration; behavioral reactivity and anxiety like-behavior during acute abstinence were evaluated. A characterization of DA-Phen CNS targeting by its quantification in the brain was also carried out. Our findings showed that DA-Phen administration was able to reduce relapse in alcohol drinking by 50% and reversed the alterations in behavioral reactivity and emotionality observed during acute abstinence. In conclusion, DA-Phen can reduce reinstatement of alcohol drinking in an operant-drinking paradigm following deprivation periods and reverse abstinence-induced behavioral phenotype. DA-Phen activity seems to be mediated by the modulation of the DAergic transmission. However further studies are needed to characterize DA-Phen pharmacodynamic and pharmacokinetic properties, and its potential therapeutic profile in alcohol addiction.

  4. PX-52, A novel inhibitor of 14 kDa secretory and 85 kDa cytosolic phospholipases A2.

    PubMed

    Franson, R C; Rosenthal, M D

    1997-01-01

    Previously we reported that PGBx, a prostaglandin oligomer with anti-inflammatory activity, inhibited 14 kDa phospholipase A2 (PLA2) activity and blocked arachidonic acid mobilization in prelabeled human neutrophils (Biochim. Biophys. Acta 1006:272-277, 278-286, 1989) This study describes a new inhibitor of phospholipase A2, PX-52, that also blocks agonist induced arachidonic acid mobilization in prelabeled cells. PX-52, a fatty acid polymer, inhibited hydrolysis of 14C-oleate labeled E.coli by a variety of 14 kDa PLA2s including human PMN, sperm, synovial fluid and disc, as well as porcine pancreas, N. naja, and bee venom in a dose-dependent manner with IC50s ranging from 1.0-3.7 uM. Inhibition of activity was comparable at different Ca2+ concentrations, but was relieved by increasing substrate concentration or by methylation of PX-52. Hydrolysis of [14C]-arachidonyl phosphatidylcholine by 85 kDa, cytosolic PLA2 from U937 cells was similarly inhibited by PX-52, the IC50 = 5 uM. Arachidonic acid mobilization induced by A23187 in prelabeled human PMNs was blocked by PX-52; IC50 = 10-15 uM while concentrations of up to 80 uM oleate had no effect. These results demonstrate that PX-52 inhibits the in vitro activity of secretory and cytosolic PLA2s and agonist-induced arachidonic acid release from human cells. Given its ability to block the arachidonic acid cascade, PX-52 may be useful in the control of inflammation.

  5. Purification and characterization of human 72-kDa gelatinase (type IV collagenase). Use of immunolocalisation to demonstrate the non-coordinate regulation of the 72-kDa and 95-kDa gelatinases by human fibroblasts.

    PubMed

    Hipps, D S; Hembry, R M; Docherty, A J; Reynolds, J J; Murphy, G

    1991-04-01

    Human gingival fibroblast gelatinase (type IV collagenase) has been purified to homogeneity using a combination of ion exchange chromatography, gel filtration and affinity chromatography. The purified proenzyme electrophoresed under reducing conditions as a single band of 72 kDa which could be activated to a species of 65 kDa. Gelatinase was activated by organomercurials by a process apparently initiated by a conformational change and involving self-cleavage. It was not activated by trypsin or plasmin unlike the other family members, collagenase and stromelysin. Gelatinase otherwise exhibited properties typical of the metalloproteinases: it was inhibited by metal chelating agents and by the specific inhibitor TIMP (tissue inhibitor of metalloproteinases). Its major substrate was shown to be denatured collagen although it was also able to degrade native type IV and V collagens. A polyclonal antibody was raised in a sheep using the purified enzyme as antigen. The antiserum recognised and specifically inhibited the 72-kDa gelatinase but not a 95-kDa gelatinase from pig leukocytes. It was used in immunolocalisation studies on human fibroblasts to investigate the regulation of the production of the two Mr forms of gelatinase. These studies clearly demonstrate that human fibroblasts constitutively synthesize and secrete 72-kDa gelatinase but that 95-kDa gelatinase was inducible by agents such as cytokines. The significance of these results in relation to the likely in vivo rôle of gelatinases is discussed.

  6. The complex impact structure Serra da Cangalha, Tocantins State, Brazil

    NASA Astrophysics Data System (ADS)

    Kenkmann, Thomas; Vasconcelos, Marcos A. R.; Crósta, Alvaro P.; Reimold, Wolf U.

    2011-06-01

    Serra da Cangalha is a complex impact structure with a crater diameter of 13,700 m and a central uplift diameter of 5800 m. New findings of shatter cones, planar fractures, feather features, and possible planar deformation features are presented. Several ring-like features that are visible on remote sensing imagery are caused by selective erosion of tilted strata. The target at Serra da Cangalha is composed of Devonian to Permian sedimentary rocks, mainly sandstones that are interlayered with siltstone and claystones. NNE-SSW and WNW-ESE-striking joint sets were present prior to the impact and also overprinted the structure after its formation. As preferred zones of weakness, these joint sets partly controlled the shape of the outer perimeter of the structure and, in particular, affected the deformation within the central uplift. Joints in radial orientation to the impact center did not undergo a change in orientation during tilting of strata when the central uplift was formed. These planes were used as major displacement zones. The asymmetry of the central uplift, with preferred overturning of strata in the northern to western sector, may suggest a moderately oblique impact from a southerly direction. Buckle folding of tilted strata, as well as strata overturning, indicates that the central uplift became gravitationally unstable at the end of crater formation.

  7. Chandra Confirmation of a Pulsar Wind Nebula in DA 495

    NASA Technical Reports Server (NTRS)

    Arzoumanian, Z.; Safi-Harb, S.; Landecker, T.L.; Kothes, R.; Camilo, F.

    2008-01-01

    As part of a multiwavelength study of the unusual radio supernova remnant DA 495, we present observations made with the Chandra X-ray Observatory. Imaging and spectroscopic analysis confirms the previously detected X-ray source at the heart of the annular radio nebula, establishing the radiative properties of two key emission components: a soft unresolved source with a blackbody temperature of 1 MK consistent with a neutron star, surrounded by a nontherma1 nebula 40" in diameter exhibiting a power-law spectrum with photon index Gamma = 1.63, typical of a pulsar wind nebula. Morphologically, the nebula appears to be slightly extended along a direction, in projection on the sky, previously demonstrated to be of significance in radio and ASCA observations; we argue that this represents the orientation of the pulsar spin axis. At smaller scales, a narrow X-ray feature is seen extending out 5" from the point source, but energetic arguments suggest that it is not the resolved termination shock of the pulsar wind against the ambient medium. Finally, we argue based on synchrotron lifetimes in the nebular magnetic field that DA 495 represents the first example of a pulsar wind nebula in which electromagnetic flux makes up a significant part, together with particle flux, of the neutron star's wind.

  8. Calibration of Synthetic Photometry Using DA White Dwarfs

    NASA Astrophysics Data System (ADS)

    Holberg, J. B.; Bergeron, Pierre

    2006-09-01

    We have calibrated four major ground-based photometric systems with respect to the Hubble Space Telescope (HST) absolute flux scale, which is defined by Vega and four fundamental DA white dwarfs. These photometric systems include the Johnson-Kron-Cousins UBVRI, the Strömgren uvby filters, the Two Micron All Sky Survey JHKs, and the Sloan Digital Sky Survey ugriz filters. Synthetic magnitudes are calculated from model white dwarf spectra folded through the published filter response functions; these magnitudes in turn are absolutely calibrated with respect to the HST flux scale. Effective zero-magnitude fluxes and zero-point offsets of each system are determined. In order to verify the external observational consistency, as well as to demonstrate the applicability of these definitions, the synthetic magnitudes are compared with the respective observed magnitudes of larger sets of DA white dwarfs that have well-determined effective temperatures and surface gravities and span a wide range in both of these parameters.

  9. D-A and D-2 dopamine receptor function in the rabbit retina: a model for the central nervous system

    SciTech Connect

    Hensler, J.G.

    1987-01-01

    Studies were done investigating the effect of the synaptic concentration of the transmitter DA, modified by changes in the frequency of electrical field stimulation and by the DA uptake inhibitor nomifensine, on the modulation of /sup 3/H-DA release by D-2 DA autoreceptors and by melatonin receptor sites. At lower synaptic concentrations of the transmitter dopamine, D-2 DA receptor agonists were more potent, while antagonists were more potent when the synaptic concentration of transmitter was higher. The potency of melatonin to inhibit DA release was not altered by the frequency of field stimulation of by frequency-dependent changes in the synaptic concentration of the transmitter.

  10. Axonal patterns and targets of dA1 interneurons in the chick hindbrain.

    PubMed

    Kohl, Ayelet; Hadas, Yoav; Klar, Avihu; Sela-Donenfeld, Dalit

    2012-04-25

    Hindbrain dorsal interneurons that comprise the rhombic lip relay sensory information and coordinate motor outputs. The progenitor dA1 subgroup of interneurons, which is formed along the dorsal-most region of the caudal rhombic lip, gives rise to the cochlear and precerebellar nuclei. These centers project sensory inputs toward upper-brain regions. The fundamental role of dA1 interneurons in the assembly and function of these brainstem nuclei is well characterized. However, the precise en route axonal patterns and synaptic targets of dA1 interneurons are not clear as of yet. Novel genetic tools were used to label dA1 neurons and trace their axonal trajectories and synaptic connections at various stages of chick embryos. Using dA1-specific enhancers, two contralateral ascending axonal projection patterns were identified; one derived from rhombomeres 6-7 that elongated in the dorsal funiculus, while the other originated from rhombomeres 2-5 and extended in the lateral funiculus. Targets of dA1 axons were followed at later stages using PiggyBac-mediated DNA transposition. dA1 axons were found to project and form synapses in the auditory nuclei and cerebellum. Investigation of mechanisms that regulate the patterns of dA1 axons revealed a fundamental role of Lim-homeodomain (HD) proteins. Switch in the expression of the specific dA1 Lim-HD proteins Lhx2/9 into Lhx1, which is typically expressed in dB1 interneurons, modified dA1 axonal patterns to project along the routes of dB1 subgroup. Together, the results of this research provided new tools and knowledge to the assembly of trajectories and connectivity of hindbrain dA1 interneurons and of molecular mechanisms that control these patterns.

  11. Immunoreactivity of the Mycobacterium avium subsp. paratuberculosis 19-kDa lipoprotein

    PubMed Central

    Huntley, Jason FJ; Stabel, Judith R; Bannantine, John P

    2005-01-01

    Background The Mycobacterium tuberculosis 19-kDa lipoprotein has been reported to stimulate both T and B cell responses as well as induce a number of Th1 cytokines. In order to evaluate the Mycobacterium avium subsp. paratuberculosis (M. avium subsp. paratuberculosis) 19-kDa lipoprotein as an immunomodulator in cattle with Johne's disease, the gene encoding the 19-kDa protein (MAP0261c) was analyzed. Results MAP0261c is conserved in mycobacteria, showing a 95% amino acid identity in M. avium subspecies avium, 84% in M. intracellulare and 76% in M. bovis and M. tuberculosis. MAP0261c was cloned, expressed, and purified as a fusion protein with the maltose-binding protein (MBP-19 kDa) in Escherichia coli. IFN-γ production was measured from 21 naturally infected and 9 control cattle after peripheral blood mononuclear cells (PBMCs) were stimulated with a whole cell lysate (WCL) of M. avium subsp. paratuberculosis or the recombinant MBP-19 kDa. Overall, the mean response to MBP-19 kDa was not as strong as the mean response to the WCL. By comparison, cells from control, non-infected cattle did not produce IFN-γ after stimulation with either WCL or MBP-19 kDa. To assess the humoral immune response to the 19-kDa protein, sera from cattle with clinical Johne's disease were used in immunoblot analysis. Reactivity to MBP-19 kDa protein, but not MBP alone, was observed in 9 of 14 infected cattle. Antibodies to the 19-kDa protein were not observed in 8 of 9 control cows. Conclusions Collectively, these results demonstrate that while the 19-kDa protein from M. avium subsp. paratuberculosis stimulates a humoral immune response and weak IFN-γ production in infected cattle, the elicited responses are not strong enough to be used in a sensitive diagnostic assay. PMID:15663791

  12. The 37kDa/67kDa Laminin Receptor acts as a receptor for Aβ42 internalization

    PubMed Central

    Da Costa Dias, Bianca; Jovanovic, Katarina; Gonsalves, Danielle; Moodley, Kiashanee; Reusch, Uwe; Knackmuss, Stefan; Weinberg, Marc S.; Little, Melvyn; Weiss, Stefan F. T.

    2014-01-01

    Neuronal loss is a major neuropathological hallmark of Alzheimer's disease (AD). The associations between soluble Aβ oligomers and cellular components cause this neurotoxicity. The 37 kDa/67 kDa laminin receptor (LRP/LR) has recently been implicated in Aβ pathogenesis. In this study the mechanism underlying the pathological role of LRP/LR was elucidated. Försters Resonance Energy Transfer (FRET) revealed that LRP/LR and Aβ form a biologically relevant interaction. The ability of LRP/LR to form stable associations with endogenously shed Aβ was confirmed by pull down assays and Aβ-ELISAs. Antibody blockade of this association significantly lowered Aβ42 induced apoptosis. Furthermore, antibody blockade and shRNA mediated downregulation of LRP/LR significantly hampered Aβ42 internalization. These results suggest that LRP/LR is a receptor for Aβ42 internalization, mediating its endocytosis and contributing to the cytotoxicity of the neuropeptide by facilitating intra-cellular Aβ42 accumulation. These findings recommend anti-LRP/LR specific antibodies and shRNAs as potential therapeutic tools for AD treatment. PMID:24990253

  13. Foxa2 acts as a co-activator potentiating expression of the Nurr1-induced DA phenotype via epigenetic regulation.

    PubMed

    Yi, Sang-Hoon; He, Xi-Biao; Rhee, Yong-Hee; Park, Chang-Hwan; Takizawa, Takumi; Nakashima, Kinichi; Lee, Sang-Hun

    2014-02-01

    Understanding how dopamine (DA) phenotypes are acquired in midbrain DA (mDA) neuron development is important for bioassays and cell replacement therapy for mDA neuron-associated disorders. Here, we demonstrate a feed-forward mechanism of mDA neuron development involving Nurr1 and Foxa2. Nurr1 acts as a transcription factor for DA phenotype gene expression. However, Nurr1-mediated DA gene expression was inactivated by forming a protein complex with CoREST, and then recruiting histone deacetylase 1 (Hdac1), an enzyme catalyzing histone deacetylation, to DA gene promoters. Co-expression of Nurr1 and Foxa2 was established in mDA neuron precursor cells by a positive cross-regulatory loop. In the presence of Foxa2, the Nurr1-CoREST interaction was diminished (by competitive formation of the Nurr1-Foxa2 activator complex), and CoREST-Hdac1 proteins were less enriched in DA gene promoters. Consequently, histone 3 acetylation (H3Ac), which is responsible for open chromatin structures, was strikingly increased at DA phenotype gene promoters. These data establish the interplay of Nurr1 and Foxa2 as the crucial determinant for DA phenotype acquisition during mDA neuron development.

  14. Tuning PEG-DA hydrogel properties via solvent-induced phase separation (SIPS)†

    PubMed Central

    Bailey, Brennan Margaret; Hui, Vivian; Fei, Ruochong

    2012-01-01

    Poly(ethylene glycol) diacrylate (PEG-DA) hydrogels are widely utilized to probe cell-material interactions and ultimately for a material-guided approach to tissue regeneration. In this study, PEG-DA hydrogels were fabricated via solvent-induced phase separation (SIPS) to obtain hydrogels with a broader range of tunable physical properties including morphology (e.g. porosity), swelling and modulus (G′). In contrast to conventional PEG-DA hydrogels prepared from an aqueous precursor solution, the reported SIPS protocol utilized a dichloromethane (DCM) precursor solution which was sequentially photopolymerized, dried and hydrated. Physical properties were further tailored by varying the PEG-DA wt% concentration (5 wt%–25 wt%) and Mn (3.4k and 6k g mol −1). SIPS produced PEG-DA hydrogels with a macroporous morphology as well as increased G′ values versus the corresponding conventional PEG-DA hydrogels. Notably, since the total swelling was not significantly changed versus the corresponding conventional PEG-DA hydrogels, pairs or series of hydrogels represent scaffolds in which morphology and hydration or G′ and hydration are uncoupled. In addition, PEG-DA hydrogels prepared via SIPS exhibited enhanced degradation rates. PMID:22956857

  15. 40 CFR 60.44Da - Standard for nitrogen oxides (NOX).

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 6 2010-07-01 2010-07-01 false Standard for nitrogen oxides (NOX). 60.44Da Section 60.44Da Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS... for nitrogen oxides (NOX). (a) On and after the date on which the initial performance test is...

  16. Paśu Ayurvĕda (veterinary medicine) in Garudapurăņa.

    PubMed

    Varanasi, Subhose; Narayana, A

    2007-01-01

    The history of veterinary medicine is closely tied to the development of human medicine. Evidence of animal medicine has been found in ancient civilizations, such as those of the Hindu, Babylonians, Hebrews, Arabs, Greeks, and Romans. Ancient Indian literature in the form of the holy Vĕda, Purăna, Brăhmaņa, epics, etc. is flooded with information on animal care. The Purăņa are ancient scriptures discuss varied topics like devotion to God and his various aspects, traditional sciences like Ayurvĕda, Jyŏtişa (Astrology), cosmology, concepts like dharma, karma, reincarnation and many others. The treatment of animal diseases using Ayurvedic medicine has been mentioned in Garudapurăna, Agnipurăņa, Atri-samhită, Matsyapurăņa and many other texts. The Garudapurăņa is one of the important Săttvika purăna, the subject matter is divided into two parts, viz. Pŭrvakhaņda (first part) and an Uttarakhaņda (subsequent part). Gavăyurvĕda, Gajăyurvĕda narrated briefly and Aśvăyurvĕda described detailly in Pŭrvakhaņda.

  17. 32 CFR Appendix F to Part 623 - Power of Attorney (DA Form 4881-4-R)

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 3 2011-07-01 2009-07-01 true Power of Attorney (DA Form 4881-4-R) F Appendix F to Part 623 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY SUPPLIES AND EQUIPMENT LOAN OF ARMY MATERIEL Pt. 623, App. F Appendix F to Part 623—Power of Attorney (DA Form...

  18. 32 CFR Appendix E to Part 623 - Surety Bond (DA Form 4881-3-R)

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true Surety Bond (DA Form 4881-3-R) E Appendix E to Part 623 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY SUPPLIES AND EQUIPMENT LOAN OF ARMY MATERIEL Pt. 623, App. E Appendix E to Part 623—Surety Bond (DA Form...

  19. 32 CFR Appendix E to Part 623 - Surety Bond (DA Form 4881-3-R)

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 3 2013-07-01 2013-07-01 false Surety Bond (DA Form 4881-3-R) E Appendix E to Part 623 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY SUPPLIES AND EQUIPMENT LOAN OF ARMY MATERIEL Pt. 623, App. E Appendix E to Part 623—Surety Bond (DA Form 4881-3-R...

  20. 32 CFR Appendix E to Part 623 - Surety Bond (DA Form 4881-3-R)

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 3 2014-07-01 2014-07-01 false Surety Bond (DA Form 4881-3-R) E Appendix E to Part 623 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY SUPPLIES AND EQUIPMENT LOAN OF ARMY MATERIEL Pt. 623, App. E Appendix E to Part 623—Surety Bond (DA Form 4881-3-R...

  1. 32 CFR Appendix E to Part 623 - Surety Bond (DA Form 4881-3-R)

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 3 2012-07-01 2009-07-01 true Surety Bond (DA Form 4881-3-R) E Appendix E to Part 623 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY SUPPLIES AND EQUIPMENT LOAN OF ARMY MATERIEL Pt. 623, App. E Appendix E to Part 623—Surety Bond (DA Form 4881-3-R...

  2. Augmenting Effect of DA-9601 on Ghrelin in an Acute Gastric Injury Model

    PubMed Central

    Baek, Yoo Hum; Lee, Kang Nyeong; Jun, Dae Won; Yoon, Byung Chul; Kim, Ju Mi; Oh, Tae Young

    2011-01-01

    Background/Aims Acute gastric injury by alcohol or indomethacin has been reported to be prevented by DA-9601, an extract of the herb Artemisia asiatica. Ghrelin, an endogenously produced gastrointestinal peptide hormone, has also been demonstrated to play a role in gastric mucosal defense. The aim of this study was to investigate the effects of DA-9601 on ghrelin in an acute gastric injury model induced by alcohol or indomethacin. Methods A total of 140 Sprague-Dawley rats were divided into two groups, a placebo group and a DA-9601-pretreated group. Thirty minutes later, half of the rats in each group received ethanol injury and the other half received indomethacin injury. Levels of serum ghrelin and gastric mucosal ghrelin mRNA were measured by ELISA and RT-PCR, respectively. Results Immediately after ethanol administration, ghrelin increased in both groups pretreated with DA-9601 and placebo. However, the increase occurred more rapidly and was higher in the DA-9601-pretreated rats than in the controls that did not receive DA-9601-pretreatment. Similarly, from 30 minutes to 2 hours after indomethacin administration, the DA-9601-pretreated rats showed a significant increase in serum and gastric mucosal ghrelin concentrations, whereas placebo-pretreated rats showed only a mild increase. Conclusions DA-9601 potentiates the endogenous production and secretion of ghrelin in acute gastric injury models induced by ethanol or indomethacin. PMID:21461072

  3. A Creative Approach to the Common Core Standards: The Da Vinci Curriculum

    ERIC Educational Resources Information Center

    Chaucer, Harry

    2012-01-01

    "A Creative Approach to the Common Core Standards: The Da Vinci Curriculum" challenges educators to design programs that boldly embrace the Common Core State Standards by imaginatively drawing from the genius of great men and women such as Leonardo da Vinci. A central figure in the High Renaissance, Leonardo made extraordinary contributions as a…

  4. A Proposal to Build Evaluation Capacity at the Bunche-Da Vinci Learning Partnership Academy

    ERIC Educational Resources Information Center

    King, Jean A.

    2005-01-01

    The author describes potential evaluation capacity-building activities in contrast to the specifics of an evaluation design. Her response to the case of the Bunche-Da Vinci Learning Partnership Academy is developed in three parts: (1) an initial framing of the Bunche-Da Vinci situation; (2) what should be done before signing a contract; and (3)…

  5. A Creative Approach to the Common Core Standards: The Da Vinci Curriculum

    ERIC Educational Resources Information Center

    Chaucer, Harry

    2012-01-01

    "A Creative Approach to the Common Core Standards: The Da Vinci Curriculum" challenges educators to design programs that boldly embrace the Common Core State Standards by imaginatively drawing from the genius of great men and women such as Leonardo da Vinci. A central figure in the High Renaissance, Leonardo made extraordinary contributions as a…

  6. 32 CFR Appendix E to Part 623 - Surety Bond (DA Form 4881-3-R)

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 3 2011-07-01 2009-07-01 true Surety Bond (DA Form 4881-3-R) E Appendix E to Part 623 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY SUPPLIES AND EQUIPMENT LOAN OF ARMY MATERIEL Pt. 623, App. E Appendix E to Part 623—Surety Bond (DA Form 4881-3-R...

  7. The analysis Arabidopsis thaliana overexpressing a 14kDa self-folding protein [abstract

    USDA-ARS?s Scientific Manuscript database

    A recent study in banana identified a 14kDa protein that has been hypothesized to function in regulating the nucleation and growth of the needle-shaped crystals of calcium oxalate that accumulate within the tissues of this plant. To gain further insight in to the functional role of this 14 kDa prote...

  8. The Indicative and Subjunctive "da"-complements in Serbian A Syntactic-Semantic Approach

    ERIC Educational Resources Information Center

    Todorovic, Natasa

    2012-01-01

    A syntactic-semantic investigation of subjunctive and indicative "da"-complements in Serbian is conducted in this project. After a careful comparison of Serbian sentence constructions with "da"-complements to the equivalent sentence structures in languages of the Balkans as well as other Slavic languages, it is clearly…

  9. Tyrosine phosphorylation of two cytosolic proteins of 50 kDa and 35 kDa in rat liver by insulin-receptor kinase in vitro.

    PubMed Central

    Kwok, Y C; Yip, C C

    1987-01-01

    Insulin-receptor tyrosine kinase can phosphorylate a variety of artificial substrates in vitro. Its physiological substrate(s), however, remains unknown. In the present study, we show that immobilized insulin receptors phosphorylate tyrosine residues of two cytosolic proteins of 50 kDa and 35 kDa in rat liver. Phosphorylation of these two proteins required Mn2+- or Mg2+-ATP as the phosphate donor. Phosphorylation was time- and temperature-dependent. Furthermore, the rate of phosphorylation of the two proteins was related to the autophosphorylated state of the insulin receptor. The pI of the phosphorylated 50 kDa and 35 kDa proteins was 5.4 and 5.6 respectively. These proteins were present in low abundance. They were not related to each other, nor to the insulin receptor, as demonstrated by in-gel proteolytic digestion and by immunoprecipitation using antibodies produced against them. They were specific substrates for the insulin receptor kinase, since they were not phosphorylated by epidermal-growth-factor-receptor kinase. These observations suggest that the 50 kDa and 35 kDa cytosolic proteins may be endogenous substrates for the insulin-receptor kinase. Images Fig. 1. Fig. 2. Fig. 5. Fig. 6. Fig. 7. Fig. 8. Fig. 9. PMID:2829823

  10. DA-9701: A New Multi-Acting Drug for the Treatment of Functional Dyspepsia.

    PubMed

    Kwon, Yong Sam; Son, Miwon

    2013-05-30

    Motilitone(®) (DA-9701) is a new herbal drug that was launched for the treatment of functional dyspepsia in December 2011 in Korea. The heterogeneous symptom pattern and multiple causes of functional dyspepsia have resulted in multiple drug target strategies for its treatment. DA-9701, a compound consisting of a combination of Corydalis Tuber and Pharbitidis Semen, has being developed for treatment of functional dyspepsia. It has multiple mechanisms of action such as fundus relaxation, visceral analgesia, and prokinetic effects. Furthermore, it was found to significantly enhance meal-induced gastric accommodation and increase gastric compliance in dogs. DA-9701 also showed an analgesic effect in rats with colorectal distension induced visceral hypersensitivity and an antinociceptive effect in beagle dogs with gastric distension-induced nociception. The pharmacological effects of DA-9701 also include conventional effects, such as enhanced gastric emptying and gastrointestinal transit. The safety profi le of DA-9701 is also preferable to that of other treatments.

  11. DA-9701: A New Multi-Acting Drug for the Treatment of Functional Dyspepsia

    PubMed Central

    Kwon, Yong Sam; Son, Miwon

    2013-01-01

    Motilitone® (DA-9701) is a new herbal drug that was launched for the treatment of functional dyspepsia in December 2011 in Korea. The heterogeneous symptom pattern and multiple causes of functional dyspepsia have resulted in multiple drug target strategies for its treatment. DA-9701, a compound consisting of a combination of Corydalis Tuber and Pharbitidis Semen, has being developed for treatment of functional dyspepsia. It has multiple mechanisms of action such as fundus relaxation, visceral analgesia, and prokinetic effects. Furthermore, it was found to significantly enhance meal-induced gastric accommodation and increase gastric compliance in dogs. DA-9701 also showed an analgesic effect in rats with colorectal distension induced visceral hypersensitivity and an antinociceptive effect in beagle dogs with gastric distension-induced nociception. The pharmacological effects of DA-9701 also include conventional effects, such as enhanced gastric emptying and gastrointestinal transit. The safety profi le of DA-9701 is also preferable to that of other treatments. PMID:24265862

  12. The 10 kDa protein of Taenia solium metacestodes shows genus specific antigenicity.

    PubMed

    Park, S K; Yun, D H; Chung, J Y; Kong, Y; Cho, S Y

    2000-09-01

    Genus specific antigenicity of the 10 kDa protein in cyst fluid (CF) of Taenia solium metacestodes was demonstrated by comparative immunoblot analysis. When CFs from taeniid metacestodes of T. saginata, T. solium, T. taeniaeformis and T. crassiceps were probed with specific monoclonal antibody (mAb) raised against 150 kDa protein of T. solium metacestodes, specific antibody reactions were observed in 7 and 10 kDa proteins of T. solium and in 7/8 kDa of T. saginata, T. taeniaeformis and T. crassiceps. The mAb did not react with any protein in hydatid fluid of Echinococcus granulosus and E. multilocularis. This result revealed that the 10 kDa peptide of T. solium metacestodes and its equivalent proteins of different Taenia metacestodes are genus specific antigens that are shared among different Taenia species.

  13. Growth-Promoting Hormone DA-6 Assists Phytoextraction and Detoxification of Cd by Ryegrass.

    PubMed

    He, Shanying; Wu, Qiuling; He, Zhenli

    2015-01-01

    A pot experiment was carried out to study the effect of growth-promoting hormone diethyl aminoethyl hexanoate (DA-6) on Cd phytoextraction and detoxification in ryegrass. Foliar spray of DA-6 significantly enhanced Cd extraction efficiency (P<0.05), with 1 μM DA-6 the most effective. At the subcellular level, 43-53% of Cd was soluble fraction and 23-46% in cell wall, and 9-25% in organelles. Chemical speciation analysis showed that 52.7-58.5% of Cd was NaCl extractable, 12.1-22.7% ethanol extractable, followed by other fractions. DA-6 alleviated metal toxicity by fixing more Cd in cell wall and decreasing Cd migration in plant. In conclusion, ryegrass tolerates Cd by cell wall compartmentalization along with protein and organic acids combination, and the treatment of 1 μM DA-6 appears to be optimal for enhancing the remediation efficiency of ryegrass for Cd contaminated soil.

  14. Xanthelasma and lipoma in Leonardo da Vinci's Mona Lisa.

    PubMed

    Dequeker, Jan; Muls, Erik; Leenders, Kathleen

    2004-08-01

    The painting Mona Lisa in the Louvre, Paris, by Leonardo da Vinci (1503-1506), shows skin alterations at the inner end of the left upper eyelid similar to xanthelasma, and a swelling of the dorsum of the right hand suggestive of a subcutaneous lipoma. These findings in a 25-30 year old woman, who died at the age of 37, may be indicative of essential hyperlipidemia, a strong risk factor for ischemic heart disease in middle age. As far as is known, this portrait of Mona Lisa painted in 1506 is the first evidence that xanthelasma and lipoma were prevalent in the sixteenth century, long before the first description by Addison and Gall in 1851.

  15. Leonardo da Vinci: the search for the soul.

    PubMed

    Del Maestro, R F

    1998-11-01

    The human race has always contemplated the question of the anatomical location of the soul. During the Renaissance the controversy crystallized into those individuals who supported the heart ("cardiocentric soul") and others who supported the brain ("cephalocentric soul") as the abode for this elusive entity. Leonardo da Vinci (1452-1519) joined a long list of other explorers in the "search for the soul." The method he used to resolve this anatomical problem involved the accumulation of information from ancient and contemporary sources, careful notetaking, discussions with acknowledged experts, and his own personal search for the truth. Leonardo used a myriad of innovative methods acquired from his knowledge of painting, sculpture, and architecture to define more clearly the site of the "senso comune"--the soul. In this review the author examines the sources of this ancient question, the knowledge base tapped by Leonardo for his personal search for the soul, and the views of key individuals who followed him.

  16. [Regarding the Manuscript D " Dell' occhio " of Leonardo da Vinci].

    PubMed

    Heitz, Robert F

    2009-01-01

    Leonardo da Vinci's Manuscript D consists of five double pages sheets, which, folded in two, comprise ten folios. This document, in the old Tuscan dialect and mirror writing, reveals the ideas of Leonardo on the anatomy of the eye in relation to the formation of images and visual perception. Leonardo explains in particular the behavior of the rays in the eye in terms of refraction and reflection, and is very mechanistic in his conception of the eye and of the visual process. The most significant innovations found in these folios are the concept of the eye as a camera obscura and the intersection of light rays in the interior of the eye. His texts nevertheless show hesitation, doubts and a troubled confusion, reflecting the ideas and uncertainties of his era. He did not share his results in his lifetime, despite both printing and etching being readily available to him.

  17. Sine ars scientia nihil est: Leonardo da Vinci and beyond.

    PubMed

    Kickhöfel, Eduardo H P

    2009-01-01

    The aim of this article is to reflect on the relationship between art and science so far as it concerns a symposium on neurosciences. We undertake a historical overview of that relationship, paying particular attention to the sui generis case of Leonardo da Vinci, who very often is regarded as the man who worked on art and science with equal ease. We then explain why his idea of merging these two forms of knowledge failed, considering the clear-cut distinction between art and science in his time. With this clarification, we explore the matter today. We look at Raphael's The Transfiguration, in which the representation of the possessed boy is seen by neuroscientists as indicative of an epileptic seizure. We also look at the ideas of neuroscientists Semir Zeki and Vilayanur Ramachandran, who study particular aspects of brain function and suggest a new merging of art and science.

  18. A 92-kDa human immunostimulatory protein.

    PubMed Central

    Fontan, E; Briend, E; Saklani-Jusforgues, H; d'Alayer, J; Vandekerckhove, J; Fauve, R M

    1994-01-01

    We purified to apparent homogeneity a human urinary glycoprotein of 92 kDa (HGP.92) that, administered intravenously at 250 micrograms/kg, fully protected mice against a lethal inoculum of Listeria monocytogenes. Since HGP.92 protected scid mice, which lack B and T lymphocytes, this increased resistance to Listeria did not appear to be lymphocyte mediated. Furthermore, inflammatory macrophages incubated with 6 nM HGP.92 inhibited the growth of Lewis carcinoma cells in vitro. These two activities appeared to depend on an oligosaccharide moiety, as they were lost after N-Glycanase treatment of HGP.92. Thus, the biological activity of HGP.92 was in some way related to a glycan moiety. Images PMID:8078887

  19. VLBI Radar of the 2012 DA14 Asteroid

    NASA Astrophysics Data System (ADS)

    Nechaeva, M. B.; Dugin, N. A.; Antipenko, A. A.; Bezrukov, D. A.; Bezrukov, V. V.; Voytyuk, V. V.; Dement'ev, A. F.; Jekabsons, N.; Klapers, M.; Konovalenko, A. A.; Kulishenko, V. F.; Nabatov, A. S.; Nesteruk, V. N.; Putillo, D.; Reznichenko, A. M.; Salerno, E.; Snegirev, S. D.; Tikhomirov, Yu. V.; Khutornoy, R. V.; Skirmante, K.; Shmeld, I.; Chagunin, A. K.

    2015-03-01

    An experiment on VLBI radar of the 2012 DA14 asteroid was carried out on February 15-16, 2011 at the time of its closest approach to the Earth. The research teams of Kharkov (Institute of Radio Astronomy of the National Academy of Sciences of Ukraine), Evpatoria (National Space Facilities Control and Test Center), Nizhny Novgorod (Radiophysical Research Institute), Bologna (Istituto di Radioastronomia (INAF)), and Ventspils (Ventspils International Radioastronomy Center) took part in the experiment. The asteroid was irradiated by the RT-70 planetary radar (Evpatoria) at a frequency of 5 GHz. The reflected signal was received using two 32-m radio telescopes in Medicina (Italy) and Irbene (Latvia) in radiointerferometric mode. The Doppler frequency shifts in bi-static radar mode and interference frequency in VLBI mode were measured. Accuracy of the VLBI radar method for determining the radial and angular velocities of the asteroid were estimated.

  20. Discovery of Photospheric Germanium in Hot DA White Dwarfs

    NASA Astrophysics Data System (ADS)

    Vennes, Stéphane; Chayer, Pierre; Dupuis, Jean

    2005-04-01

    We report the identification of Ge IV resonance lines in ultraviolet spectra of the hot DA white dwarfs Feige 24, G191-B2B, and GD 246. The lines originate in the stellar photosphere, and we measure low Ge/H abundance ratios ranging between -8.0 and -8.7. We also tentatively identify a resonance line of Sn IV blended with an Fe V line in the spectrum of G191-B2B. The presence of germanium extends our knowledge of the abundance pattern in hot white dwarfs beyond the iron group. The abundance ratio appears nearly solar, which implies either that the germanium abundance mixture in these stars has remained unaltered since leaving the main sequence or that diffusion processes (e.g., selective radiation pressure) are coincidentally reproducing a solar Ge/H ratio.