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Sample records for simulated pwr primary

  1. 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.

  2. 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.

  3. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  5. The Effects of Metallurgical Factors on PWSCC Crack Growth Rates in TT Alloy 690 in Simulated PWR Primary Water

    NASA Astrophysics Data System (ADS)

    Yonezawa, Toshio; Watanabe, Masashi; Hashimoto, Atsushi

    2015-06-01

    Primary water stress corrosion cracking growth rates (PWSCCGRs) in highly cold-worked thermally treated (TT) Alloy 690 have been recently reported as exhibiting significant heat-to-heat variability. Authors hypothesized that these significant differences could be due to the metallurgical characteristics of each heat. In order to confirm this hypothesis, the effect of fundamental metallurgical characteristics on PWSCCGR measurements in cold-worked TT Alloy 690 has been investigated. The following new observations were made in this study: (1) Microcracks and voids were observed in or near eutectic crystals of grain boundary (GB) M23C6 carbides (primary carbides) after cold rolling, but were not observed before cold rolling. These primary carbides with microcracks and voids were observed in both lightly forged and as-cast and cold-rolled TT Alloy 690 (heat A) as well as in a cold-rolled TT Alloy 690 (heat Y) that simulated the chemical composition and carbide banded structure of the material previously tested by Paraventi and Moshier. However, this was not observed in precipitated (secondary) M23C6 GB carbides in heavily forged and cold-rolled TT Alloy 690 heat A and a cold-rolled commercial TT Alloy 690. (2) From microstructural analyses carried out on the various TT Alloy 690 test materials before and after cold rolling, the amount of eutectic crystals (primary carbides and nitrides) M23C6 and TiN depended on the chemical composition. In particular, the amount of M23C6 depended on the fabrication process. Microcracks and voids in or near the M23C6 and TiN precipitates were generated by the cold rolling process. (3) The PWSCCGRs observed in TT Alloy 690 were different for each heat and fabrication process. The PWSCCGR decreased with increasing Vickers hardness of each heat. However, for the same heats and fabrication processes, the PWSCCGR increased with increasing Vickers hardness due to cold work. Thus, the PWSCCGR must be affected not only by hardness (or

  6. 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.

  7. 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.

  8. 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.

  9. VERA Core Simulator Methodology for PWR Cycle Depletion

    SciTech Connect

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

    2015-01-01

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

  10. 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.

  11. 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.

  12. 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.

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

    SciTech Connect

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

    1981-11-01

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

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

    SciTech Connect

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

    1990-01-01

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

  15. 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.

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

    SciTech Connect

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

    1980-01-01

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

  17. 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.

  18. 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.

  19. 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).

  20. 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.

  1. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2013-09-01

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

  3. 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.

  4. 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)

  5. 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

    In the general context of the nuclear activities, life extension of the existing plants is the interesting option for countries that are already well equipped with NPPs. As the working life of 60 years is now expected possible for some well maintained plants, their safety measures needs to be improved such that they should be comparable to the new or future designs, taken into account the results from the probabilistic and the deterministic accident analysis. To accomplish this aim, the Accident Management (AM) is the important part of the process that must be utilized including possible automation of some processes. At INSTN, the extensive sensitivity studies related to the feed and bleed process on the primary and the secondary side had been carried out with the SIPACT simulator, based on the Cathare code, for a 900 MWe pressurized water reactor. The simulations had been mainly conducted for the Beyond Design Basis Accident (BDBA) condition. This condition included the total loss of feed-water and a small break with the loss of the high pressure injection system (HPIS). From these studies, several interesting findings had been obtained. For AM purpose and with the bleeding process, the criterion called 'the safety time margin' for core uncover was introduced. By plotting the safety time margin against the bleeding time, the relation between them was established and used to optimize, when possible, the AM measures. For the scenario that involved the total loss of feed water, in case of full bleeding, a window was found for the bleeding time around the degradation of the heat exchange in SGs would be resulted. In this scenario, one of the solutions was to open only one relief valve at first in order to let through only the minimal mass. At the time of the injection by the accumulator, the other two relief valves were then opened. As a result, the flow through the relief valves could be effectively compensated by the flow from the accumulator, the mass balance in

  6. 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.

  7. 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.

  8. 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.

  9. Life Prediction and Stress Evolvement for Low Cycle Fatigue in PWR Primary Pipe Material

    NASA Astrophysics Data System (ADS)

    Fei, Xue; Wei-wei, Yu; Zhao-xi, Wang; Wen-xin, Ti; Lei, Lin; Xin-ming, Men

    2010-05-01

    The low cycle fatigue (LCF) behavior of primary pipe material Z3CN20.09M cast stainless stell (CASS) was studied at room temperature (RT) and elevated temperature of 350° C by conducting total axial stain controlled tests in air with strain amplitude in the range ±0.175% to ±0.8%. Based on the test results, the cyclic stress response of material was analyzed, and a dynamic strain aging (DSA) phenomena was discovered at 350° C. Besides, the evaluation of elastic modulus during cyclic tests was studied, and the effect of elastic modulus on parameters of low cycle fatigue was investigated based on the Manson-Coffin model. It is shown that elastic modulus for Z3CN20.09M decreases constantly during the whole fatigue life, but fluctuates more frequently at elevated temperature. Both the static and dynamic elastic modulus result in a same life trend in low cycle fatigue, but the elastic modulus affects the precision of fatigue life prediction to some extent when the fatigue life exceeded 105.

  10. 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

  11. 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

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

    SciTech Connect

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

    1987-01-01

    For a severe pressurized water reactor accident that leads to a loss of feedwater to the stream generators, such as might occur in a station blackout, fission product decay heating causes a water boil-off. Without effective decay heat removal, the fuel elements will be uncovered. Eventually, steam will oxidize the overheated cladding. The noble gases and volatile fission products, such as cesium and iodine, 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.

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

    SciTech Connect

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

    1982-01-01

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

  14. Verification of Optimal Control Strategy Search Using a Simplest 3-D PWR Xenon Oscillation Simulator

    SciTech Connect

    Shimazu, Yoichiro

    2006-07-01

    Power spatial oscillations due to the transient xenon spatial distribution are well known as xenon oscillation in large PWRs. When the reactor size becomes larger than the current design, then even radial oscillations can be also divergent. Even if the radial oscillation is convergent, when some control rods malfunction occurs, it is necessary to suppress the oscillation in as short time as possible. In such cases, optimal control strategy is required. Generally speaking the optimality search based on the modern control theory requires a lot of calculation for the evaluation of state variables. In the case of control rod malfunctions the xenon oscillation could be three dimensional. In such case, direct core calculations would be inevitable. From this point of view a very simple model, only four point reactor model, has been developed and verified. In this paper, an example of a procedure and the validity of the results for optimal control strategy search are presented by comparing it with the result by a three dimensional nuclear design code The simplest simulator can predict optimal strategy in less than 10 seconds on a PC. Thus it is recommended that a strategy generator, which is quick in analyzing and easy to use, might be installed in a monitoring system or in an operator guiding system. (author)

  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. Effects of dissolved hydrogen on general corrosion behavior and oxide films of alloy 690TT in PWR primary water

    NASA Astrophysics Data System (ADS)

    Jeon, Soon-Hyeok; Lee, Eun-Hee; Hur, Do Haeng

    2017-03-01

    The effect of dissolved hydrogen (DH) on the general corrosion behavior and oxide films of Alloy 690TT is investigated in simulated primary water at 330 °C. With increasing DH, the structure of oxide film significantly changed and the corrosion rate decreased. At DH = 5 cm3/kg H2O, the oxide layer was thick, and consisted of outer Ni oxide layer and inner Cr2O3 layer. Under the conditions of DH = 35 and 100 cm3/kg H2O, the oxide films grew thinner and composed of outer polyhedral spinel oxide particles such as NiCr2O4 or NiCrFeO4 and an intermediate metallic Ni-rich layer, with inner Cr2O3 layer. The general corrosion rate significantly decreased by about 72% as DH concentration increased from 5 to 35 cm3/kg H2O. In the range of 35-65 cm3/kg H2O, the corrosion rate slightly decreased with increasing DH concentration. However, no further changes were observed in the range of 65-100 cm3/kg H2O.

  17. 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.

  18. 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.

  19. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    SciTech Connect

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  20. STRESS CORROSION CRACK GROWTH RESPONSE FOR ALLOY 152/52 DISSIMILAR METAL WELDS IN PWR PRIMARY WATER

    SciTech Connect

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

    2015-08-15

    As part of ongoing research into primary water stress corrosion cracking (PWSCC) susceptibility of alloy 690 and its welds, SCC tests have been conducted on alloy 152/52 dissimilar metal (DM) welds with cracks positioned with the goal to assess weld dilution and fusion line effects on SCC susceptibility. No increased crack growth rate was found when evaluating a 20% Cr dilution zone in alloy 152M joined to carbon steel (CS) that had not undergone a post-weld heat treatment (PWHT). However, high SCC crack growth rates were observed when the crack reached the fusion line of that material where it propagated both on the fusion line and in the heat affected zone (HAZ) of the carbon steel. Crack surface and crack profile examinations of the specimen revealed that cracking in the weld region was transgranular (TG) with weld grain boundaries not aligned with the geometric crack growth plane of the specimen. The application of a typical pressure vessel PWHT on a second set of alloy 152/52 – carbon steel DM weld specimens was found to eliminate the high SCC susceptibility in the fusion line and carbon steel HAZ regions. PWSCC tests were also performed on alloy 152-304SS DM weld specimens. Constant K crack growth rates did not exceed 5x10-9 mm/s in this material with post-test examinations revealing cracking primarily on the fusion line and slightly into the 304SS HAZ.

  1. 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.

  2. 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)

  3. 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.

  4. 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.

  5. 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

  6. Accessing primary care: a simulated patient study

    PubMed Central

    Campbell, John L; Carter, Mary; Davey, Antoinette; Roberts, Martin J; Elliott, Marc N; Roland, Martin

    2013-01-01

    Background Simulated patient, or so-called ‘mystery-shopper’, studies are a controversial, but potentially useful, approach to take when conducting health services research. Aim To investigate the construct validity of survey questions relating to access to primary care included in the English GP Patient Survey. Design and setting Observational study in 41 general practices in rural, urban, and inner-city settings in the UK. Method Between May 2010 and March 2011, researchers telephoned practices at monthly intervals, simulating patients requesting routine, but prompt, appointments. Seven measures of access and appointment availability, measured from the mystery-shopper contacts, were related to seven measures of practice performance from the GP Patient Survey. Results Practices with lower access scores in the GP Patient Survey had poorer access and appointment availability for five out of seven items measured directly, when compared with practices that had higher scores. Scores on items from the national survey that related to appointment availability were significantly associated with direct measures of appointment availability. Patient-satisfaction levels and the likelihood that patients would recommend their practice were related to the availability of appointments. Patients’ reports of ease of telephone access in the national survey were unrelated to three out of four measures of practice call handling, but were related to the time taken to resolve an appointment request, suggesting responders’ possible confusion in answering this question. Conclusion Items relating to the accessibility of care in a the English GP patient survey have construct validity. Patients’ satisfaction with their practice is not related to practice call handling, but is related to appointment availability. PMID:23561783

  7. Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

    NASA Astrophysics Data System (ADS)

    Iqbal, M. Javed; Mirza, Nasir M.; Mirza, Sikander M.

    2008-01-01

    During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

  8. 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).

  9. 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)

  10. 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.

  11. Expanding primary care opportunities: simulation for clinical reasoning.

    PubMed

    Phillippi, Julia C; Bull, Amy; Holley, Sharon L

    2013-05-01

    Many nurse practitioner specialties are requiring that basic primary care be included in their curricula. However, some experienced faculty within the specialty lack primary care experience. With a national shortage of nursing faculty, it is more important than ever to maximize available resources without overtaxing faculty workloads. Revision of our primary care practicum allowed nurse-midwifery faculty to lead a primary care clinical conference, using Family Nurse Practitioner (FNP) faculty as primary care experts. We revamped the clinical conference time to simulate clinical visits to guide the students through the clinical reasoning process. Low-fidelity simulation allowed students time to take a systematic approach to patient assessment, planning, and charting. The FNP "experts" were used to critique student chart notes prior to grading. This collaborative approach to the primary care clinical conference was well received by students, faculty, and preceptors and was quick and inexpensive to implement.

  12. 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)

  13. Primary Connections: Simulating the Classroom in Initial Teacher Education

    NASA Astrophysics Data System (ADS)

    Hume, Anne Christine

    2012-06-01

    The challenge of preparing novice primary teachers for teaching in an educational environment, where science education has low status and many teachers have limited science content knowledge and lack the confidence to teach science, is great. This paper reports on an innovation involving a sustained simulation in an undergraduate science education course as a mediational tool to connect two communities of practice—initial teacher education and expert primary science teaching. The course lecturer and student teachers role-played the expert classroom teacher and primary students (Years 7/8) respectively in an attempt to gain insights into teaching and learning through authentic activity that models good practice in primary science teaching and learning. Activity theory was used to help frame and analyse the data. Findings from the first trial indicate that the simulation was very effective in initiating science pedagogical content knowledge (PCK) development of primary student teachers.

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

    SciTech Connect

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

    1982-08-01

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

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

    SciTech Connect

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

    2014-01-01

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

  16. Primary Connections: Simulating the Classroom in Initial Teacher Education

    ERIC Educational Resources Information Center

    Hume, Anne Christine

    2012-01-01

    The challenge of preparing novice primary teachers for teaching in an educational environment, where science education has low status and many teachers have limited science content knowledge and lack the confidence to teach science, is great. This paper reports on an innovation involving a sustained simulation in an undergraduate science education…

  17. 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.

  18. 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.

  19. Simulations of Porcine Eye Exposure to Primary Blast Insult

    PubMed Central

    Watson, Richard; Gray, Walt; Sponsel, William E.; Lund, Brian J.; Glickman, Randolph D.; Groth, Sylvia L.; Reilly, Matthew A.

    2015-01-01

    Purpose A computational model of the porcine eye was developed to simulate primary blast exposure. This model facilitates understanding of blast-induced injury mechanisms. Methods A computational model of the porcine eye was used to simulate the effects of primary blast loading for comparison with experimental findings from shock tube experiments. The eye model was exposed to overpressure-time histories measured during physical experiments. Deformations and mechanical stresses within various ocular tissues were then examined for correlation with pathological findings in the experiments. Results Stresses and strains experienced in the eye during a primary blast event increase as the severity of the blast exposure increases. Peak stresses in the model occurred in locations in which damage was most often observed in the physical experiments. Conclusions Blast injuries to the anterior chamber may be due to inertial displacement of the lens and ciliary body while posterior damage may arise due to contrecoup interactions of the vitreous and retina. Correlation of modeling predictions with physical experiments lends confidence that the model accurately represents the conditions found in the physical experiments. Translational Relevance This computational model offers insights into the mechanisms of ocular injuries arising due to primary blast and may be used to simulate the effects of new protective eyewear designs. PMID:26336633

  20. 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.

  1. 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.

  2. Simulation of primary fuel atomization processes at subcritical pressures.

    SciTech Connect

    Arienti, Marco

    2013-06-01

    This report documents results from an LDRD project for the first-principles simulation of the early stages of spray formation (primary atomization). The first part describes a Cartesian embedded-wall method for the calculation of flow internal to a real injector in a fully coupled primary calculation. The second part describes the extension to an all-velocity formulation by introducing a momentum-conservative semi-Lagrangian advection and by adding a compressible term in the Poissons equation. Accompanying the description of the new algorithms are verification tests for simple two-phase problems in the presence of a solid interface; a validation study for a scaled-up multi-hole Diesel injector; and demonstration calculations for the closing and opening transients of a single-hole injector and for the high-pressure injection of liquid fuel at supersonic velocity.

  3. High Fidelity Simulation of Primary Atomization in Diesel Engine Sprays

    NASA Astrophysics Data System (ADS)

    Ivey, Christopher; Bravo, Luis; Kim, Dokyun

    2014-11-01

    A high-fidelity numerical simulation of jet breakup and spray formation from a complex diesel fuel injector at ambient conditions has been performed. A full understanding of the primary atomization process in fuel injection of diesel has not been achieved for several reasons including the difficulties accessing the optically dense region. Due to the recent advances in numerical methods and computing resources, high fidelity simulations of atomizing flows are becoming available to provide new insights of the process. In the present study, an unstructured un-split Volume-of-Fluid (VoF) method coupled to a stochastic Lagrangian spray model is employed to simulate the atomization process. A common rail fuel injector is simulated by using a nozzle geometry available through the Engine Combustion Network. The working conditions correspond to a single orifice (90 μm) JP-8 fueled injector operating at an injection pressure of 90 bar, ambient condition at 29 bar, 300 K filled with 100% nitrogen with Rel = 16,071, Wel = 75,334 setting the spray in the full atomization mode. The experimental dataset from Army Research Lab is used for validation in terms of spray global parameters and local droplet distributions. The quantitative comparison will be presented and discussed. Supported by Oak Ridge Associated Universities and the Army Research Laboratory.

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

    SciTech Connect

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

    2006-07-01

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

  5. Delayed auditory feedback simulates features of nonfluent primary progressive aphasia

    PubMed Central

    Maruta, Carolina; Makhmood, Sonya; Downey, Laura E.; Golden, Hannah L.; Fletcher, Phillip D.; Witoonpanich, Pirada; Rohrer, Jonathan D.; Warren, Jason D.

    2014-01-01

    The pathophysiology of nonfluent primary progressive aphasia (nfvPPA) remains poorly understood. Here, we compared quantitatively speech parameters in patients with nfvPPA versus healthy older individuals under altered auditory feedback, which has been shown to modulate normal speech output. Patients (n = 15) and healthy volunteers (n = 17) were recorded while reading aloud under delayed auditory feedback [DAF] with latency 0, 50 or 200 ms and under DAF at 200 ms plus 0.5 octave upward pitch shift. DAF in healthy older individuals was associated with reduced speech rate and emergence of speech sound errors, particularly at latency 200 ms. Up to a third of the healthy older group under DAF showed speech slowing and frequency of speech sound errors within the range of the nfvPPA cohort. Our findings suggest that (in addition to any anterior, primary language output disorder) these key features of nfvPPA may reflect distorted speech input signal processing, as simulated by DAF. DAF may constitute a novel candidate pathophysiological model of posterior dorsal cortical language pathway dysfunction in nfvPPA. PMID:25305712

  6. 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.

  7. MARS simulations of the NuMI primary beamline

    SciTech Connect

    Sergei I Striganov

    2004-05-18

    MARS is a Monte Carlo code for simulation of three-dimensional hadronic and electromagnetic cascades, muon and low-energy neutron transport in shielding and in accelerator and detector components in the energy range from a fraction of an eV up to 100 TeV. This report uses MARS to both transport the 120 GeV primary proton beam from the NuMI extraction Lambertsons through the NuMI Pre-target Hall and calculate the radiological effect of beam losses at various locations and for a variety of conditions. These results are used to: anticipate where beam losses will be significant; determine the level of activation of components; and calculate ground water activation and confirm adequacy of shielding. The results are presented in tables and figures along with drawings of the magnets as they were modeled in MARS. Details of the model elements are found in Appendix A. Further details of beam loss case studies are included in Appendix B.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600

    NASA Astrophysics Data System (ADS)

    Panter, J.; Viguier, B.; Cloué, J.-M.; Foucault, M.; Combrade, P.; Andrieu, E.

    2006-01-01

    In the present study alloy 600 was tested in simulated pressurised water reactor (PWR) primary water, at 360 °C, under an hydrogen partial pressure of 30 kPa. These testing conditions correspond to the maximum sensitivity of alloy 600 to crack initiation. The resulting oxidised structures (corrosion scale and underlying metal) were characterised. A chromium rich oxide layer was revealed, the underlying metal being chromium depleted. In addition, analysis of the chemical composition of the metal close to the oxide scale had allowed to detect oxygen under the oxide scale and particularly in a triple grain boundary. Implication of such a finding on the crack initiation of alloy 600 is discussed. Significant diminution of the crack initiation time was observed for sample oxidised before stress corrosion tests. In view of these results, a mechanism for stress corrosion crack initiation of alloy 600 in PWR primary water was proposed.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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)

  19. 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.

  20. Application of statistical method for determination of primary mass composition of cosmic rays using simulated data

    NASA Astrophysics Data System (ADS)

    Kalita, D.; Boruah, K.

    2013-03-01

    In this paper we have studied the reconstruction of primary mass composition based on simulated longitudinal shower development using a statistical method viz. the multiparametric topological analysis (MTA) and show its applicability for the determination of the primary mass composition. In particular, the sensitivity of X max distribution is tested for simulated data using CORSIKA-6990 code assuming a number of uniform and non-uniform mixed compositions of proton(p), oxygen(O) and iron(Fe) nuclei at primary energies 1017 eV and 1018 eV.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    SciTech Connect

    Miro, R.; Maggini, F.; Barrachina, T.; Verdu, G.; Gomez, A.; Ortego, A.; Murillo, J. C.

    2006-07-01

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

  6. 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.

  7. 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)

  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. 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.

  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. 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)

  12. Automatic determination of primary electron beam parameters in Monte Carlo simulation

    SciTech Connect

    Pena, Javier; Gonzalez-Castano, Diego M.; Gomez, Faustino; Sanchez-Doblado, Francisco; Hartmann, Guenther H.

    2007-03-15

    In order to obtain realistic and reliable Monte Carlo simulations of medical linac photon beams, an accurate determination of the parameters that define the primary electron beam that hits the target is a fundamental step. In this work we propose a new methodology to commission photon beams in Monte Carlo simulations that ensures the reproducibility of a wide range of clinically useful fields. For such purpose accelerated Monte Carlo simulations of 2x2, 10x10, and 20x20 cm{sup 2} fields at SSD=100 cm are carried out for several combinations of the primary electron beam mean energy and radial FWHM. Then, by performing a simultaneous comparison with the correspondent measurements for these same fields, the best combination is selected. This methodology has been employed to determine the characteristics of the primary electron beams that best reproduce a Siemens PRIMUS and a Varian 2100 CD machine in the Monte Carlo simulations. Excellent agreements were obtained between simulations and measurements for a wide range of field sizes. Because precalculated profiles are stored in databases, the whole commissioning process can be fully automated, avoiding manual fine-tunings. These databases can also be used to characterize any accelerators of the same model from different sites.

  13. 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.

  14. 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)

  15. 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.

  16. 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

  17. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect

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

    2000-08-21

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

  18. Three-Dimension Visualization for Primary Wheat Diseases Based on Simulation Model

    NASA Astrophysics Data System (ADS)

    Shijuan, Li; Yeping, Zhu

    Crop simulation model has been becoming the core of agricultural production management and resource optimization management. Displaying crop growth process makes user observe the crop growth and development intuitionisticly. On the basis of understanding and grasping the occurrence condition, popularity season, key impact factors for main wheat diseases of stripe rust, leaf rust, stem rust, head blight and powdery mildew from research material and literature, we designed 3D visualization model for wheat growth and diseases occurrence. The model system will help farmer, technician and decision-maker to use crop growth simulation model better and provide decision-making support. Now 3D visualization model for wheat growth on the basis of simulation model has been developed, and the visualization model for primary wheat diseases is in the process of development.

  19. 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)

  20. Measurement of Primary Ejecta From Normal Incident Hypervelocity Impact on Lunar Regolith Simulant

    NASA Technical Reports Server (NTRS)

    Edwards, David L.; Cooke, William; Moser, Danielle; Swift, Wesley

    2007-01-01

    The National Aeronautics and Space Administration (NASA) continues to make progress toward long-term lunar habitation. Critical to the design of a lunar habitat is an understanding of the lunar surface environment. A subject for further definition is the lunar primary ejecta environment. The document NASA SP-8013 was developed for the Apollo program and is the latest definition of the primary ejecta environment. There is concern that NASA SP-8013 may over-estimate the lunar primary ejecta environment. NASA's Meteoroid Environment Office (MEO) has initiated several tasks to improve the accuracy of our understanding of the lunar surface primary ejecta environment. This paper reports the results of experiments on projectile impact into pumice targets, simulating lunar regolith. The Ames Vertical Gun Range (AVGR) was used to accelerate spherical Pyrex projectiles of 0.29g to velocities ranging between 2.5 km/s and 5.18 km/s. Impact on the pumice target occurred at normal incidence. The ejected particles were detected by thin aluminum foil targets placed around the pumice target in a 0.5 Torr vacuum. A simplistic technique to characterize the ejected particles was formulated. Improvements to this technique will be discussed for implementation in future tests.

  1. 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.

  2. 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.

  3. Structural uncertainty in model-simulated trends of global gross primary production

    NASA Astrophysics Data System (ADS)

    Hashimoto, H.; Wang, W.; Milesi, C.; Xiong, J.; Ganguly, S.; Zhu, Z.; Nemani, R. R.

    2012-12-01

    Accurate representation of the effect of drought on terrestrial vegetation functioning is important for understanding the interannual variability in global Gross Primary Production(GPP) and for projecting carbon sequestration potential by vegetation. Drought effect is usually modeled as a function of Vapor Pressure Deficit (VPD) and/or soil moisture. Global warming is likely to accelerate increasing trend in VPD, while a relatively stable precipitation is predicted. This difference in projections between VPD and precipitation can cause serious discrepancies in vegetation behavior depending on how the ecosystem models represent the drought effect. In this study, we scrutinized the model responses to drought using the 30-year record of GIMMS 3G dataset (1982-2010). A diagnostic ecosystem model, Terrestrial Observation and Prediction System (TOPS), was used to estimate global GPP from 1982 to 2009 with 9 different experimental simulations. The control run of global GPP increased until around 2000, but stayed flat after 2000. Among the simulations with single climate constraint, only the VPD-driven simulation showed a decrease in 2000s, while the other scenarios simulated an increase in GPP. These different responses in 2000s can be attributed to the difference in the representation of water stress in models, i.e. using VPD and/or precipitation. When we compared the trend of simulated GPP with CO2 growth rate, VPD-driven model had the highest correlation with CO2 growth rate. However, spatial map of trend in simulated GPP using GIMMS 3G data showed more consistent with the GPP driven by soil moisture than the GPP driven by VPD. Thus, the GPP driven by soil moisture is close to satellite observations in TOPS model, and high correlation of VPD-driven simulation with CO2 growth rate can be attributed to spurious correlation that is likely induced by the previously reported high correlations between CO2 growth rate and temperature variability.

  4. Estimation of primary pH measurement uncertainty using Monte Carlo simulation

    NASA Astrophysics Data System (ADS)

    Damasceno, J. C.; Borges, R. M. H.; Couto, P. R. G.; Ordine, A. P.; Getrouw, M. A.; Borges, P. P.; Fraga, I. C. S.

    2006-06-01

    pH is a widely used control parameter for several industrial processes. Thus, its correct determination and uncertainty estimation are extremely important. The Guide to the Expression of Uncertainty in Measurement (ISO-GUM) has been extensively used for pH uncertainty estimation. This work uses Monte Carlo simulation to estimate pH uncertainty in a primary pH system for the measurements of a regional comparison (SIM 8.11P-1) in which INMETRO has participated. The results are compared with the ISO-GUM analytical estimation approach and good agreement was found.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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)

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

    SciTech Connect

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

    1981-01-01

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

  11. High-Fidelity Simulation of Primary Blast: Direct Effects on the Head.

    PubMed

    Sawyer, Thomas W; Wang, Yushan; Ritzel, David V; Josey, Tyson; Villanueva, Mercy; Shei, Yimin; Nelson, Peggy; Hennes, Grant; Weiss, Tracy; Vair, Cory; Fan, Changyang; Barnes, Julia

    2016-07-01

    The role of primary blast in blast-induced traumatic brain injury (bTBI) is controversial in part due to the technical difficulties of generating free-field blast conditions in the laboratory. The use of traditional shock tubes often results in artifacts, particularly of dynamic pressure, whereas the forces affecting the head are dependent on where the animal is placed relative to the tube, whether the exposure is whole-body or head-only, and on how the head is actually exposed to the insult (restrained or not). An advanced blast simulator (ABS) has been developed that enables high-fidelity simulation of free-field blastwaves, including sharply defined static and dynamic overpressure rise times, underpressures, and secondary shockwaves. Rats were exposed in head-only fashion to single-pulse blastwaves of 15 to 30 psi static overpressure. Head restraints were configured so as to eliminate concussive and minimize whiplash forces exerted on the head, as shown by kinematic analysis. No overt signs of trauma were present in the animals post-exposure. However, significant changes in brain 2',3'-cyclic nucleotide 3'-phosphohydrolase (CNPase) and neurofilament heavy chain levels were evident by 7 days. In contrast to most studies of primary blast-induced TBI (PbTBI), no elevation of glial fibrillary acidic protein (GFAP) levels was noted when head movement was minimized. The ABS described in this article enables the generation of shockwaves highly representative of free-field blast. The use of this technology, in concert with head-only exposure, minimized head movement, and the kinematic analysis of the forces exerted on the head provide convincing evidence that primary blast directly causes changes in brain function and that GFAP may not be an appropriate biomarker of PbTBI.

  12. Simulation of tree-ring widths with a model for primary production, carbon allocation, and growth

    NASA Astrophysics Data System (ADS)

    Li, G.; Harrison, S. P.; Prentice, I. C.; Falster, D.

    2014-12-01

    We present a simple, generic model of annual tree growth, called "T". This model accepts input from a first-principles light-use efficiency model (the "P" model). The P model provides values for gross primary production (GPP) per unit of absorbed photosynthetically active radiation (PAR). Absorbed PAR is estimated from the current leaf area. GPP is allocated to foliage, transport tissue, and fine-root production and respiration in such a way as to satisfy well-understood dimensional and functional relationships. Our approach thereby integrates two modelling approaches separately developed in the global carbon-cycle and forest-science literature. The T model can represent both ontogenetic effects (the impact of ageing) and the effects of environmental variations and trends (climate and CO2) on growth. Driven by local climate records, the model was applied to simulate ring widths during the period 1958-2006 for multiple trees of Pinus koraiensis from the Changbai Mountains in northeastern China. Each tree was initialised at its actual diameter at the time when local climate records started. The model produces realistic simulations of the interannual variability in ring width for different age cohorts (young, mature, and old). Both the simulations and observations show a significant positive response of tree-ring width to growing-season total photosynthetically active radiation (PAR0) and the ratio of actual to potential evapotranspiration (α), and a significant negative response to mean annual temperature (MAT). The slopes of the simulated and observed relationships with PAR0 and α are similar; the negative response to MAT is underestimated by the model. Comparison of simulations with fixed and changing atmospheric CO2 concentration shows that CO2 fertilisation over the past 50 years is too small to be distinguished in the ring-width data, given ontogenetic trends and interannual variability in climate.

  13. Simulation of tree ring-widths with a model for primary production, carbon allocation and growth

    NASA Astrophysics Data System (ADS)

    Li, G.; Harrison, S. P.; Prentice, I. C.; Falster, D.

    2014-07-01

    We present a simple, generic model of annual tree growth, called "T". This model accepts input from a first-principles light-use efficiency model (the P model). The P model provides values for Gross Primary Production (GPP) per unit of absorbed photosynthetically active radiation (PAR). Absorbed PAR is estimated from the current leaf area. GPP is allocated to foliage, transport-tissue, and fine root production and respiration, in such a way as to satisfy well-understood dimensional and functional relationships. Our approach thereby integrates two modelling approaches separately developed in the global carbon-cycle and forest-science literature. The T model can represent both ontogenetic effects (impact of ageing) and the effects of environmental variations and trends (climate and CO2) on growth. Driven by local climate records, the model was applied to simulate ring widths during 1958-2006 for multiple trees of Pinus koraiensis from the Changbai Mountain, northeastern China. Each tree was initialised at its actual diameter at the time when local climate records started. The model produces realistic simulations of the interannual variability in ring width for different age cohorts (young, mature, old). Both the simulations and observations show a significant positive response of tree-ring width to growing-season total photosynthetically active radiation (PAR0) and the ratio of actual to potential evapotranspiration (α), and a significant negative response to mean annual temperature (MAT). The slopes of the simulated and observed relationships with PAR0 and α are similar; the negative response to MAT is underestimated by the model. Comparison of simulations with fixed and changing atmospheric CO2 concentration shows that CO2 fertilization over the past 50 years is too small to be distinguished in the ring-width data given ontogenetic trends and interannual variability in climate.

  14. Measurements and simulation of forest leaf area index and net primary productivity in Northern China.

    PubMed

    Wang, P; Sun, R; Hu, J; Zhu, Q; Zhou, Y; Li, L; Chen, J M

    2007-11-01

    Large scale process-based modeling is a useful approach to estimate distributions of global net primary productivity (NPP). In this paper, in order to validate an existing NPP model with observed data at site level, field experiments were conducted at three sites in northern China. One site is located in Qilian Mountain in Gansu Province, and the other two sites are in Changbaishan Natural Reserve and Dunhua County in Jilin Province. Detailed field experiments are discussed and field data are used to validate the simulated NPP. Remotely sensed images including Landsat Enhanced Thematic Mapper plus (ETM+, 30 m spatial resolution in visible and near infrared bands) and Advanced Spaceborne Thermal Emission and Reflection Radiometer (ASTER, 15m spatial resolution in visible and near infrared bands) are used to derive maps of land cover, leaf area index, and biomass. Based on these maps, field measured data, soil texture and daily meteorological data, NPP of these sites are simulated for year 2001 with the boreal ecosystem productivity simulator (BEPS). The NPP in these sites ranges from 80 to 800 gCm(-2)a(-1). The observed NPP agrees well with the modeled NPP. This study suggests that BEPS can be used to estimate NPP in northern China if remotely sensed images of high spatial resolution are available.

  15. [Simulating net primary production of rice and wheat crops: model validation and scenario prediction].

    PubMed

    Yang, Zhao-fang; Yu, Yong-qiang; Huang, Yao

    2005-03-01

    A model developed by the authors was validated against a total of 98 independent data sets to simulate net primary production (NPP) of rice and wheat crops. These data sets come from literature review and include field measurements conducted in different regions of China with various rates of N application. Model validation indicates that NPP of rice and wheat crops in main cultivated-area of China can be well simulated from weather, soil and N fertilization. A comparison between the simulated (y) and the observed NPP (x) resulted in a regression of y = 1.05x- 16.8 (r2= 0.771,p < 0.001, n = 98). Model scenario prediction for Nanjing area suggests that the increase of atmospheric CO2 concentration will enhance carbon fixation, while the increase of air temperature will reduce carbon fixation by rice and wheat crops. Effect of global warming on the wheat carbon fixation is less than on the rice. Under present and future scenario with atmospheric CO2 concentration of 540 micromol x mol(-1) and temperature increment of 1-4 degrees C, N fertilization will enhance carbon fixation of rice and wheat crops. The enhancement for wheat is more significant than that for rice crop. However, the application of N will not significantly improve the carbon fixation, even reduce rice NPP when the N application rate is higher than 150 kg x hm(-2).

  16. Numerical Methodology for Coupled Time-Accurate Simulations of Primary and Secondary Flowpaths in Gas Turbines

    NASA Technical Reports Server (NTRS)

    Przekwas, A. J.; Athavale, M. M.; Hendricks, R. C.; Steinetz, B. M.

    2006-01-01

    Detailed information of the flow-fields in the secondary flowpaths and their interaction with the primary flows in gas turbine engines is necessary for successful designs with optimized secondary flow streams. Present work is focused on the development of a simulation methodology for coupled time-accurate solutions of the two flowpaths. The secondary flowstream is treated using SCISEAL, an unstructured adaptive Cartesian grid code developed for secondary flows and seals, while the mainpath flow is solved using TURBO, a density based code with capability of resolving rotor-stator interaction in multi-stage machines. An interface is being tested that links the two codes at the rim seal to allow data exchange between the two codes for parallel, coupled execution. A description of the coupling methodology and the current status of the interface development is presented. Representative steady-state solutions of the secondary flow in the UTRC HP Rig disc cavity are also presented.

  17. Observation and simulation of net primary productivity in Qilian Mountain, western China.

    PubMed

    Zhou, Y; Zhu, Q; Chen, J M; Wang, Y Q; Liu, J; Sun, R; Tang, S

    2007-11-01

    We modeled net primary productivity (NPP) at high spatial resolution using an advanced spaceborne thermal emission and reflection radiometer (ASTER) image of a Qilian Mountain study area using the boreal ecosystem productivity simulator (BEPS). Two key driving variables of the model, leaf area index (LAI) and land cover type, were derived from ASTER and moderate resolution imaging spectroradiometer (MODIS) data. Other spatially explicit inputs included daily meteorological data (radiation, precipitation, temperature, humidity), available soil water holding capacity (AWC), and forest biomass. NPP was estimated for coniferous forests and other land cover types in the study area. The result showed that NPP of coniferous forests in the study area was about 4.4 tCha(-1)y(-1). The correlation coefficient between the modeled NPP and ground measurements was 0.84, with a mean relative error of about 13.9%.

  18. Modeling and Simulation of the ITER First Wall/Blanket Primary Heat Transfer System

    SciTech Connect

    Ying, Alice; Popov, Emilian L

    2011-01-01

    ITER inductive power operation is modeled and simulated using a thermal-hydraulics system code (RELAP5) integrated with a 3-D CFD (SC-Tetra) code. The Primary Heat Transfer System (PHTS) functions are predicted together with the main parameters operational ranges. The control algorithm strategy and derivation are summarized as well. The First Wall and Blanket modules are the primary components of PHTS, used to remove the major part of the thermal heat from the plasma. The modules represent a set of flow channels in solid metal structure that serve to absorb the radiation heat and nuclear heating from the fusion reactions and to provide shield for the vacuum vessel. The blanket modules are water cooled. The cooling is forced convective with constant blanket inlet temperature and mass flow rate. Three independent water loops supply coolant to the three blanket sectors. The main equipment of each loop consists of a pump, a steam pressurizer and a heat exchanger. A major feature of ITER is the pulsed operation. The plasma does not burn continuously, but on intervals with large periods of no power between them. This specific feature causes design challenges to accommodate the thermal expansion of the coolant during the pulse period and requires active temperature control to maintain a constant blanket inlet temperature.

  19. Large-eddy simulation of cavitating nozzle flow and primary jet break-up

    NASA Astrophysics Data System (ADS)

    Ã-rley, F.; Trummler, T.; Hickel, S.; Mihatsch, M. S.; Schmidt, S. J.; Adams, N. A.

    2015-08-01

    We employ a barotropic two-phase/two-fluid model to study the primary break-up of cavitating liquid jets emanating from a rectangular nozzle, which resembles a high aspect-ratio slot flow. All components (i.e., gas, liquid, and vapor) are represented by a homogeneous mixture approach. The cavitating fluid model is based on a thermodynamic-equilibrium assumption. Compressibility of all phases enables full resolution of collapse-induced pressure wave dynamics. The thermodynamic model is embedded into an implicit large-eddy simulation (LES) environment. The considered configuration follows the general setup of a reference experiment and is a generic reproduction of a scaled-up fuel injector or control valve as found in an automotive engine. Due to the experimental conditions, it operates, however, at significantly lower pressures. LES results are compared to the experimental reference for validation. Three different operating points are studied, which differ in terms of the development of cavitation regions and the jet break-up characteristics. Observed differences between experimental and numerical data in some of the investigated cases can be caused by uncertainties in meeting nominal parameters by the experiment. The investigation reveals that three main mechanisms promote primary jet break-up: collapse-induced turbulent fluctuations near the outlet, entrainment of free gas into the nozzle, and collapse events inside the jet near the liquid-gas interface.

  20. Large-eddy simulation of cavitating nozzle flow and primary jet break-up

    SciTech Connect

    Örley, F. Trummler, T.; Mihatsch, M. S.; Schmidt, S. J.; Adams, N. A.; Hickel, S.

    2015-08-15

    We employ a barotropic two-phase/two-fluid model to study the primary break-up of cavitating liquid jets emanating from a rectangular nozzle, which resembles a high aspect-ratio slot flow. All components (i.e., gas, liquid, and vapor) are represented by a homogeneous mixture approach. The cavitating fluid model is based on a thermodynamic-equilibrium assumption. Compressibility of all phases enables full resolution of collapse-induced pressure wave dynamics. The thermodynamic model is embedded into an implicit large-eddy simulation (LES) environment. The considered configuration follows the general setup of a reference experiment and is a generic reproduction of a scaled-up fuel injector or control valve as found in an automotive engine. Due to the experimental conditions, it operates, however, at significantly lower pressures. LES results are compared to the experimental reference for validation. Three different operating points are studied, which differ in terms of the development of cavitation regions and the jet break-up characteristics. Observed differences between experimental and numerical data in some of the investigated cases can be caused by uncertainties in meeting nominal parameters by the experiment. The investigation reveals that three main mechanisms promote primary jet break-up: collapse-induced turbulent fluctuations near the outlet, entrainment of free gas into the nozzle, and collapse events inside the jet near the liquid-gas interface.

  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. 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

  3. Effect of lithium hydroxide on stability of fuel cladding oxide film in simulated pressurized water reactor primary water environments

    SciTech Connect

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

    1997-09-01

    The trend in pressurized water reactors (PWR) toward higher burnups, increasing lithium concentrations, and higher coolant temperatures imposes a demand for better fuel cladding corrosion and hydriding properties. There is a lack of reliable and fast in-situ techniques to investigate zirconium alloys in high-temperature water environments. The contact electric resistance (CER) technique was used to measure the electric resistance of the oxide growing on a zirconium-based fuel cladding material. Lithium hydroxide (LiOH) decreased electric resistance of the oxide when LiOH was in excess of {approximately} 70 ppm in PWR water at 300 C. Electric resistance of the oxide was dependent upon LiOH concentration and was shown to correlate inversely with the effect of LiOH on weight gain. Kinetics of the decrease of electric resistance indicated the mechanism of degradation was a phase transformation rather than a diffusion-limited process.

  4. 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

  5. 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.

  6. Simulation of the effects of bottom topography on net primary production induced by riverine input

    NASA Astrophysics Data System (ADS)

    Hoshiba, Yasuhiro; Yamanaka, Yasuhiro

    2016-04-01

    Riverine input often leads to high biological productivity in coastal areas. In coastal areas termed as region of freshwater influence (ROFI), horizontal anticyclonic gyres and vertical circulation form by density differences between buoyant river water and sea water. Previous physical oceanography studies have shown that the horizontal pattern of anticyclonic gyres and the strength of vertical circulation are dependent on the bottom topography of ROFI. However, the dependencies of biogeochemical cycles such as the net primary production (NPP) on the bottom topography have not been verified. In order to clarify how the bottom topography affects the NPP in phytoplankton blooms caused by riverine input through the physical processes in ROFI, we used an ocean general circulation model (OGCM) including a simple ecosystem model and conducted several case studies varying the bottom slope angle in the ideal settings. We estimated NPP categorized into three nutrients supplied from the river, the sea-subsurface layer and via regeneration: RI-NPP, S-NPP and RE-NPP. S-NPP and RE-NPP are larger and smaller with a steeper slope, respectively, while RI-NPP is not affected by the slope angle. As a result, total NPP is weakly dependent on the slope angle, i.e., because S- and RE-NPPs cancel each other out through two physical processes, (1) S-NPP is controlled by the strength of the vertical circulation and (2) RE-NPP is controlled by the shape of the horizontal gyre, which both vary with the bottom slope angle. We also conducted realistic simulations for Ishikari Bay, Japan and confirmed a similar dependency to that in the above ideal settings. That is, the simulation results are consistent with the regime of ideal settings and show that RI- and RE-NPPs are important variables for Ishikari Bay which has a gentle slope.

  7. Effects of topography on simulated net primary productivity at landscape scale.

    PubMed

    Chen, X F; Chen, J M; An, S Q; Ju, W M

    2007-11-01

    Local topography significantly affects spatial variations of climatic variables and soil water movement in complex terrain. Therefore, the distribution and productivity of ecosystems are closely linked to topography. Using a coupled terrestrial carbon and hydrological model (BEPS-TerrainLab model), the topographic effects on the net primary productivity (NPP) are analyzed through four modelling experiments for a 5700 km(2) area in Baohe River basin, Shaanxi Province, northwest of China. The model was able to capture 81% of the variability in NPP estimated from tree rings, with a mean relative error of 3.1%. The average NPP in 2003 for the study area was 741 gCm(-2)yr(-1) from a model run including topographic effects on the distributions of climate variables and lateral flow of ground water. Topography has considerable effect on NPP, which peaks near 1350 m above the sea level. An elevation increase of 100 m above this level reduces the average annual NPP by about 25 gCm(-2). The terrain aspect gives rise to a NPP change of 5% for forests located below 1900 m as a result of its influence on incident solar radiation. For the whole study area, a simulation totally excluding topographic effects on the distributions of climatic variables and ground water movement overestimated the average NPP by 5%.

  8. Primary and secondary restraints of human and ovine knees for simulated in vivo gait kinematics.

    PubMed

    Nesbitt, Rebecca J; Herfat, Safa T; Boguszewski, Daniel V; Engel, Andrew J; Galloway, Marc T; Shearn, Jason T

    2014-06-27

    Knee soft tissue structures are frequently injured, leading to the development of osteoarthritis even with treatment. Understanding how these structures contribute to knee function during activities of daily living (ADLs) is crucial in creating more effective treatments. This study was designed to determine the role of different knee structures during a simulated ADL in both human knees and ovine stifle joints. A six degree-of-freedom robot was used to reproduce each species' in vivo gait while measuring three-dimensional joint forces and torques. Using a semi-randomized selective cutting method, we determined the primary and secondary structures contributing to the forces and torques along and about each anatomical axis. In both species, the bony interaction, ACL, and medial meniscus provided most of the force contributions during stance, whereas the ovine MCL, human bone, and ACLs of both species were the key contributors during swing. This study contributes to our overarching goal of establishing functional tissue engineering parameters for knee structures by further validating biomechanical similarities between the ovine model and the human to provide a platform for measuring biomechanics during an in vivo ADL. These parameters will be used to develop more effective treatments for knee injuries to reduce or eliminate the incidence of osteoarthritis.

  9. 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.

  10. 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.

  11. Sensitivity of Crop Gross Primary Production Simulations to In-situ and Reanalysis Meteorological Data

    NASA Astrophysics Data System (ADS)

    Jin, C.; Xiao, X.; Wagle, P.

    2014-12-01

    Accurate estimation of crop Gross Primary Production (GPP) is important for food securityand terrestrial carbon cycle. Numerous publications have reported the potential of the satellite-based Production Efficiency Models (PEMs) to estimate GPP driven by in-situ climate data. Simulations of the PEMs often require surface reanalysis climate data as inputs, for example, the North America Regional Reanalysis datasets (NARR). These reanalysis datasets showed certain biases from the in-situ climate datasets. Thus, sensitivity analysis of the PEMs to the climate inputs is needed before their application at the regional scale. This study used the satellite-based Vegetation Photosynthesis Model (VPM), which is driven by solar radiation (R), air temperature (T), and the satellite-based vegetation indices, to quantify the causes and degree of uncertainties in crop GPP estimates due to different meteorological inputs at the 8-day interval (in-situ AmeriFlux data and NARR surface reanalysis data). The NARR radiation (RNARR) explained over 95% of the variability in in-situ RAF and TAF measured from AmeriFlux. The bais of TNARR was relatively small. However, RNARR had a systematical positive bias of ~3.5 MJ m-2day-1 from RAF. A simple adjustment based on the spatial statistic between RNARR and RAF produced relatively accurate radiation data for all crop site-years by reducing RMSE from 4 to 1.7 MJ m-2day-1. The VPM-based GPP estimates with three climate datasets (i.e., in-situ, and NARR before and after adjustment, GPPVPM,AF, GPPVPM,NARR, and GPPVPM,adjNARR) showed good agreements with the seasonal dynamics of crop GPP derived from the flux towers (GPPAF). The GPPVPM,AF differed from GPPAF by 2% for maize, and -8% to -12% for soybean on the 8-day interval. The positive bias of RNARR resulted in an overestimation of GPPVPM,NARR at both maize and soybean systems. However, GPPVPM,adjNARR significantly reduced the uncertainties of the maize GPP from 25% to 2%. The results from this

  12. Leaf chlorophyll constraint on model simulated gross primary productivity in agricultural systems

    NASA Astrophysics Data System (ADS)

    Houborg, Rasmus; McCabe, Matthew F.; Cescatti, Alessandro; Gitelson, Anatoly A.

    2015-12-01

    Leaf chlorophyll content (Chll) may serve as an observational proxy for the maximum rate of carboxylation (Vmax), which describes leaf photosynthetic capacity and represents the single most important control on modeled leaf photosynthesis within most Terrestrial Biosphere Models (TBMs). The parameterization of Vmax is associated with great uncertainty as it can vary significantly between plants and in response to changes in leaf nitrogen (N) availability, plant phenology and environmental conditions. Houborg et al. (2013) outlined a semi-mechanistic relationship between Vmax25 (Vmax normalized to 25 °C) and Chll based on inter-linkages between Vmax25, Rubisco enzyme kinetics, N and Chll. Here, these relationships are parameterized for a wider range of important agricultural crops and embedded within the leaf photosynthesis-conductance scheme of the Community Land Model (CLM), bypassing the questionable use of temporally invariant and broadly defined plant functional type (PFT) specific Vmax25 values. In this study, the new Chll constrained version of CLM is refined with an updated parameterization scheme for specific application to soybean and maize. The benefit of using in-situ measured and satellite retrieved Chll for constraining model simulations of Gross Primary Productivity (GPP) is evaluated over fields in central Nebraska, U.S.A between 2001 and 2005. Landsat-based Chll time-series records derived from the Regularized Canopy Reflectance model (REGFLEC) are used as forcing to the CLM. Validation of simulated GPP against 15 site-years of flux tower observations demonstrate the utility of Chll as a model constraint, with the coefficient of efficiency increasing from 0.91 to 0.94 and from 0.87 to 0.91 for maize and soybean, respectively. Model performances particularly improve during the late reproductive and senescence stage, where the largest temporal variations in Chll (averaging 35-55 μg cm-2 for maize and 20-35 μg cm-2 for soybean) are observed. While

  13. 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

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

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

    NASA Astrophysics Data System (ADS)

    Barbet, M.; Guillery, M.

    1981-06-01

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

  2. 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.

  3. 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.

  4. 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.

  5. Sensitivity of Global Terrestrial Gross Primary Production to Hydrologic States Simulated by the Community Land Model Using Two Runoff Parameterizations

    SciTech Connect

    Lei, Huimin; Huang, Maoyi; Leung, Lai-Yung R.; Yang, Dawen; Shi, Xiaoying; Mao, Jiafu; Hayes, Daniel J.; Schwalm, C.; Wei, Yaxing; Liu, Shishi

    2014-09-01

    The terrestrial water and carbon cycles interact strongly at various spatio-temporal scales. To elucidate how hydrologic processes may influence carbon cycle processes, differences in terrestrial carbon cycle simulations induced by structural differences in two runoff generation schemes were investigated using the Community Land Model 4 (CLM4). Simulations were performed with runoff generation using the default TOPMODEL-based and the Variable Infiltration Capacity (VIC) model approaches under the same experimental protocol. The comparisons showed that differences in the simulated gross primary production (GPP) are mainly attributed to differences in the simulated leaf area index (LAI) rather than soil moisture availability. More specifically, differences in runoff simulations can influence LAI through changes in soil moisture, soil temperature, and their seasonality that affect the onset of the growing season and the subsequent dynamic feedbacks between terrestrial water, energy, and carbon cycles. As a result of a relative difference of 36% in global mean total runoff between the two models and subsequent changes in soil moisture, soil temperature, and LAI, the simulated global mean GPP differs by 20.4%. However, the relative difference in the global mean net ecosystem exchange between the two models is small (2.1%) due to competing effects on total mean ecosystem respiration and other fluxes, although large regional differences can still be found. Our study highlights the significant interactions among the water, energy, and carbon cycles and the need for reducing uncertainty in the hydrologic parameterization of land surface models to better constrain carbon cycle modeling.

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

    SciTech Connect

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

    2015-09-01

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

  7. Assessment of simulated lesions on primary teeth with near-infrared imaging

    PubMed Central

    Tam, Wilson; Lee, Robert C.; Lin, Brent; Simon, Jacob C.; Fried, Daniel

    2016-01-01

    Previous studies have demonstrated that the structural changes on enamel due to demineralization and remineralization can be exploited through optical imaging methods such as QLF, thermal and NIR imaging. The purpose of this study is to investigate whether PS-OCT and NIR reflectance imaging can be utilized to assess lesion structure in artificial enamel lesions on the smooth surfaces of primary teeth exposed to fluoride. The smooth coronal surfaces of primary teeth (n=25) were divided into 4 windows: sound, demineralization, demineralization with remineralization and APF with demineralization. Windows were treated with either acidulated phosphate fluoride (APF) for 1 minute, a demineralization solution for 4 days, and/or an acidic remineralization solution for 12 days. The samples were imaged using PS-OCT, QLF and NIR reflectance at 1400–1700 nm wavelengths. This study demonstrated that both PS-OCT and NIR reflectance imaging were suitable for assessing lesion structure in the smooth surfaces of primary dentition. PMID:26997743

  8. Assessment of simulated lesions on primary teeth with near-infrared imaging

    NASA Astrophysics Data System (ADS)

    Tam, Wilson; Lee, Robert C.; Lin, Brent; Simon, Jacob C.; Fried, Daniel

    2016-02-01

    Previous studies have demonstrated that the structural changes on enamel due to demineralization and remineralization can be exploited through optical imaging methods such as QLF, thermal and NIR imaging. The purpose of this study is to investigate whether PS-OCT and NIR reflectance imaging can be utilized to assess lesion structure in artificial enamel lesions on the smooth surfaces of primary teeth exposed to fluoride. The smooth coronal surfaces of primary teeth (n=25) were divided into 4 windows: sound, demineralization, demineralization with remineralization and APF with demineralization. Windows were treated with either acidulated phosphate fluoride (APF) for 1 minute, a demineralization solution for 4 days, and/or an acidic remineralization solution for 12 days. The samples were imaged using PS-OCT, QLF and NIR reflectance at 1400-1700 nm wavelengths. This study demonstrated that both PS-OCT and NIR reflectance imaging were suitable for assessing lesion structure in the smooth surfaces of primary dentition.

  9. PRELIMINARY COUPLING OF THE MONTE CARLO CODE OPENMC AND THE MULTIPHYSICS OBJECT-ORIENTED SIMULATION ENVIRONMENT (MOOSE) FOR ANALYZING DOPPLER FEEDBACK IN MONTE CARLO SIMULATIONS

    SciTech Connect

    Matthew Ellis; Derek Gaston; Benoit Forget; Kord Smith

    2011-07-01

    In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes. An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.

  10. A station blackout simulation for the Advanced Neutron Source Reactor using the integrated primary and secondary system model

    SciTech Connect

    Schneider, E.A.

    1994-06-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at Oak Ridge National Laboratory. This paper deals with thermal-hydraulic analysis of ANSR`s cooling systems during nominal and transient conditions, with the major effort focusing upon the construction and testing of computer models of the reactor`s primary, secondary and reflector vessel cooling systems. The code RELAP5 was used to simulate transients, such as loss of coolant accidents and loss of off-site power, as well as to model the behavior of the reactor in steady state. Three stages are involved in constructing and using a RELAP5 model: (1) construction and encoding of the desired model, (2) testing and adjustment of the model until a satisfactory steady state is achieved, and (3) running actual transients using the steady-state results obtained earlier as initial conditions. By use of the ANSR design specifications, a model of the reactor`s primary and secondary cooling systems has been constructed to run a transient simulating a loss of off-site power. This incident assumes a pump coastdown in both the primary and secondary loops. The results determine whether the reactor can survive the transition from forced convection to natural circulation.

  11. MD simulations of phase stability of PuGa alloys: Effects of primary radiation defects and helium bubbles

    SciTech Connect

    Dremov, V. V.; Sapozhnikov, F. A.; Ionov, G. V.; Karavaev, A. V.; Vorobyova, M. A.; Chung, B. W.

    2013-05-14

    We present classical molecular dynamics (MD) with Modified Embedded Atom Model (MEAM) simulations to investigate the role of primary radiation defects and radiogenic helium as factors affecting the phase stability of PuGa alloys in cooling–heating cycles at ambient pressure. The models of PuGa alloys equilibrated at ambient conditions were subjected to cooling–heating cycles in which they were initially cooled down to 100 K and then heated up to 500 K at ambient pressure. The rate of temperature change in the cycles was 10 K/ns. The simulations showed that the initial FCC phase of PuGa alloys undergo polymorphous transition in cooling to a lower symmetry α'-phase. All the alloys undergo direct and reverse polymorphous transitions in the cooling–heating cycles. The alloys containing vacancies shift in both transitions to lower temperatures relative to the defect-free alloys. The radiogenic helium has much less effect on the phase stability compared to that of primary radiation defects (in spite of the fact that helium concentration is twice of that for the primary radiation defects). Lastly, this computational result agrees with experimental data on unconventional stabilization mechanism of PuGa alloys.

  12. MD simulations of phase stability of PuGa alloys: Effects of primary radiation defects and helium bubbles

    DOE PAGES

    Dremov, V. V.; Sapozhnikov, F. A.; Ionov, G. V.; ...

    2013-05-14

    We present classical molecular dynamics (MD) with Modified Embedded Atom Model (MEAM) simulations to investigate the role of primary radiation defects and radiogenic helium as factors affecting the phase stability of PuGa alloys in cooling–heating cycles at ambient pressure. The models of PuGa alloys equilibrated at ambient conditions were subjected to cooling–heating cycles in which they were initially cooled down to 100 K and then heated up to 500 K at ambient pressure. The rate of temperature change in the cycles was 10 K/ns. The simulations showed that the initial FCC phase of PuGa alloys undergo polymorphous transition in coolingmore » to a lower symmetry α'-phase. All the alloys undergo direct and reverse polymorphous transitions in the cooling–heating cycles. The alloys containing vacancies shift in both transitions to lower temperatures relative to the defect-free alloys. The radiogenic helium has much less effect on the phase stability compared to that of primary radiation defects (in spite of the fact that helium concentration is twice of that for the primary radiation defects). Lastly, this computational result agrees with experimental data on unconventional stabilization mechanism of PuGa alloys.« less

  13. 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)

  14. 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.

  15. 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.

  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. 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)

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

    SciTech Connect

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

    2006-10-10

    development of a detector cluster and corresponding high-precision driving system to collect radiation signatures inside PWR spent fuel assemblies. The data obtained would provide the spatial distribution of the neutron and gamma flux fields within the spent fuel assembly, while the data analysis would be used to help identify missing or replaced pins. Monte Carlo simulations have been performed to help validate this concept using a realistic 17 x 17 PWR spent fuel assembly [4-5]. The initial results of this study show that neutron profile in the guide tubes, when obtained in the presence of missing pins, can be identifiably different from the profiles obtained without missing pins, Our latest simulations have focused upon a specific type of fission chamber that could be tested for this application.

  19. [Parameter sensitivity of simulating net primary productivity of Larix olgensis forest based on BIOME-BGC model].

    PubMed

    He, Li-hong; Wang, Hai-yan; Lei, Xiang-dong

    2016-02-01

    Model based on vegetation ecophysiological process contains many parameters, and reasonable parameter values will greatly improve simulation ability. Sensitivity analysis, as an important method to screen out the sensitive parameters, can comprehensively analyze how model parameters affect the simulation results. In this paper, we conducted parameter sensitivity analysis of BIOME-BGC model with a case study of simulating net primary productivity (NPP) of Larix olgensis forest in Wangqing, Jilin Province. First, with the contrastive analysis between field measurement data and the simulation results, we tested the BIOME-BGC model' s capability of simulating the NPP of L. olgensis forest. Then, Morris and EFAST sensitivity methods were used to screen the sensitive parameters that had strong influence on NPP. On this basis, we also quantitatively estimated the sensitivity of the screened parameters, and calculated the global, the first-order and the second-order sensitivity indices. The results showed that the BIOME-BGC model could well simulate the NPP of L. olgensis forest in the sample plot. The Morris sensitivity method provided a reliable parameter sensitivity analysis result under the condition of a relatively small sample size. The EFAST sensitivity method could quantitatively measure the impact of simulation result of a single parameter as well as the interaction between the parameters in BIOME-BGC model. The influential sensitive parameters for L. olgensis forest NPP were new stem carbon to new leaf carbon allocation and leaf carbon to nitrogen ratio, the effect of their interaction was significantly greater than the other parameter' teraction effect.

  20. Primary pancreatic sinus histiocytosis with massive lymphadenopathy (Rosai-Dorfman disease): an unusual extranodal manifestation clinically simulating malignancy.

    PubMed

    Podberezin, Mark; Angeles, Ronald; Guzman, Grace; Peace, David; Gaitonde, Sujata

    2010-02-01

    Abstract Sinus histiocytosis with massive lymphadenopathy (SHML), also called Rosai-Dorfman disease, is a rare entity. Its etiology and pathogenesis are still essentially unclear. The histologic hallmark of this disease is proliferation of distinctive histiocytes within lymph node sinuses and in extranodal sites. Approximately 23% of patients with SHML, documented in the SHML Registry, presented with disease primarily in extranodal sites, and very few cases of SHML (<1%) involving the gastrointestinal system have been described in the literature. We report an unusual case of primary pancreatic SHML with infiltration of the process into peripancreatic, perinephric, and perisplenic adipose tissue, simulating malignancy.

  1. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    SciTech Connect

    McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard; Grey, Carissa; Engelhardt, Charles; Saltzstein, Sylvia J.; Sorenson, Ken B.

    2016-05-01

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  2. The Design and Development of a Simulation to Teach Water Conservation to Primary School Students

    ERIC Educational Resources Information Center

    Campbell, Lee

    2004-01-01

    Information and Communications Technology (ICT) plays a dominant role in enhancing teaching and learning. Similar advances have been made in the use of multimedia in the classroom. These advances are coupled with newer developmental tools and techniques. This paper examines the design and development of a simulation on water conservation. Science…

  3. 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.

  4. 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)

  5. 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.

  6. Patient-specific simulations and measurements of the magneto-hemodynamic effect in human primary vessels.

    PubMed

    Kyriakou, Adamos; Neufeld, Esra; Szczerba, Dominik; Kainz, Wolfgang; Luechinger, Roger; Kozerke, Sebastian; McGregor, Robert; Kuster, Niels

    2012-02-01

    This paper investigates the main characteristics of the magneto-hemodynamic (MHD) response for application as a biomarker of vascular blood flow. The induced surface potential changes of a volunteer exposed to a 3 T static B0 field of a magnetic resonance imaging (MRI) magnet were measured over time at multiple locations by an electrocardiogram device and compared to simulation results. The flow simulations were based on boundary conditions derived from MRI flow measurements restricted to the aorta and vena cava. A dedicated and validated low-frequency electromagnetic solver was applied to determine the induced temporal surface potential change from the obtained 4D flow distribution using a detailed whole-body model of the volunteer. The simulated MHD signal agreed with major characteristics of the measured signal (temporal location of main peak, magnitude, variation across chest and along torso) except in the vicinity of the heart. The MHD signal is mostly influenced by the aorta; however, more vessels and better boundary conditions are needed to analyze the finer details of the response. The results show that the MHD signal is strongly position dependent with highly variable but reproducibly measurable distinguished characteristics. Additional investigations are necessary before determining whether the MHD effect is a reliable reference for location-specific information on blood flow.

  7. Mechanisms controlling primary and new production in a global ecosystem model - Part I: Validation of the biological simulation

    NASA Astrophysics Data System (ADS)

    Popova, E. E.; Coward, A. C.; Nurser, G. A.; de Cuevas, B.; Fasham, M. J. R.; Anderson, T. R.

    2006-12-01

    A global general circulation model coupled to a simple six-compartment ecosystem model is used to study the extent to which global variability in primary and export production can be realistically predicted on the basis of advanced parameterizations of upper mixed layer physics, without recourse to introducing extra complexity in model biology. The "K profile parameterization" (KPP) scheme employed, combined with 6-hourly external forcing, is able to capture short-term periodic and episodic events such as diurnal cycling and storm-induced deepening. The model realistically reproduces various features of global ecosystem dynamics that have been problematic in previous global modelling studies, using a single generic parameter set. The realistic simulation of deep convection in the North Atlantic, and lack of it in the North Pacific and Southern Oceans, leads to good predictions of chlorophyll and primary production in these contrasting areas. Realistic levels of primary production are predicted in the oligotrophic gyres due to high frequency external forcing of the upper mixed layer (accompanying paper Popova et al., 2006) and novel parameterizations of zooplankton excretion. Good agreement is shown between model and observations at various JGOFS time series sites: BATS, KERFIX, Papa and HOT. One exception is the northern North Atlantic where lower grazing rates are needed, perhaps related to the dominance of mesozooplankton there. The model is therefore not globally robust in the sense that additional parameterizations are needed to realistically simulate ecosystem dynamics in the North Atlantic. Nevertheless, the work emphasises the need to pay particular attention to the parameterization of mixed layer physics in global ocean ecosystem modelling as a prerequisite to increasing the complexity of ecosystem models.

  8. A conceptual model of primary productivity in shallow streams using biomass simulation. Technical completion report

    SciTech Connect

    Elliott, J.C.; McDonnell, A.J.

    1982-06-01

    A conceptual model for primary productivity was developed for application to rooted aquatic macrophytes in streams to assist studies of eutrophication and control of water quality in supplementing outputs of dissolved oxygen (DO) models of pollution loads. This model included a first-order differential equation of biomass, with specific rates for photosynthesis, respiration, and death. A model component was developed to describe available light spatially/temporally in the weed bed, as reduced from extraterrestrial solar radiation. A DO model component included terms for photosynthetic production, plant respiration, and a benthal sink due to dead plant matter decay. The latter, a first-order exponential oxygen sink, had not been previously included in DO models.

  9. Primary male neuroendocrine adenocarcinoma involving the nipple simulating Merkel cell carcinoma - a diagnostic pitfall.

    PubMed

    Mecca, Patricia; Busam, Klaus

    2008-02-01

    Male breast cancer is a rare entity accounting for < 1% of all breast cancer cases in the United States, but with a rate that has been rising over the last 25 years. Nipple skin/subcutaneous tumors in men are even rarer. Likewise, true neuroendocrine carcinoma of the breast, defined as > 50% of tumor cells staining for either chromogranin or synaptophysin, is not a common entity, usually occurring in older women. We present the case of a 70-year-old man with a slowly growing nipple mass that had enlarged over the previous 1.5 years. The histology consisted of nests, trabeculae and sheets of basaloid cells with rare abortive gland formation and a pushing edge. The case was originally misdiagnosed as a Merkel cell carcinoma, based largely on histologic morphology. Strong staining for synaptophysin (in greater than 50% of cells), CD56, keratins AE1 : AE3 and Cam 5.2, as well as estrogen receptor and progesterone receptor was noted. Myoepithelial cells within in situ areas were identified using stains for calponin and 4A4, supporting a primary mammary duct origin. Additionally, a substantial portion of cells stained for Gross Cystic Disease Fluid Protein-15 (GCDFP-15), confirming some overlap with sweat duct differentiation. To the best of our knowledge, although reported in the male breast, no case of primary nipple neuroendocrine carcinoma in a male patient has been reported in the literature. The gender of the patient and association with the skin of the chest wall probably contributed to the original misdiagnosis of Merkel cell carcinoma in this patient.

  10. 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

  11. Sliding simulation of automotive brake primary contact with variable amounts of copper and graphite nanoparticles

    NASA Astrophysics Data System (ADS)

    Dmitriev, A. I.; Österle, W.

    2016-11-01

    Copper is one of the most important components in brake pads and its amount can reach up to 14%. In spite of a number of positive features copper usage in brake pad formulations has recently become the subject of considerable discussions, primarily due to concerns about potential risks related to environmental impacts of copper particles. So, for developing new pad formulations with possible replacements of copper content, it is very important to understand the functionality of copper additions to brake friction materials. In the paper theoretical investigation of the role of copper as a pad ingredient was carried out on the basis of modelling by the method of movable cellular automata (MCA). In the study the concentration of copper particles in a Fe3O4-matrix was varied. The sliding simulations were performed while assuming material properties at 500°C in order to assess the beneficial role of copper during severe braking conditions corresponding to fading cycles during dynamometer testing.

  12. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    SciTech Connect

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  13. Using numerical simulation methods to predict the effects of balancing coal and primary air flow rates on furnace emissions

    SciTech Connect

    Schwab, M.J.; Nelson, R.K.; Hardman, R.R.; Facchiano, T.

    1996-12-31

    This paper presents the technical results of a computer modeling exercise to quantify the impacts of balanced and unbalanced coal flows on NO{sub x} emissions and other boiler performance indicators. Using Airflow Sciences Corporation`s proprietary codes, separate computational fluid dynamics models of the furnace region and coal nozzles of a 200 MW{sub e} tangentially-fired boiler equipped with an ABB C-E Services Low NO{sub x} Concentric Firing System (Level II) were constructed. In modeling the coal combustion process, the numerical simulation of gas conditions within the furnace is accomplished by coupling the fluid dynamics relationships with sub-models that predict heat transfer (conduction, convection and radiation), turbulence, coal particle trajectories and temperatures, coal devolatilization, char combustion and equilibrium (mixing limited) chemistry. The equilibrium chemistry sub-model defines concentrations of the products of combustion at all locations within the furnace, with the exception of NO{sub x} concentrations. The generation of NO{sub x} is decoupled from the CFD simulation and is determined using finite-rate chemistry. The model was validated using test results from a recently completed US Department of Energy-sponsored Clean Coal Project at Gulf Power Company`s Plant Lansing Smith Unit 2. Validation was accomplished through comparison of the model results with experimental data including NO{sub x} emissions, unburned carbon, furnace exit gas temperatures, carbon monoxide levels, and excess oxygen values. Following validation, additional simulations were run to quantify the effect of balanced and unbalanced coal flows. Conditions simulated included the as-found condition, a fully balanced condition, a mill-by-mill fully balanced condition, and a {+-}10 percent balanced condition. The results showed that NO{sub x} emissions were not significantly affected by improving the distributions of primary air and coal between the burners.

  14. 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

  15. A large scale simulation of excitation propagation in layer 2/3 of primary and secondary visual cortices of mice.

    PubMed

    Ohtsu, Shoya; Nomura, Taishin; Uno, Shota; Maeda, Kazuki; Hayashida, Yuki; Yagi, Tetsuya

    2015-01-01

    Analyzing network architecture and spatio-temporal dynamics of the visual cortical areas can facilitate understanding visual information processing in the brain. Recently, several physiological experiments utilizing the fast in-vivo imaging technique have demonstrated that the primary visual cortex (V1) and the secondary visual cortex (V2) in mice exhibit complex properties of the responses to visual and electrical stimuli. In order to provide a tool for quantitatively analyzing such a complex dynamics of the cortices at the level of neurons and circuits, here, we constructed a physiologically plausible large-scale network model of the layers 2/3 of V1 and V2, composed of 14,056 multi-compartment neuron models. The Message-Passing-Interface-based parallel simulations of our network model were able to reproduce, at least quantitatively, the neural responses experimentally observed in mouse V1 and V2 with the voltage-sensitive dye imaging.

  16. Simulation study of the correlation (Xmaxμ, Nμ) in view of obtaining information on primary mass of the UHECRs

    NASA Astrophysics Data System (ADS)

    Arsene, Nicusor; Sima, Octavian; Haungs, Andreas; Rebel, Heinigerd

    2016-10-01

    In this paper we study, using Monte Carlo simulations, the possibility to discriminate the mass of the Ultra High Energy Cosmic Rays (UHECRs) by combining information obtained from the maximum Xmaxμ of the muon production rate longitudinal profile of Extensive Air Showers (EAS) and the number of muons, Nμ, which hit an array of detectors located in the horizontal plane. We investigate the sensitivity of the 2D distribution Xmaxμ versus Nμ to the mass of the primary particle generating the air shower. To this purpose we analyze a set of CORSIKA showers induced by protons and iron nuclei at energies of 1019 eV and 1020 eV, at five angles of incidence, 0°, 37°, 48°, 55° and 60°. Using the simulations we obtain the 2D Probability Functions Prob(Xmaxμ, Nμ | p) and Prob(Xmaxμ, Nμ | Fe) which give the probability that a shower induced by a proton or iron nucleus contributes to a specific point on the plane (Xmaxμ, Nμ). Then we construct the probability functions Prob(p | Xmaxμ, Nμ) and Prob(Fe | Xmaxμ, Nμ) which give the probability that a certain point on the plane (Xmaxμ, Nμ) corresponds to a shower initiated by a proton or an iron nucleus, respectively. Finally, a test of this procedure using a Bayesian approach, confirms an improved accuracy of the primary mass estimation in comparison with the results obtained using only the Xmaxμ distributions.

  17. Primary implant stability in a bone model simulating clinical situations for the posterior maxilla: an in vitro study

    PubMed Central

    2016-01-01

    Purpose The aim of this study was to determine the influence of anatomical conditions on primary stability in the models simulating posterior maxilla. Methods Polyurethane blocks were designed to simulate monocortical (M) and bicortical (B) conditions. Each condition had four subgroups measuring 3 mm (M3, B3), 5 mm (M5, B5), 8 mm (M8, B8), and 12 mm (M12, B12) in residual bone height (RBH). After implant placement, the implant stability quotient (ISQ), Periotest value (PTV), insertion torque (IT), and reverse torque (RT) were measured. Two-factor ANOVA (two cortical conditions×four RBHs) and additional analyses for simple main effects were performed. Results A significant interaction between cortical condition and RBH was demonstrated for all methods measuring stability with two-factor ANOVA. In the analyses for simple main effects, ISQ and PTV were statistically higher in the bicortical groups than the corresponding monocortical groups, respectively. In the monocortical group, ISQ and PTV showed a statistically significant rise with increasing RBH. Measurements of IT and RT showed a similar tendency, measuring highest in the M3 group, followed by the M8, the M5, and the M12 groups. In the bicortical group, all variables showed a similar tendency, with different degrees of rise and decline. The B8 group showed the highest values, followed by the B12, the B5, and the B3 groups. The highest coefficient was demonstrated between ISQ and PTV. Conclusions Primary stability was enhanced by the presence of bicortex and increased RBH, which may be better demonstrated by ISQ and PTV than by IT and RT. PMID:27588215

  18. Comparative assessment of primary and secondary infection risks in a norovirus outbreak using a household model simulation.

    PubMed

    Miura, Fuminari; Watanabe, Toru; Watanabe, Kozo; Takemoto, Kazuhiko; Fukushi, Kensuke

    2016-12-01

    Diarrheal diseases can be transmitted via both primary infection due to exposures to contaminated materials from the environment and secondary infection due to person-to-person contacts. Usually, the importance of secondary infection is empirically assessed by fitting mathematical models to the epidemic curves. However, these empirical models may not be applicable to other epidemic cases because they are developed only for the target epidemics and they don't consider the detail routes of infection. In our previous study, we developed a theoretical model taking into account the various routes of infection that commonly occur in households (e.g., shaking hands, food handling, and changing diapers). This model was made flexible and applicable to any epidemics by means of adjusting model parameters. In this study, we proposed a new index "Vulnerability indicator to secondary infection (VISI)", which expressed a ratio of secondary infection to primary infection risks and calculated this index in a simulated norovirus (NoV) epidemic that involved 10,000 households. The results demonstrated that households composed of more than three members including infant(s) had much higher levels of VISI (5-45) than two-member-households with VISI (0.1-4). These results concluded that the infants were likely to be a hub of secondary infections in highly dense families and therefore careful handling of diapers was deemed indispensible in such families to effectively control the secondary infections.

  19. 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)

  20. Single and two-phase natural circulation in Westinghouse pressurized water reactor simulators: Phenomena, analysis and scaling

    SciTech Connect

    Schultz, R.R.; Chapman, J.C.; Kukita, Y.; Motley, F.E.; Stumpf, H.; Chen, Y.S.; Tasaka, K.

    1987-01-01

    Natural circulation data obtained in the 1/48 scale W four loop PWR simulator - the Large Scale Test Facility (LSTF) are discussed and summarized. Core cooling modes, the primary fluid state, the primary loop mass flow and localized natural circulation phenomena occurring in the steam generator are presented. TRAC-PF1 LSTF model (using both a 1 U-tube and a 3 U-tube steam generator model) analyses of the LSTF natural circulation data including the SG recirculation patterns are presented and compared to the data. The LSTF data are then compared to similar natural circulation data obtained in the Primarkreislaufe (PKL) and the Semiscale facilities. Based on the 1/48 to 1/1705 scaling range which exists between the facilities, the implications of these data towrard natural circulation behavior in commercial plants are briefly discussed.

  1. MULTI - TRACER CONTROL ROOM AIR INLEAKAGE PROTOCOL AND SIMULATED PRIMARY AND EXTENDED MULTI - ZONE RESULTS.

    SciTech Connect

    DIETZ,R.N.

    2002-01-01

    The perfluorocarbon tracer (PFT) technology can be applied simultaneously to the wide range in zonal flowrates (from tens of cfms in some Control Rooms to almost 1,000,000 cfm in Turbine Buildings), to achieve the necessary uniform tagging for subsequent determination of the desired air inleakage and outleakage from all zones surrounding a plant's Control Room (CR). New types of PFT sources (Mega sources) were devised and tested to handle the unusually large flowrates in a number of HVAC zones in power stations. A review of the plans of a particular nuclear power plant and subsequent simulations of the tagging and sampling results confirm that the technology can provide the necessary concentration measurement data to allow the important ventilation pathways involving the Control Room and its air flow communications with all adjacent zones to be quantitatively determined with minimal uncertainty. Depending on need, a simple single or 3-zone scheme (involving the Control Room alone or along with the Aux. Bldg. and Turbine Bldg.) or a more complex test involving up to 7 zones simultaneously can be accommodated with the current revisions to the technology; to test all the possible flow pathways, several different combinations of up to 7 zones would need to be run. The potential exists that for an appropriate investment, in about 2 years, it would be possible to completely evaluate an entire power plant in a single extended multizone test with up to 12 to 13 separate HVAC zones. With multiple samplers in the Control Room near each of the contiguous zones, not only will the prevalent inleakage or outleakage zones be documented, but the particular location of the pathway's room of ingress can be identified. The suggested protocol is to perform a 3-zone test involving the Control Room, Aux. Bldg., and Turbine Bldg. to (1) verify CR total inleakage and (2) proportion that inleakage to distinguish that from the other 2 major buildings and any remaining untagged locations

  2. [Rapid analysis of 28 primary aromatic amines in aqueous food simulants by high performance liquid chromatography-tandem mass spectrometry].

    PubMed

    Xiao, Xiaofeng; Wang, Jianling; Yang, Juanjuan; Liu, Tingfei; Chen, Tong; He, Jun; Deng, Hongyi; Gao, Qiyan

    2013-01-01

    A novel method for rapid analysis of the migration amounts of 28 primary aromatic amines (PAAs) in aqueous food simulants (10% ethanol, 30 g/L acetic acid and 20% ethanol aqueous solution) was developed using high performance liquid chromatography-tandem mass spectrometry (HPLC-MS/MS). After the migration test, the soaking solution was cooled down from 100 degrees C, vortexed, filtered through a hydrophilic polytetrafluoroethylene filter with a disposable syringe, and then the filtrate was analyzed by HPLC-MS/MS. A Zorbax SB-Phenyl column (250 mm x 4.6 mm, 5 microm) was selected for chromatography. The mobile phase consisted of water and 0.1% formic acid-25% acetonitrile-methanol solution with gradient elution. The 28 PAAs in aqueous food simulants were detected by tandem mass spectrometer operated in positive electrospray ionization (ESI+) and multiple reaction monitoring (MRM) mode. The limits of quantification for the 28 PAAs were between 0.002 microg/L and 10 microg/L. The linearity of the method was good with correlation coefficients (r2) greater than 0.995 over the concentration range from 5 microg/L or 10 microg/L to 100 microg/L. The average recoveries of the 28 PAAs were between 76.6% and 114% with the relative standard deviations between 1.53% and 8.97% at the levels of 10, 20, and 40 microg/L. The method shows rapid pretreatment, the lower limits of quantification, good recoveries and accuracies, and meets the requirement of European Union (EU) No 10/2011 regulation for the specific migration of PAAs. The method has been applied to analyze the 28 PAAs in different aqueous food simulants from the migration test of 30 batches of food contact material samples exported to EU.

  3. 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)

  4. 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.

  5. Diagnosing the Dynamics of Observed and Simulated Ecosystem Gross Primary Productivity with Time Causal Information Theory Quantifiers.

    PubMed

    Sippel, Sebastian; Lange, Holger; Mahecha, Miguel D; Hauhs, Michael; Bodesheim, Paul; Kaminski, Thomas; Gans, Fabian; Rosso, Osvaldo A

    2016-01-01

    Data analysis and model-data comparisons in the environmental sciences require diagnostic measures that quantify time series dynamics and structure, and are robust to noise in observational data. This paper investigates the temporal dynamics of environmental time series using measures quantifying their information content and complexity. The measures are used to classify natural processes on one hand, and to compare models with observations on the other. The present analysis focuses on the global carbon cycle as an area of research in which model-data integration and comparisons are key to improving our understanding of natural phenomena. We investigate the dynamics of observed and simulated time series of Gross Primary Productivity (GPP), a key variable in terrestrial ecosystems that quantifies ecosystem carbon uptake. However, the dynamics, patterns and magnitudes of GPP time series, both observed and simulated, vary substantially on different temporal and spatial scales. We demonstrate here that information content and complexity, or Information Theory Quantifiers (ITQ) for short, serve as robust and efficient data-analytical and model benchmarking tools for evaluating the temporal structure and dynamical properties of simulated or observed time series at various spatial scales. At continental scale, we compare GPP time series simulated with two models and an observations-based product. This analysis reveals qualitative differences between model evaluation based on ITQ compared to traditional model performance metrics, indicating that good model performance in terms of absolute or relative error does not imply that the dynamics of the observations is captured well. Furthermore, we show, using an ensemble of site-scale measurements obtained from the FLUXNET archive in the Mediterranean, that model-data or model-model mismatches as indicated by ITQ can be attributed to and interpreted as differences in the temporal structure of the respective ecological time

  6. Diagnosing the Dynamics of Observed and Simulated Ecosystem Gross Primary Productivity with Time Causal Information Theory Quantifiers

    PubMed Central

    Sippel, Sebastian; Mahecha, Miguel D.; Hauhs, Michael; Bodesheim, Paul; Kaminski, Thomas; Gans, Fabian; Rosso, Osvaldo A.

    2016-01-01

    Data analysis and model-data comparisons in the environmental sciences require diagnostic measures that quantify time series dynamics and structure, and are robust to noise in observational data. This paper investigates the temporal dynamics of environmental time series using measures quantifying their information content and complexity. The measures are used to classify natural processes on one hand, and to compare models with observations on the other. The present analysis focuses on the global carbon cycle as an area of research in which model-data integration and comparisons are key to improving our understanding of natural phenomena. We investigate the dynamics of observed and simulated time series of Gross Primary Productivity (GPP), a key variable in terrestrial ecosystems that quantifies ecosystem carbon uptake. However, the dynamics, patterns and magnitudes of GPP time series, both observed and simulated, vary substantially on different temporal and spatial scales. We demonstrate here that information content and complexity, or Information Theory Quantifiers (ITQ) for short, serve as robust and efficient data-analytical and model benchmarking tools for evaluating the temporal structure and dynamical properties of simulated or observed time series at various spatial scales. At continental scale, we compare GPP time series simulated with two models and an observations-based product. This analysis reveals qualitative differences between model evaluation based on ITQ compared to traditional model performance metrics, indicating that good model performance in terms of absolute or relative error does not imply that the dynamics of the observations is captured well. Furthermore, we show, using an ensemble of site-scale measurements obtained from the FLUXNET archive in the Mediterranean, that model-data or model-model mismatches as indicated by ITQ can be attributed to and interpreted as differences in the temporal structure of the respective ecological time

  7. 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.

  8. Evaluation and comparison of gross primary production estimates for the Northern Great Plains grasslands

    USGS Publications Warehouse

    Zhang, L.; Wylie, B.; Loveland, T.; Fosnight, E.; Tieszen, L.L.; Ji, L.; Gilmanov, T.

    2007-01-01

    Two spatially-explicit estimates of gross primary production (GPP) are available for the Northern Great Plains. An empirical piecewise regression (PWR) GPP model was developed from flux tower measurements to map carbon flux across the region. The Moderate Resolution Imaging Spectrometer (MODIS) GPP model is a process-based model that uses flux tower data to calibrate its parameters. Verification and comparison of the regional PWR GPP and the global MODIS GPP are important for the modeling of grassland carbon flux. This study compared GPP estimates from PWR and MODIS models with five towers in the grasslands. Among them, PWR GPP and MODIS GPP showed a good agreement with tower-based GPP at three towers. The global MODIS GPP, however, did not agree well with tower-based GPP at two other towers, probably because of the insensitivity of MODIS model to regional ecosystem and climate change and extreme soil moisture conditions. Cross-validation indicated that the PWR model is relatively robust for predicting regional grassland GPP. However, the PWR model should include a wide variety of flux tower data as the training data sets to obtain more accurate results. In addition, GPP maps based on the PWR and MODIS models were compared for the entire region. In the northwest and south, PWR GPP was much higher than MODIS GPP. These areas were characterized by the higher water holding capacity with a lower proportion of C4 grasses in the northwest and a higher proportion of C4 grasses in the south. In the central and southeastern regions, PWR GPP was much lower than MODIS GPP under complicated conditions with generally mixed C3/C4 grasses. The analysis indicated that the global MODIS GPP model has some limitations on detecting moisture stress, which may have been caused by the facts that C3 and C4 grasses are not distinguished, water stress is driven by vapor pressure deficit (VPD) from coarse meteorological data, and MODIS land cover data are unable to differentiate the sub

  9. Deriving therapies for children with primary CNS tumors using pharmacokinetic modeling and simulation of cerebral microdialysis data.

    PubMed

    Jacus, M O; Throm, S L; Turner, D C; Patel, Y T; Freeman, B B; Morfouace, M; Boulos, N; Stewart, C F

    2014-06-16

    The treatment of children with primary central nervous system (CNS) tumors continues to be a challenge despite recent advances in technology and diagnostics. In this overview, we describe our approach for identifying and evaluating active anticancer drugs through a process that enables rational translation from the lab to the clinic. The preclinical approach we discuss uses tumor subgroup-specific models of pediatric CNS tumors, cerebral microdialysis sampling of tumor extracellular fluid (tECF), and pharmacokinetic modeling and simulation to overcome challenges that currently hinder researchers in this field. This approach involves performing extensive systemic (plasma) and target site (CNS tumor) pharmacokinetic studies. Pharmacokinetic modeling and simulation of the data derived from these studies are then used to inform future decisions regarding drug administration, including dosage and schedule. Here, we also present how our approach was used to examine two FDA approved drugs, simvastatin and pemetrexed, as candidates for new therapies for pediatric CNS tumors. We determined that due to unfavorable pharmacokinetic characteristics and insufficient concentrations in tumor tissue in a mouse model of ependymoma, simvastatin would not be efficacious in further preclinical trials. In contrast to simvastatin, pemetrexed was advanced to preclinical efficacy studies after our studies determined that plasma exposures were similar to those in humans treated at similar tolerable dosages and adequate unbound concentrations were found in tumor tissue of medulloblastoma-bearing mice. Generally speaking, the high clinical failure rates for CNS drug candidates can be partially explained by the fact that therapies are often moved into clinical trials without extensive and rational preclinical studies to optimize the transition. Our approach addresses this limitation by using pharmacokinetic and pharmacodynamic modeling of data generated from appropriate in vivo models to

  10. Deriving Therapies for Children with Primary CNS Tumors Using Pharmacokinetic Modeling and Simulation of Cerebral Microdialysis Data

    PubMed Central

    Jacus, M.O.; Throm, S.L.; Turner, D.C.; Patel, Y.T.; Freeman, B.B.; Morfouace, M.; Boulos, N.; Stewart, C. F.

    2014-01-01

    The treatment of children with primary central nervous system (CNS) tumors continues to be a challenge despite recent advances in technology and diagnostics. In this overview, we describe our approach for identifying and evaluating active anticancer drugs through a process that enables rational translation from the lab to the clinic. The preclinical approach we discuss uses tumor subgroup-specific models of pediatric CNS tumors, cerebral microdialysis sampling of tumor extracellular fluid (tECF), and pharmacokinetic modeling and simulation to overcome challenges that currently hinder researchers in this field. This approach involves performing extensive systemic (plasma) and target site (CNS tumor) pharmacokinetic studies. Pharmacokinetic modeling and simulation of the data derived from these studies are then used to inform future decisions regarding drug administration, including dosage and schedule. Here, we also present how our approach was used to examine two FDA approved drugs, simvastatin and pemetrexed, as candidates for new therapies for pediatric CNS tumors. We determined that due to unfavorable pharmacokinetic characteristics and insufficient concentrations in tumor tissue in a mouse model of ependymoma, simvastatin would not be efficacious in further preclinical trials. In contrast to simvastatin, pemetrexed was advanced to preclinical efficacy studies after our studies determined that plasma exposures were similar to those in humans treated at similar tolerable dosages and adequate unbound concentrations were found in tumor tissue of medulloblastoma-bearing mice. Generally speaking, the high clinical failure rates for CNS drug candidates can be partially explained by the fact that therapies are often moved into clinical trials without extensive and rational preclinical studies to optimize the transition. Our approach addresses this limitation by using pharmacokinetic and pharmacodynamic modeling of data generated from appropriate in vivo models to

  11. The association between net primary productivity and rainfall in CMIP5 20th and 21st century simulations

    NASA Astrophysics Data System (ADS)

    Negron Juarez, R. I.; Riley, W. J.; Koven, C. D.; Knox, R. G.; Taylor, P.; Chambers, J. Q.

    2014-12-01

    Tropical forests fix large amounts of atmospheric CO2 into biomass via net primary productivity (NPP). In this study we use the NPP-MAR (mean annual rainfall) relationship observed in tropical forests to evaluate the performance (20th century) and predictions (21st century) of tropical NPP from ten Coupled Model Intercomparison Project Phase 5 (CMIP5) Earth System Models (ESMs). Most of the CMIP5 models showed an increase in NPP concurrent with an increase in rainfall and a decrease in surface solar radiation. Models that better represented the NPP-MAR pattern did not better represent the climate, and vice versa. By the end of the 21st century the models that best reproduced the observed NPP-MAR relationship projected an increases in NPP between ~2% (RCP 4.5) and ~19% (RCP 8.5) relative to current observations (11.88±5 MgC ha-1 yr-1, 327 field sites) and increases of ~9% and ~25% relative to historical simulations (2005). By separating the effects of climate forcing and CO2 fertilization models showed that maximum productivity is likely occurring during the current climate, but this signal is masked by increases in NPP due to CO2 fertilization. Further studies addressing the individual and simultaneous effect of other climate variables on NPP are needed.

  12. 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.

  13. 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.

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

    SciTech Connect

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

    2001-04-04

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

  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. 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)

  17. 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.

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

    SciTech Connect

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

    2014-09-30

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

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Evaluating Simulations of Primary Anthropogenic and Biomass Burning Organic Aerosols using Aerosol Mass Spectrometer Data and Positive Matrix Factorization Analysis

    NASA Astrophysics Data System (ADS)

    Fast, J.; Aiken, A.; Alexander, L.; Canagaratna, M.; Decarlo, P.; Herndon, S.; Jimenez, J.; Kleinman, L.; Ochoa, C.; Onasch, T.; Song, C.; Wiedinmyer, C.; Yu, X.; Zaveri, R.

    2008-12-01

    Most model predictions of organic matter are currently underestimated because the processes contributing to secondary organic aerosol (SOA) formation and transformation are not well understood. Since research associated with developing a better framework to improve the representation of specific gas-to-particle partitioning processes controlling SOA based on new measurements and theoretical relationships is on- going, this study seeks to determine whether 3-D models can adequately predict concentrations of primary organic aerosols (POA). If one assumes POA is non-volatile, then errors in POA predictions will results from uncertainties in the emission inventories and errors in transport and mixing processes. The WRF-chem model is used to predict POA in the vicinity of Mexico City during the 2006 MILAGRO field campaign. Particulate matter emission rates were obtained from urban and regional Mexican emission inventories and from biomass burning estimates derived from MODIS "hotspot" and vegetation databases. Organic aerosol predictions are evaluated using data from Aerodyne Aerosol Mass Spectrometer (AMS) instruments deployed at four ground sites and on two research aircraft and from Sunset Laboratory OCEC instruments deployed at two ground sites. Positive Matrix Factorization (PMF) has recently been applied to derive components of organic aerosols including: hydrocarbon-like organic aerosol (HOA), oxidized organic aerosol (OOA), and biomass burning organic aerosols (BBOA). The temporal variation of HOA is often similar to primary emissions of other species in urban areas. PMF analysis is currently available for three of the ground sites and for some of the aircraft flights. We found that the predicted POA was consistently lower than the measured organic matter at the ground sites, which is consistent with the expectation that SOA should be a large fraction of the total organic aerosol mass. A much better agreement was found when predicted POA was compared with HOA

  6. 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

  7. Using discrete event simulation to compare the performance of family health unit and primary health care centre organizational models in Portugal

    PubMed Central

    2011-01-01

    Background Recent reforms in Portugal aimed at strengthening the role of the primary care system, in order to improve the quality of the health care system. Since 2006 new policies aiming to change the organization, incentive structures and funding of the primary health care sector were designed, promoting the evolution of traditional primary health care centres (PHCCs) into a new type of organizational unit - family health units (FHUs). This study aimed to compare performances of PHCC and FHU organizational models and to assess the potential gains from converting PHCCs into FHUs. Methods Stochastic discrete event simulation models for the two types of organizational models were designed and implemented using Simul8 software. These models were applied to data from nineteen primary care units in three municipalities of the Greater Lisbon area. Results The conversion of PHCCs into FHUs seems to have the potential to generate substantial improvements in productivity and accessibility, while not having a significant impact on costs. This conversion might entail a 45% reduction in the average number of days required to obtain a medical appointment and a 7% and 9% increase in the average number of medical and nursing consultations, respectively. Conclusions Reorganization of PHCC into FHUs might increase accessibility of patients to services and efficiency in the provision of primary care services. PMID:21999336

  8. Amorphous and Nanocrystalline High Temperature Magnetic Material for PWR

    DTIC Science & Technology

    2006-03-01

    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...times that of conventional ferrites at room temperature); 2) Frequency: 200 kHz to 1 MHz; 3) Temperature: 200 °C and above. The goals of the DUST...Department in DUST Program 44 Appendix III: Benchmark core loss comparisons between HITPERM and Magnetics, Inc. ferrite cores. 46 Appendix IV

  9. 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.

  10. 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)

  11. Prediction of phase distribution pattern in phase field simulations on Mo5SiB2-primary areas in near eutectic Mo-Si-B alloy

    NASA Astrophysics Data System (ADS)

    Kazemi, O.; Hasemann, G.; Krüger, M.; Halle, T.

    2017-03-01

    A Mo-10.9Si-20.3B (if not stated otherwise all compositions are given in at.%) alloy was modeled using the phase field method with linearized phase diagrams and thermodynamic data. The simulation results showed that there are two specific microstructural constituents. According to the simulations and experimental microstructural investigations the primary Mo5SiB2 phase observed in this study was combined by the Moss-Mo3Si-Mo5SiB2 eutectics. The overall composition of the both primary and eutectic area was probed and the phase evolution along with the concentration change of the core areas were explored and visualized in the solidification domain. To verify the accuracy of the simulation results, they were comprised with the experimental achievements. In terms of the fraction of phases, the portion of the primary phase and the eutectic constituent and the phase distribution pattern our results were in good agreement with the experimental observations.

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

    SciTech Connect

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

    1989-01-01

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

  13. 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.

  14. 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.

  15. 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.

  16. 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).

  17. 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.

  18. 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

  19. Monte Carlo simulation of electron depth distribution and backscattering for carbon films deposited on aluminium as a function of incidence angle and primary energy

    NASA Astrophysics Data System (ADS)

    Dapor, Maurizio

    2005-01-01

    Carbon films are deposited on various substrates (polymers, polyester fabrics, polyester yarns, metal alloys) both for experimental and technological motivations (medical devices, biocompatible coatings, food package and so on). Computational studies of the penetration of electron beams in supported thin film of carbon are very useful in order to compare the simulated results with analytical techniques data (obtained by scanning electron microscopy and/or Auger electron spectroscopy) and investigate the film characteristics. In the present paper, the few keV electron depth distribution and backscattering coefficient for the special case of film of carbon deposited on aluminium are investigated, by a Monte Carlo simulation, as a function of the incidence angle and primary electron energy. The simulated results can be used as a way to evaluate the carbon film thickness by a set of measurements of the backscattering coefficient.

  20. Some observations on simulated molten debris-coolant layer dynamics. [PWR; BWR

    SciTech Connect

    Greene, G.A.; Klein, J.; Klages, J.; Schwarz, E.; Sanborn, Y.

    1983-04-01

    Experiments are being performed to investigate high temperature liquid-liquid film boiling between a pool of liquid metal and an overlying coolant pool of R-11 or water. Film boiling has been observed to be stable for R-11; however, considerable liquid-liquid contact has been observed with water well beyond the minimum film boiling temperature. Unstable liquid-liquid film boiling of water has been observed to escalate into dispersive, non-energetic vapor explosions when the interface contact temperature exceeded the spontaneous nucleation temperature. Other parametric trends in the data are discussed.

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

    DTIC Science & Technology

    1987-06-01

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

  2. 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.

  3. 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.

  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. Sci—Fri PM: Topics — 07: Monte Carlo Simulation of Primary Dose and PET Isotope Production for the TRIUMF Proton Therapy Facility

    SciTech Connect

    Lindsay, C; Jirasek, A; Blackmore, E; Hoehr, C; Schaffer, P; Trinczek, M; Sossi, V

    2014-08-15

    Uveal melanoma is a rare and deadly tumour of the eye with primary metastases in the liver resulting in an 8% 2-year survival rate upon detection. Large growths, or those in close proximity to the optic nerve, pose a particular challenge to the commonly employed eye-sparing technique of eye-plaque brachytherapy. In these cases external beam charged particle therapy offers improved odds in avoiding catastrophic side effects such as neuropathy or blindness. Since 1995, the British Columbia Cancer Agency in partnership with the TRIUMF national laboratory have offered proton therapy in the treatment of difficult ocular tumors. Having seen 175 patients, yielding 80% globe preservation and 82% metastasis free survival as of 2010, this modality has proven to be highly effective. Despite this success, there have been few studies into the use of the world's largest cyclotron in patient care. Here we describe first efforts of modeling the TRIUMF dose delivery system using the FLUKA Monte Carlo package. Details on geometry, estimating beam parameters, measurement of primary dose and simulation of PET isotope production are discussed. Proton depth dose in both modulated and pristine beams is successfully simulated to sub-millimeter precision in range (within limits of measurement) and 2% agreement to measurement within in a treatment volume. With the goal of using PET signals for in vivo dosimetry (alignment), a first look at PET isotope depth distribution is presented — comparing favourably to a naive method of approximating simulated PET slice activity in a Lucite phantom.

  6. Dose and linear energy transfer distributions of primary and secondary particles in carbon ion radiation therapy: A Monte Carlo simulation study in water

    PubMed Central

    Johnson, Daniel; Chen, Yong; Ahmad, Salahuddin

    2015-01-01

    The factors influencing carbon ion therapy can be predicted from accurate knowledge about the production of secondary particles from the interaction of carbon ions in water/tissue-like materials, and subsequently the interaction of the secondary particles in the same materials. The secondary particles may have linear energy transfer (LET) values that potentially increase the relative biological effectiveness of the beam. Our primary objective in this study was to classify and quantify the secondary particles produced, their dose averaged LETs, and their dose contributions in the absorbing material. A 1 mm diameter carbon ion pencil beam with energies per nucleon of 155, 262, and 369 MeV was used in a geometry and tracking 4 Monte Carlo simulation to interact in a 27 L water phantom containing 3000 rectangular detector voxels. The dose-averaged LET and the dose contributions of primary and secondary particles were calculated from the simulation. The results of the simulations show that the secondary particles that contributed a major dose component had LETs <100 keV/µm. The secondary particles with LETs >600 keV/µm contributed only <0.3% of the dose. PMID:26865757

  7. Simulation of the ATIC-2 Silicon Matrix for Protons and Helium GCR Primaries at 0.3, 10, and 25 TeV/Nucleon

    NASA Technical Reports Server (NTRS)

    Watts, J.; Adams, J. H.; Bashindzhagyan, G.; Batkov, K. E.; Chang, J.; Christl, M.; Fazely, A. R.; Ganel, O.; Gunasingha R. M.; Guzik, T. G.

    2005-01-01

    The energy deposition distribution for protons and helium galactic cosmic ray primaries at 0.3, 10, and 25 TeV/nucleon in the ATIC-2 silicon matrix detector are simulated with GEANT4. The GEANT3 geometrical model of ATIC developed by the University of Maryland was combined with a GEANT4 application developed for the Deep Space Test Bed (DSTB) detector package. The new code included relatively minor modifications to completely describe the ATIC materials and a more detailed model of the Silicon Matrix detector. For this analysis all particles were started as a unidirectional beam at a single point near the center of the Silicon Matrix front surface. The point was selected such that each primary passed through at least two of the overlapping silicon pixels.

  8. Simulation of Human Plasma Concentrations of Thalidomide and Primary 5-Hydroxylated Metabolites Explored with Pharmacokinetic Data in Humanized TK-NOG Mice.

    PubMed

    Nishiyama, Sayako; Suemizu, Hiroshi; Shibata, Norio; Guengerich, F Peter; Yamazaki, Hiroshi

    2015-11-16

    Plasma concentrations of thalidomide and primary 5-hydroxylated metabolites including 5,6-dihydroxythalidomide and glutathione (GSH) conjugate(s) were investigated in chimeric mice with highly "humanized" liver cells harboring cytochrome P450 3A5*1. Following oral administration of thalidomide (100 mg/kg), plasma concentrations of GSH conjugate(s) of 5-hydroxythalidomide were higher in humanized mice than in controls. Simulation of human plasma concentrations of thalidomide were achieved with a simplified physiologically based pharmacokinetic model in accordance with reported thalidomide concentrations. The results indicate that the pharmacokinetics in humans of GSH conjugate and/or catechol primary 5-hydroxylated thalidomide contributing in vivo activation can be estimated for the first time.

  9. 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.

  10. An Investigation of Attitudes towards the Use of Computer Games and Simulations in the Primary/Elementary Classroom Environment.

    ERIC Educational Resources Information Center

    Godfreyson, John Ernst

    This report summarizes an investigation of the attitudes of elementary students, teachers, and parents toward the use of educational computer games and simulations in a teaching/learning situation. Following a statement of the project purpose and a literature review, the courseware selection process is described. (This process resulted in use of…

  11. Modeling, simulation and analysis of the evacuation process on stairs in a multi-floor classroom building of a primary school

    NASA Astrophysics Data System (ADS)

    Li, Wenhang; Li, Yi; Yu, Ping; Gong, Jianhua; Shen, Shen; Huang, Lin; Liang, Jianming

    2017-03-01

    Few studies have focused on the evacuation of multi-floor classroom buildings in a primary school, a process that differs from evacuations in other buildings. A stair-unit model was proposed to describe the spatial topology of twisting stairwells and to describe the spatial relationship between stairwells and floors. Based on the stair-unit model, a schedule-line model was proposed to calculate evacuation paths in stair-units; a modified algorithm to calculate pedestrian forces were proposed to describe the evacuee movements in stairwells; and a projection strategy was proposed to model the 3-dimensional evacuation process in multi-floor buildings. The simulated processes were compared with a real evacuation drill. The results showed that the simulated process achieved qualitative and quantitative consistencies with the real drill, proving the appropriateness of the proposed models and algorithms. Based on the validation, further simulations were conducted and a few rules for evacuations in stairwells were identified including rules governing the impact of the moment of entering a staircase, the number of students in a class, the stagger strategy, and the layout of grades on different floors on the time in stairwell and the total evacuation duration. The results can be used to mitigate the effects of a fire disaster, and the proposed models and algorithms can also be referenced by evacuation simulation for other multi-floor buildings such as residential buildings.

  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. Combatting Ionic Aggregation using Dielectric Forces Combining Modeling/Simulation and Experimental Results to Explain End-capping of Primary Amine Functionalized Polystyrene

    SciTech Connect

    Messman, Jamie M; Goswami, Monojoy; Pickel, Deanna L; Uhrig, David; Sumpter, Bobby G; Mays, Jimmy

    2011-01-01

    Chain-end functionalization of living poly(styryl)lithium using 1-(3-bromopropyl)-2,2,5,5-tetramethyl-1-aza-2,5-disilacyclo-pentane (BTDP) to generate primary amine end-functionalized polystyrene was investigated using high vacuum anionic polymerization techniques. 13C NMR spectroscopy and Matrix Assisted Laser Desorption Ionization Time of Flight Mass Spectrometry (MALDI-TOF MS) were used to evaluate polymer end-groups and demonstrated that quantitative amine functionalized polymer was attained under appropriate reaction conditions. In general, the polymerization of styrene was conducted in benzene and the end-capping reaction was performed by adding tetrahydrofuran (THF) to the reaction prior to the addition of BTDP in THF at room temperature. Results indicated that approximately 20% THF by volume is required to obtain 100% end-capping free from side reactions. When too little or no THF was present, side reactions such as lithium halogen exchange followed by Wurtz coupling resulted in unfunctionalized head-to-head dimer as well as other byproducts. Modeling and simulation of the solvent effects using hybrid methods (the so-called QM/MM method) suggest that THF effectively dissociated the anionic chain-end aggregation, thereby resulting in the desired primary amine functionalized polymer. Molecular dynamics (MD) simulations were conducted to develop an understanding of the physics of counterions involved in the end-functionalization process.

  14. Piloted simulation tests of propulsion control as backup to loss of primary flight controls for a mid-size jet transport

    NASA Technical Reports Server (NTRS)

    Bull, John; Mah, Robert; Davis, Gloria; Conley, Joe; Hardy, Gordon; Gibson, Jim; Blake, Matthew; Bryant, Don; Williams, Diane

    1995-01-01

    Failures of aircraft primary flight-control systems to aircraft during flight have led to catastrophic accidents with subsequent loss of lives (e.g. , DC-1O crash, B-747 crash, C-5 crash, B-52 crash, and others). Dryden Flight Research Center (DFRC) investigated the use of engine thrust for emergency flight control of several airplanes, including the B-720, Lear 24, F-15, C-402, and B-747. A series of three piloted simulation tests have been conducted at Ames Research Center to investigate propulsion control for safely landing a medium size jet transport which has experienced a total primary flight-control failure. The first series of tests was completed in July 1992 and defined the best interface for the pilot commands to drive the engines. The second series of tests was completed in August 1994 and investigated propulsion controlled aircraft (PCA) display requirements and various command modes. The third series of tests was completed in May 1995 and investigated PCA full-flight envelope capabilities. This report describes the concept of a PCA, discusses pilot controls, displays, and procedures; and presents the results of piloted simulation evaluations of the concept by a cross-section of air transport pilots.

  15. 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.

  16. 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%.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. 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)

  7. 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).

  8. 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.

  9. 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.

  10. Simulating spatiotemporal dynamics of sichuan grassland net primary productivity using the CASA model and in situ observations.

    PubMed

    Tang, Chuanjiang; Fu, Xinyu; Jiang, Dong; Fu, Jingying; Zhang, Xinyue; Zhou, Su

    2014-01-01

    Net primary productivity (NPP) is an important indicator for grassland resource management and sustainable development. In this paper, the NPP of Sichuan grasslands was estimated by the Carnegie-Ames-Stanford Approach (CASA) model. The results were validated with in situ data. The overall precision reached 70%; alpine meadow had the highest precision at greater than 75%, among the three types of grasslands validated. The spatial and temporal variations of Sichuan grasslands were analyzed. The absorbed photosynthetic active radiation (APAR), light use efficiency (ε), and NPP of Sichuan grasslands peaked in August, which was a vigorous growth period during 2011. High values of APAR existed in the southwest regions in altitudes from 2000 m to 4000 m. Light use efficiency (ε) varied in the different types of grasslands. The Sichuan grassland NPP was mainly distributed in the region of 3000-5000 m altitude. The NPP of alpine meadow accounted for 50% of the total NPP of Sichuan grasslands.

  11. Multiparametric topological analysis (MTA) for the study of the primary CR composition: Performances with Auger simulated data

    NASA Astrophysics Data System (ADS)

    D'Urso, D.; Ambrosio, M.; Aramo, C.; Guarino, F.; Valore, L.; Pierre Auger Collaboration

    2008-04-01

    We describe the application of a multiparametric analysis to estimate the UHE Cosmic Rays composition. The proposed method, MTA (Multiparametric Topological Analysis), is based on the study of the correlations among different shower observables. This technique is designed to fully exploit the complementarity of Auger fluorescence and ground array data. In the present work, we report the results of the application to Conex showers, fully simulated through the Auger detector, using only parameters describing the longitudinal development of air showers as recorded by fluorescence detector for hybrid data.

  12. 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

  13. Effect of Charge Relaxation in Three-Dimensional Numerical Simulations of Turbulent Primary Atomization of Electrically Charged Liquid Jets

    NASA Astrophysics Data System (ADS)

    Courtine, Emilien; van Poppel, Bret; Daily, John; Desjardins, Olivier

    2012-11-01

    Electrohydrodynamics (EHD) is an interdisciplinary topic that describes the complex interaction between fluid mechanics and electric fields. In the context of combustion applications, EHD may enable improved spray control and finer atomization so that fuel injection schemes can be inexpensively developed for small engines. Moreover, EHD may provide efficient enhancements to hydrocarbon fuel atomization that could benefit a much broader range of engines and non-combustion applications. In this work, high-fidelity numerical simulations of an electrically charged kerosene jet undergoing turbulent atomization are presented. The simulations make use of first-principle-based methods designed to accurately represent the interfacial stresses and discontinuities. Under the assumption of a large electric Reynolds number, it can be appropriate to assume that the charges do not have time to relax to the liquid-gas interface, and that they do not drift within the liquid volume. Alternatively, one can solve a free charge conservation equation to fully account for charge drift. These two approaches are compared in details, and the role of charge drift in EHD atomization is analyzed. The implementation of the charge transport equation, which is discontinuous in nature, is discussed as well.

  14. 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)

  15. Quantification and attribution of errors in the simulated annual gross primary production and latent heat fluxes by two global land surface models

    NASA Astrophysics Data System (ADS)

    Li, Jianduo; Wang, Ying-Ping; Duan, Qingyun; Lu, Xingjie; Pak, Bernard; Wiltshire, Andy; Robertson, Eddy; Ziehn, Tilo

    2016-09-01

    Differences in the predicted carbon and water fluxes by different global land models have been quite large and have not decreased over the last two decades. Quantification and attribution of the uncertainties of global land surface models are important for improving the performance of global land surface models, and are the foci of this study. Here we quantified the model errors by comparing the simulated monthly global gross primary productivity (GPP) and latent heat flux (LE) by two global land surface models with the model-data products of global GPP and LE from 1982 to 2005. By analyzing model parameter sensitivities within their ranges, we identified about 2-11 most sensitive model parameters that have strong influences on the simulated GPP or LE by two global land models, and found that the sensitivities of the same parameters are different among the plant functional types (PFT). Using parameter ensemble simulations, we found that 15%-60% of the model errors were reduced by tuning only a few (<4) most sensitive parameters for most PFTs, and that the reduction in model errors varied spatially within a PFT or among different PFTs. Our study shows that future model improvement should optimize key model parameters, particularly those parameters relating to leaf area index, maximum carboxylation rate, and stomatal conductance.

  16. Predicting the apparent viscosity and yield stress of mixtures of primary, secondary and anaerobically digested sewage sludge: Simulating anaerobic digesters.

    PubMed

    Markis, Flora; Baudez, Jean-Christophe; Parthasarathy, Rajarathinam; Slatter, Paul; Eshtiaghi, Nicky

    2016-09-01

    Predicting the flow behaviour, most notably, the apparent viscosity and yield stress of sludge mixtures inside the anaerobic digester is essential because it helps optimize the mixing system in digesters. This paper investigates the rheology of sludge mixtures as a function of digested sludge volume fraction. Sludge mixtures exhibited non-Newtonian, shear thinning, yield stress behaviour. The apparent viscosity and yield stress of sludge mixtures prepared at the same total solids concentration was influenced by the interactions within the digested sludge and increased with the volume fraction of digested sludge - highlighted using shear compliance and shear modulus of sludge mixtures. However, when a thickened primary - secondary sludge mixture was mixed with dilute digested sludge, the apparent viscosity and yield stress decreased with increasing the volume fraction of digested sludge. This was caused by the dilution effect leading to a reduction in the hydrodynamic and non-hydrodynamic interactions when dilute digested sludge was added. Correlations were developed to predict the apparent viscosity and yield stress of the mixtures as a function of the digested sludge volume fraction and total solids concentration of the mixtures. The parameters of correlations can be estimated using pH of sludge. The shear and complex modulus were also modelled and they followed an exponential relationship with increasing digested sludge volume fraction.

  17. Simulating Spatiotemporal Dynamics of Sichuan Grassland Net Primary Productivity Using the CASA Model and In Situ Observations

    PubMed Central

    Tang, Chuanjiang; Fu, Xinyu; Jiang, Dong; Zhang, Xinyue; Zhou, Su

    2014-01-01

    Net primary productivity (NPP) is an important indicator for grassland resource management and sustainable development. In this paper, the NPP of Sichuan grasslands was estimated by the Carnegie-Ames-Stanford Approach (CASA) model. The results were validated with in situ data. The overall precision reached 70%; alpine meadow had the highest precision at greater than 75%, among the three types of grasslands validated. The spatial and temporal variations of Sichuan grasslands were analyzed. The absorbed photosynthetic active radiation (APAR), light use efficiency (ε), and NPP of Sichuan grasslands peaked in August, which was a vigorous growth period during 2011. High values of APAR existed in the southwest regions in altitudes from 2000 m to 4000 m. Light use efficiency (ε) varied in the different types of grasslands. The Sichuan grassland NPP was mainly distributed in the region of 3000–5000 m altitude. The NPP of alpine meadow accounted for 50% of the total NPP of Sichuan grasslands. PMID:25250396

  18. 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.

  19. 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.

  20. 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.

  1. Primary damage in tungsten using the binary collision approximation, molecular dynamic simulations and the density functional theory

    NASA Astrophysics Data System (ADS)

    De Backer, A.; Sand, A.; Ortiz, C. J.; Domain, C.; Olsson, P.; Berthod, E.; Becquart, C. S.

    2016-02-01

    The damage produced by primary knock-on atoms (PKA) in W has been investigated from the threshold displacement energy (TDE) where it produces one self interstitial atom-vacancy pair to larger energies, up to 100 keV, where a large molten volume is formed. The TDE has been determined in different crystal directions using the Born-Oppenheimer density functional molecular dynamics (DFT-MD). A significant difference has been observed without and with the semi-core electrons. Classical MD has been used with two different empirical potentials characterized as ‘soft’ and ‘hard’ to obtain statistics on TDEs. Cascades of larger energy have been calculated, with these potentials, using a model that accounts for electronic losses (Sand et al 2013 Europhys. Lett. 103 46003). Two other sets of cascades have been produced using the binary collision approximation (BCA): a Monte Carlo BCA using SDTrimSP (Eckstein et al 2011 SDTrimSP: Version 5.00. Report IPP 12/8) (similar to SRIM www.srim.org) and MARLOWE (RSICC Home Page. (https://rsicc.ornl.gov/codes/psr/psr1/psr-137.html) (accessed May, 2014)). The comparison of these sets of cascades gave a recombination distance equal to 12 Å which is significantly larger from the one we reported in Hou et al (2010 J. Nucl. Mater. 403 89) because, here, we used bulk cascades rather than surface cascades which produce more defects (Stoller 2002 J. Nucl. Mater. 307 935, Nordlund et al 1999 Nature 398 49). Investigations on the defect clustering aspect showed that the difference between BCA and MD cascades is considerably reduced after the annealing of the cascade debris at 473 K using our Object Kinetic Monte Carlo model, LAKIMOCA (Domain et al 2004 J. Nucl. Mater. 335 121).

  2. 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.

  3. 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.

  4. Equilizing of the Primary Stress State in the Rock Mass, Simulated by a Model of Layer in an Elastic-Viscous Medium

    NASA Astrophysics Data System (ADS)

    Kortas, Grzegorz

    2016-12-01

    This paper is devoted to the analysis of the stress development process in the homogeneous and non-homogeneous rock mass. The rock-mass model consists of an elastic-viscous medium containing a layer (Fig. 1) that displays distinct geomechanical strain properties. When examining the process of stress equilizing in time, the Norton-Bailey power creep law was applied in the numerical analysis. The relationship between effective stresses and time, the modulus of elasticity, Poisson's coefficient, and creep compliance were obtained. It was demonstrated that the relationship between effective stress and time or creep compliance, for the assumed conditions in a homogeneous rock-mass, was approximated by hyperbolic functions (10 and 16). The process parameter included a certain value of creep compliance or of time at which there occurred a half-way equilizing of primary stresses. An analogous function binds effective stresses with creep compliance. Our model studies indicated a number of relationships between bulk and shear strain with time and creep compliance in the homogeneous and non-homogeneous rock mass, presented in Figs. 2-14, expressed by the functions of those specific parameters. The relationships obtained in this work resulted from our model assumptions. However, they demonstrated the influence of the geomechanical strain properties of rocks on the process of shaping the primary stress state in the rock mass and the tendency to reduce the principal stress differences in time. Our research results suggested the necessity to simulate the primary stress state as an initial condition of the geomechanical numerical analysis concerning the rock-mass behaviour showing rheological properties.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Specific determination of 20 primary aromatic amines in aqueous food simulants by liquid chromatography-electrospray ionization-tandem mass spectrometry.

    PubMed

    Mortensen, Sarah Kelly; Trier, Xenia Thorsager; Foverskov, Annie; Petersen, Jens Hójslev

    2005-10-14

    A multi-analyte method without any pre-treatment steps using reversed-phase liquid chromatography-electrospray ionization-tandem mass spectrometry (LC-ESI-MS/MS) was developed and applied for the determination of 20 primary aromatic amines (PAA) associated with polyurethane (PUR) products or azo-colours. The method was validated in-house for water and 3% acetic acid food simulants using spiked migrates from plastic laminates. Detection limits ranged from 0.27 to 3 microg amine/L food simulants, and RSD values of within-laboratory reproducibility at the 2 microg PAA/L level ranged from 3.9 to 19%. PAA migration from plastic laminates and black nylon cooking utensils were determined with the method, and high levels of 4,4'-methylenedianiline and aniline were found in migrates from about half of the tested cooking utensils. The method fulfils present legislative demands in the EU for screening and verification of PAA migration from food contact materials.

  10. 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)

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

    SciTech Connect

    Crowell, Shannon L.; Alzheimer, James M.

    2011-08-08

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

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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)

  18. 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)

  19. 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.

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

    SciTech Connect

    Cheverton, R.D.

    1982-01-01

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

  1. 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)

  2. Simulation of a main steam line break with steam generator tube rupture using trace

    SciTech Connect

    Gallardo, S.; Querol, A.; Verdu, G.

    2012-07-01

    A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

  3. Evaluation of scatter-to-primary ratio, grid performance and normalized average glandular dose in mammography by Monte Carlo simulation including interference and energy broadening effects

    NASA Astrophysics Data System (ADS)

    Cunha, D. M.; Tomal, A.; Poletti, M. E.

    2010-08-01

    In this work, a computational code for the study of imaging systems and dosimetry in conventional and digital mammography through Monte Carlo simulations is described. The developed code includes interference and Doppler energy broadening for simulation of elastic and inelastic photon scattering, respectively. The code estimates the contribution of scattered radiation to image quality through the spatial distribution of the scatter-to-primary ratio (S/P). It allows the inclusion of different designs of anti-scatter grids (linear or cellular), for evaluation of contrast improvement factor (CIF), Bucky factor (BF) and signal difference-to-noise ratio improvement factor (SIF). It also allows the computation of the normalized average glandular dose, \\bar{D}_{g,N} . These quantities were studied for different breast thicknesses and compositions, anode/filter combinations and tube potentials. Results showed that the S/P increases linearly with breast thickness, varying slightly with breast composition or the spectrum used. Evaluation of grid performance showed that the cellular grid provides the highest CIF with smaller BF. The SIF was also greater for the cellular grid, although both grids showed SIF < 1 for thin breasts. Results for \\bar{D}_{g,N} showed that it increases with the half-value layer (HVL) of the spectrum, decreases considerably with breast thickness and has a small dependence on the anode/filter combination. Inclusion of interference effects of breast tissues affected the values of S/P obtained with the grid up to 25%, while the energy broadening effect produced smaller variations on the evaluated quantities.

  4. Solar-simulated ultraviolet radiation induces histone 3 methylation changes in the gene promoters of matrix metalloproteinases 1 and 3 in primary human dermal fibroblasts.

    PubMed

    Gesumaria, Lisa; Matsui, Mary S; Kluz, Thomas; Costa, Max

    2015-05-01

    Molecular signalling pathways delineating the induction of matrix metalloproteinases (MMPs) by ultraviolet radiation (UVR) are currently well-defined; however, the effects of UVR on epigenetic mechanisms of MMP induction are not as well understood. In this study, we examined solar-simulated UVR (ssUVR)-induced gene expression changes and alterations to histone methylation in the promoters of MMP1 and MMP3 in primary human dermal fibroblasts (HDF). Gene expression changes, including the increased expression of MMP1 and MMP3, were observed using Affymetrix GeneChip arrays and confirmed by qRT-PCR. Using ChIP-PCR, we showed for the first time that in HDF irradiated with 12 J/cm(2) ssUVR, the H3K4me3 transcriptional activating mark increased and the H3K9me2 transcriptional silencing mark decreased in abundance in promoters, correlating with the observed elevation of MMP1 and MMP3 mRNA levels following ssUVR exposure. Changes in mRNA levels due to a single exposure were transient and decreased 5 days after exposure.

  5. Piloted Simulation Tests of Propulsion Control as Backup to Loss of Primary Flight Controls for a B747-400 Jet Transport

    NASA Technical Reports Server (NTRS)

    Bull, John; Mah, Robert; Hardy, Gordon; Sullivan, Barry; Jones, Jerry; Williams, Diane; Soukup, Paul; Winters, Jose

    1997-01-01

    Partial failures of aircraft primary flight control systems and structural damages to aircraft during flight have led to catastrophic accidents with subsequent loss of lives (e.g. DC-10, B-747, C-5, B-52, and others). Following the DC-10 accident at Sioux City, Iowa in 1989, the National Transportation Safety Board recommended 'Encourage research and development of backup flight control systems for newly certified wide-body airplanes that utilize an alternate source of motive power separate from that source used for the conventional control system.' This report describes the concept of a propulsion controlled aircraft (PCA), discusses pilot controls, displays, and procedures; and presents the results of a PCA piloted simulation test and evaluation of the B747-400 airplane conducted at NASA Ames Research Center in December, 1996. The purpose of the test was to develop and evaluate propulsion control throughout the full flight envelope of the B747-400 including worst case scenarios of engine failures and out of trim moments. Pilot ratings of PCA performance ranged from adequate to satisfactory. PCA performed well in unusual attitude recoveries at 35,000 ft altitude, performed well in fully coupled ILS approaches, performed well in single engine failures, and performed well at aft cg. PCA performance was primarily limited by out-of-trim moments.

  6. Structural heterogeneity and unique distorted hydrogen bonding in primary ammonium nitrate ionic liquids studied by high-energy X-ray diffraction experiments and MD simulations.

    PubMed

    Song, Xuedan; Hamano, Hiroshi; Minofar, Babak; Kanzaki, Ryo; Fujii, Kenta; Kameda, Yasuo; Kohara, Shinji; Watanabe, Masayoshi; Ishiguro, Shin-ichi; Umebayashi, Yasuhiro

    2012-03-08

    Liquid structure and the closest ion-ion interactions in a series of primary alkylammonium nitrate ionic liquids [C(n)Am(+)][NO(3)(-)] (n = 2, 3, and 4) were studied by means of high-energy X-ray diffraction (HEXRD) experiments with the aid of molecular dynamics (MD) simulations. Experimental density and X-ray structure factors are in good accordance with those evaluated with MD simulations. With regard to liquid structure, characteristic peaks appeared in the low Q (Q: a scattering vector) region of X-ray structure factors S(Q)'s for all ionic liquids studied here, and they increased in intensity with a peak position shift toward the lower Q side by increasing the alkyl chain length. Experimentally evaluated S(Q(peak))(r(max)) functions, which represent the S(Q) intensity at a peak position of maximum intensity Q(peak) as a function of distance (actually a integration range r(max)), revealed that characteristic peaks in the low Q region are related to the intermolecular anion-anion correlation decrease in the r range of 10-12 Å. Appearance of the peak in the low Q region is probably related to the exclusion of the correlations among ions of the same sign in this r range by the alkyl chain aggregation. From MD simulations, we found unique and rather distorted NH···O hydrogen bonding between C(n)Am(+) (n = 2, 3, and 4) and NO(3)(-) in these ionic liquids regardless of the alkyl chain length. Subsequent ab initio calculations for both a molecular complex C(2)H(5)NH(2)···HONO(2) and an ion pair C(2)H(5)NH(3)(+)···ONO(2)(-) revealed that such distorted hydrogen bonding is specific in a liquid state of this family of ionic liquids, though the linear orientation is preferred for both the N···HO hydrogen bonding in a molecular complex and the NH···O one in an ion pair. Finally, we propose our interpretation of structural heterogeneity in PILs and also in APILs.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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)

  12. Amenorrhea - primary

    MedlinePlus

    ... of periods - primary Images Primary amenorrhea Normal uterine anatomy (cut section) Absence of menstruation (amenorrhea) References Bulun SE. The physiology and pathology of the female reproductive axis. In: ...

  13. Flow measurement by pulsed-neutron activation techniques at the PKL facility at Erlangen (Germany). [PWR

    SciTech Connect

    Kehler, P.

    1982-03-01

    Flow velocities in the downcomer at the PKL facility (in Erlangen, Germany) were measured by the Pulsed-Neutron Activation (PNA) techniques. This was the first time that a fully automated PNA system, incorporating a dedicated computer for on-line data reduction, was used for flow measurements. A prototype of a portable, pulsed, high-output neutron source, developed by the Sandia National Laboratories for the US Nuclear Regulatory Commission, was also successfully demonstrated during this test. The PNA system was the primary flow-measuring device used at the PKL, covering the whole range of velocities of interest. In this test series, the PKL simulated small-break accidents similar to the one that occurred at TMI. The flow velocities in the downcomer were, therefore, very low, ranging between 0.03 and 0.35 m/sec. Two additional flow-measuring methods were used over a smaller range of velocities. Wherever comparison was possible, the PNA-derived velocity values agreed well with the measurements performed by the two more conventional methods.

  14. 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.

  15. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2

    SciTech Connect

    Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E.; Marshall, R. K.; Mohr, C. L.

    1981-09-01

    A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500°F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050°F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500°F), on deformed rods.

  16. 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

  17. 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

  18. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    SciTech Connect

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria

  19. CFD Simulation of Slug Mixing in VVER-1000 Reactor

    SciTech Connect

    Vyskocil, Ladislav

    2006-07-01

    Recently, the safety analyses of VVER and PWR reactors have dealt with the possibility of reactivity-induced accidents related to the penetration of a water slug with low boron concentration into the reactor core. Loop seals at the reactor coolant pump (RCP) suction are the most likely places for the formation of these slugs. The slug is formed in the loop when there is neither natural nor forced circulation. When the circulation is restored, the slug travels towards the reactor and causes an insertion of positive reactivity in the core. This report deals with a CFD simulation of the most dangerous event - the start-up of the first RCP. Only several seconds are needed for slug to reach the core and the operator has no time for corrective action. Mixing of slug on its way to the core can reduce the danger of core recriticality. The primary objective of this study was to find out whether the FLUENT 6 CFD code is capable of predicting the mixing in the cold leg, downcomer and lower plenum as the slug moves toward the reactor core. Numerical simulations were based on mixing tests performed on 1:5 scale model of VVER-1000 reactor at the Gidropress Design Bureau, Russia. In the physical mixing tests, temperature was substituted for Boron concentration through the use of hot and cold water. The time history of core inlet average temperature was calculated by FLUENT and was found to be in good qualitative agreement with experimental data. This work was carried out as part of the EU project FLOMIX-R, Work Package 4. (author)

  20. Primary Hyperparathyroidism

    MedlinePlus

    ... What is PRIMARY HYPERPARATH YROIDIS M? The body’s parathyroid glands—four pea-sized glands in the neck—produce parathyroid hormone (PTH). Primary hyperparathyroidism (PHPT) is a condition ...

  1. Primary thrombocythemia

    MedlinePlus

    ... as myeloproliferative disorders. Others include: Chronic myelogenous leukemia Polycythemia vera Primary myelofibrosis This disorder is most common ... PA: Elsevier Saunders; 2013:chap 68. Tefferi A. Polycythemia vera, essential thrombocytoemia, and primary myelofibrosis. In: Goldman ...

  2. Primary Aldosteronism

    MedlinePlus

    ... Endocrinology Find an Endocrinologist Value of an Endocrinologist Learn About Clinical Trials Keep Your Body in Balance › Primary Aldosteronism Fact Sheet Primary Aldosteronism March 2012 Download PDFs English Espanol Editors Paul Stewart, MD, FRCP William Young, ...

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

    SciTech Connect

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

    1986-01-01

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

  4. 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

  5. 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

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

    NASA Astrophysics Data System (ADS)

    Walter, Daniel John

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

  7. 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).

  8. Determination for dry layer resistance of sucrose under various primary drying conditions using a novel simulation program for designing pharmaceutical lyophilization cycle.

    PubMed

    Kodama, Tatsuhiro; Sawada, Hiroyuki; Hosomi, Hiroshi; Takeuchi, Masahito; Wakiyama, Naoki; Yonemochi, Etsuo; Terada, Katsuhide

    2013-08-16

    Dry layer resistance, which is the resistance of dried cake against water vapor flow generated from sublimation, is one of the important parameters to predict maximum product temperature and drying time during primary drying in lyophilization. The purpose of this study was to develop the predictive model of dry layer resistance under various primary drying conditions using the dry layer resistance obtained from a preliminary lyophilization run. When the maximum dry layer resistance was modified under the assumption that the chamber pressure is zero, the modified dry layer resistance, which is defined as specific dry layer resistance, correlated well with the sublimation rate. From this correlation, the novel predictive model including the empirical formula of sublimation rate and specific dry layer resistance is proposed. In this model, the dry layer resistance under various conditions of shelf temperature and chamber pressure was successfully predicted based on the relationship of the sublimation rate and specific dry layer resistance of the edge and center vials obtained from the product temperature in one preliminary cycle run. It is expected that this predictive model could be a practical and useful tool to predict product temperature during primary drying.

  9. Influence of walking route choice on primary school children's exposure to air pollution--A proof of concept study using simulation.

    PubMed

    Mölter, Anna; Lindley, Sarah

    2015-10-15

    This study developed a walking network for the Greater Manchester area (UK). The walking network allows routes to be calculated either based on the shortest duration or based on the lowest cumulative nitrogen dioxide (NO2) or particulate matter (PM10) exposure. The aim of this study was to analyse the costs and benefits of faster routes versus lower pollution exposure for walking routes to primary schools. Random samples of primary schools and residential addresses were used to generate 100,000 hypothetical school routes. For 60% (59,992) and 40% (40,460) an alternative low NO2 and PM10 route was found, respectively. The median change in travel time (NO2: 4.5s, PM10: 0.5s) and average route exposure (NO2: -0.40 μg/m(3), PM10: -0.03 μg/m(3)) was small. However, quantile regression analysis indicated that for 50% of routes a 1% increase in travel time was associated with a 1.5% decrease in NO2 and PM10 exposure. The results of this study suggest that the relative decrease in pollution exposure on low pollution routes tends to be greater than the relative increase in route length. This supports the idea that a route planning tool identifying less polluted routes to primary schools could help deliver potential health benefits for children.

  10. 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).

  11. 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.

  12. 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

  13. Primary Hyperparathyroidism

    MedlinePlus

    ... D blood test. This test is recommended because vitamin D deficiency is common in people with primary hyperparathyroidism. How ... bone density measurements every 1 to 2 years. Vitamin D deficiency should be corrected if present. Patients who are ...

  14. Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)

    SciTech Connect

    Xu, H.; Fyfitch, S.; Scott, P.; Foucault, M.; Kilian, R.; Winters, M.

    2004-03-01

    Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered.

  15. The effect of abductor muscle and anterior-posterior hip contact load simulation on the in-vitro primary stability of a cementless hip stem

    PubMed Central

    2010-01-01

    Background In-vitro mechanical tests are commonly performed to assess pre-clinically the effect of implant design on the stability of hip endoprostheses. There is no standard protocol for these tests, and the forces applied vary between studies. This study examines the effect of the abductor force with and without application of the anterior-posterior hip contact force in the in-vitro assessment of cementless hip implant stability. Methods Cementless stems (VerSys Fiber Metal) were implanted in twelve composite femurs which were divided into two groups: group 1 (N = 6) was loaded with the hip contact force only, whereas group 2 (N = 6) was additionally subjected to an abductor force. Both groups were subjected to the same cranial-caudal hip contact force component, 2.3 times body weight (BW) and each specimen was subjected to three levels of anterior-posterior hip contact load: 0, -0.1 to 0.3 BW (walking), and -0.1 to 0.6 BW (stair climbing). The implant migration and micromotion relative to the femur was measured using a custom-built system comprised of 6 LVDT sensors. Results Substantially higher implant motion was observed when the anterior-posterior force was 0.6BW compared to the lower anterior-posterior load levels, particularly distally and in retroversion. The abductor load had little effect on implant motion when simulating walking, but resulted in significantly less motion than the hip contact force alone when simulating stair climbing. Conclusions The anterior-posterior component of the hip contact load has a significant effect on the axial motion of the stem relative to the bone. Inclusion of the abductor force had a stabilizing effect on the implant motion when simulating stair climbing. PMID:20576151

  16. A Simulation of the Importance of Length of Growing Season and Canopy Functional Properties on the Seasonal Gross Primary Production of Temperate Alpine Meadows

    PubMed Central

    Baptist, Florence; Choler, Philippe

    2008-01-01

    Background and Aims Along snowmelt gradients, the canopies of temperate alpine meadows differ strongly in their structural and biochemical properties. Here, a study is made of the effects of these canopy dissimilarities combined with the snow-induced changes in length of growing season on seasonal gross primary production (GPP). Methods Leaf area index (LAI) and community-aggregated values of leaf angle and leaf nitrogen content were estimated for seven alpine plant canopies distributed along a marked snowmelt gradient, and these were used as input variables in a sun–shade canopy bulk-photosynthesis model. The model was validated for plant communities of early and late snowmelt sites by measuring the instantaneous CO2 fluxes with a canopy closed-chamber technique. A sensitivity analysis was conducted to estimate the relative impact of canopy properties and environmental factors on the daily and seasonal GPP. Key Results Carbon uptake was primarily related to the LAI and total canopy nitrogen content, but not to the leaf angle. For a given level of photosynthetically active radiation, CO2 assimilation was higher under overcast conditions. Sensitivity analysis revealed that increase of the length of the growing season had a higher effect on the seasonal GPP than a similar increase of any other factor. It was also found that the observed greater nitrogen content and larger LAI of canopies in late-snowmelt sites largely compensated for the negative impact of the reduced growing season. Conclusions The results emphasize the primary importance of snow-induced changes in length of growing season on carbon uptake in alpine temperate meadows. It was also demonstrated how using leaf-trait values of the dominants is a useful approach for modelling ecosystem carbon-cycle-related processes, particularly when continuous measurements of CO2 fluxes are technically difficult. The study thus represents an important step in addressing the challenge of using a plant functional

  17. 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)

  18. 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.

  19. 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)

  20. 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.

  1. Primary hyperparathyroidism

    PubMed Central

    Madkhali, Tarıq; Alhefdhi, Amal; Chen, Herbert; Elfenbein, Dawn

    2016-01-01

    Primary hyperparathyroidism is a common endocrine disorder caused by overactivation of parathyroid glands resulting in excessive release of parathyroid hormone. The resultant hypercalcemia leads to a myriad of symptoms. Primary hyperparathyroidism may increase a patient’s morbidity and even mortality if left untreated. During the last few decades, disease presentation has shifted from the classic presentation of severe bone and kidney manifestations to most patients now being diagnosed on routine labs. Although surgery is the only curative therapy, many advances have been made over the past decades in the diagnosis and the surgical management of primary hyperparathyroidism. The aim of this review is to summarize the characteristics of the disease, the work up, and the treatment options. PMID:26985167

  2. Los Alamos PWR decay-heat-removal studies. Summary results and conclusions

    SciTech Connect

    Boyack, B E; Henninger, R J; Horley, E; Lime, J F; Nassersharif, B; Smith, R

    1986-03-01

    The adequacy of shutdown-decay-heat removal in pressurized water reactors (PWRs) is currently under investigation by the Nuclear Regulatory Commission. One part of this effort is the review of feed-and-bleed procedures that could be used if the normal cooling mode through the steam generators were unavailable. Feed-and-bleed cooling is effected by manually activating the high-pressure injection (HPI) system and opening the power-operated relief valves (PORVs) to release the core decay energy. The feasibility of the feed-and-bleed concept as a diverse mode of heat removal has been evaluated at the Los Alamos National Laboratory. The TRAC-PF1 code has been used to predict the expected performance of the Oconee-1 and Calvert Cliffs-1 reactors of Bobcock and Wilcox and Combustion Engineering, respectively, and the Zion-1 and H.B. Robinson-2 plants of Westinghouse. Feed and bleed was successfully applied in each of the four plants studied, provided it was initiated no later than the time of loss of secondary heat sink. Feed and bleed was successfully applied in two of the plants, Oconee-1 and Zion-1, provided it was initiated no later than the time of primary system saturation. Feed and bleed in Calvert Cliffs-1 when initiated at the time of primary system saturation did result in core dryout; however, the core heatup was eventually terminated by coolant injection. Feed-and-bleed initiation at primary system saturation was not studied for H.B. Robinson-2. Insights developed during the analyses of specific plant transients have been identified and documented. 33 refs., 107 figs., 26 tabs.

  3. IM3D: A parallel Monte Carlo code for efficient simulations of primary radiation displacements and damage in 3D geometry

    PubMed Central

    Li, Yong Gang; Yang, Yang; Short, Michael P.; Ding, Ze Jun; Zeng, Zhi; Li, Ju

    2015-01-01

    SRIM-like codes have limitations in describing general 3D geometries, for modeling radiation displacements and damage in nanostructured materials. A universal, computationally efficient and massively parallel 3D Monte Carlo code, IM3D, has been developed with excellent parallel scaling performance. IM3D is based on fast indexing of scattering integrals and the SRIM stopping power database, and allows the user a choice of Constructive Solid Geometry (CSG) or Finite Element Triangle Mesh (FETM) method for constructing 3D shapes and microstructures. For 2D films and multilayers, IM3D perfectly reproduces SRIM results, and can be ∼102 times faster in serial execution and > 104 times faster using parallel computation. For 3D problems, it provides a fast approach for analyzing the spatial distributions of primary displacements and defect generation under ion irradiation. Herein we also provide a detailed discussion of our open-source collision cascade physics engine, revealing the true meaning and limitations of the “Quick Kinchin-Pease” and “Full Cascades” options. The issues of femtosecond to picosecond timescales in defining displacement versus damage, the limitation of the displacements per atom (DPA) unit in quantifying radiation damage (such as inadequacy in quantifying degree of chemical mixing), are discussed. PMID:26658477

  4. [Primary hyperoxaluria].

    PubMed

    Cochat, Pierre; Fargue, Sonia; Bacchetta, Justine; Bertholet-Thomas, Aurélia; Sabot, Jean-François; Harambat, Jérôme

    2011-07-01

    Primary hyperoxalurias are rare recessive inherited inborn errors of glyoxylate metabolism. They are responsible for progressive renal involvement, which further lead to systemic oxalate deposition, which can even occur in infants. Primary hyperoxaluria type 1 is the most common form in Europe and is due to alanine-glyoxylate aminostransferase deficiency, a hepatic peroxisomal pyridoxin-dependent enzyme. Therefore primary hyperoxaluria type 1 is responsible for hyperoxaluria leading to aggressive stone formation and nephrocalcinosis. As glomerular filtration rate decreases, systemic oxalate storage occurs throughout all the body, and mainly in the skeleton. The diagnosis is first based on urine oxalate measurement, then on genotyping, which may also allow prenatal diagnosis to be proposed. Conservative measures - including hydration, crystallization inhibitors and pyridoxine - are safe and may allow long lasting renal survival, provided it is given as soon as the diagnosis has been even suspected. No dialysis procedure can remove enough oxalate to compensate oxalate overproduction from the sick liver, therefore a combined liver and kidney transplantation should be planned before advanced renal disease has occurred, in order to limit/avoid systemic oxalate deposition. In the future, primary hyperoxaluria type 1 may benefit from hepatocyte transplantation, chaperone molecules, etc.

  5. Evaluation on Influence of Unstable Primary-Energy Price in a Deregulated Electric Power Market—Analysis based on a simulation model approach—

    NASA Astrophysics Data System (ADS)

    Maitani, Tatsuyuki; Tezuka, Tetsuo

    The electric power market of Japan has been locally monopolized for a long time. But, like many countries, Japan is moving forward with the deregulation of its electric power industry so that any power generation company could sell electric power in the market. The power price, however, will fluctuate inevitably to balance the power supply and demand. A new appropriate market design is indispensable when introducing new market mechanisms in the electric power market to avoid undesirable results of the market. The first stage of deregulation will be the competition between an existing large-scaled power utility and a new power generation company. In this paper we have investigated the wholesale market with competition of these two power companies based on a simulation model approach. Under the competitive situation the effects of exogenous disturbance may bring serious results and we estimated the influence on the market when the price of fossil fuel rises. The conclusion of this study is that several types of Nash equilibriums have been found in the market: the larger the new power generation company becomes, the higher the electricity price under the Nash equilibriums rises. Because of the difference in their structure of generation capacity, the existing large-scaled power utility gets more profit while the new power generation company loses its profit when the price of fossil fuel rises.

  6. 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.

  7. Properties of colloidal corrosion products and their effects on nuclear plants. Final report. [PWR; BWR

    SciTech Connect

    Matijevic, E.

    1982-09-01

    Detailed results from the first two years of work on the properties of corrosion product oxides common to light water nuclear reactor systems are presented. A smaller companion volume describes the results in overview fashion. Numerous methods are described for producing these model oxides in forms making their study simpler, i.e., particles with uniform diameter and composition. A number of studies of particle adhesion to simulated power plant surfaces are described. The magnetic properties of hematite of various particle sizes are described - a property important to the use of electromagnetic filtration in LWRs.

  8. Primary hyperparathyroidism.

    PubMed

    Govett, G; White, J

    1989-07-01

    Primary hyperparathyroidism is a pathological entity due to excessive secretion of parathormone from a single or multiple parathyroid glands. The biochemical hallmark of this disorder is an elevated serum calcium. The relationship of the parathyroid glands with the thymus gland in fetal development accounts for the occasional aberrant location of the parathyroids. By utilizing computed tomography or nuclear scanning or both preoperatively, the surgeon can isolate the hyperfunctioning adenoma and resect it, thus minimizing potential complications.

  9. Computer-assisted design of a PWR (pressurized-water reactor) digital control system: Final report

    SciTech Connect

    Shah, S.C.; Chen, D.P.

    1988-01-01

    Modern control design methods and Computer-Aided-Control-Design tools can provide improved control system performance in power plants. Conventional PID controller designs rely on simplistic models of the process to be controlled. This requires a fair amount of ad-hoc on-line tuning for good performance. Modern controller designs based on validated nonlinear simulation models can provide higher performance control and yet require fewer design iterations than PID designs. Low power steam generator level control displays the so-called shrink-swell phenomenon. This non-minimum phase response is difficult to model and control using classical PID tuning methods. A library of nuclear power plant component models is created using MATRIX/sub x/, a Computer-Aided-Control Design tool. Particular emphasis is placed on the low power steam generator level response. The modern controller performance is evaluated against a one element PID controller for low power level control. The modern controller is a digital hierarchical controller with a lower level feedwater flow controller. Global linearization of feedwater valve, a full state estimator, and a state feedback level controller results in superior control over a wider range of parameter variations. Nonlinear simulations show an improvement of four to six fold in the speed of response over a PID controller tuned with classical methods for the non-minimum phase response of the the steam generator level.

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

    SciTech Connect

    Morris, Edgar

    2014-10-01

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

  11. 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.

  12. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-06-01

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

  14. 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.

  15. On the effect of accident conditions on the molten core debris relocation into lower head of a PWR vessel

    NASA Astrophysics Data System (ADS)

    An, Xuegao

    From 1975 to present, it has been found that the primary risk to the public health and safety from nuclear power reactors lies in ``beyond design basis'' accidents. During such severe accidents, melting of the reactor core may lead to a loss of primary system integrity, or even containment failure, which will allow escape of significant amounts of radioactive material to the environment. It is very important to understand the mechanism of reactor core degradation during a severe accident. In this study, the damage progression of the reactor core and the slumping mechanism of molten material to the lower head of the reactor vessel were examined through simulation of severe accident scenarios that lead to large-scale core damage. The calculations were carried out using the computer code SCDAP/RELAP5. Different modeling parameters or models were used in calculations by version MOD3.2. The cladding oxidation shell ``durability'' parameter, which can control the timing of fuel clad failure, was varied. The heat flux model of steady-state natural convection of the molten pool was changed. The ultimate strength of the crust supporting the molten pool was doubled. These changes were made to examine the effects on the calculated core damage, and the molten pool expansion and its slumping. Different accident scenarios were simulated. The HPI/makeup flow rates were changed. The timing of opening and closing the PORV was considered. Reflood by restart of coolant pump 2B was also studied. Finally, the size of the PORV opening was also changed. The effects of these accident scenarios on accident progression and core damage process were studied. From the calculated results, it was concluded that the accurate modeling of core damage phenomena was very important to the prediction of the later stage of an accident. According to code MOD3.2, the molten material in a pool slumped to the lower head of the reactor vessel when the juncture of the top and side crusts failed after the

  16. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    SciTech Connect

    Shah, Vikram Naginbhai; Ware, Arthur Gates; Atwood, Corwin Lee; Sattison, Martin Blaine; Hartley, Robert Scott; Hsu, C.

    1999-08-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number of rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  17. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    SciTech Connect

    A. G. Ware; C. Hsu; C. L. Atwood; M. B. Sattison; R. S. Hartley; V. N. Shah

    1999-02-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number and rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  18. Carbon dioxide flux and net primary production of a boreal treed bog: Responses to warming and water-table-lowering simulations of climate change

    NASA Astrophysics Data System (ADS)

    Munir, T. M.; Perkins, M.; Kaing, E.; Strack, M.

    2015-02-01

    Midlatitude treed bogs represent significant carbon (C) stocks and are highly sensitive to global climate change. In a dry continental treed bog, we compared three sites: control, recent (1-3 years; experimental) and older drained (10-13 years), with water levels at 38, 74 and 120 cm below the surface, respectively. At each site we measured carbon dioxide (CO2) fluxes and estimated tree root respiration (Rr; across hummock-hollow microtopography of the forest floor) and net primary production (NPP) of trees during the growing seasons (May to October) of 2011-2013. The CO2-C balance was calculated by adding the net CO2 exchange of the forest floor (NEff-Rr) to the NPP of the trees. From cooler and wetter 2011 to the driest and the warmest 2013, the control site was a CO2-C sink of 92, 70 and 76 g m-2, the experimental site was a CO2-C source of 14, 57 and 135 g m-2, and the drained site was a progressively smaller source of 26, 23 and 13 g CO2-C m-2. The short-term drainage at the experimental site resulted in small changes in vegetation coverage and large net CO2 emissions at the microforms. In contrast, the longer-term drainage and deeper water level at the drained site resulted in the replacement of mosses with vascular plants (shrubs) on the hummocks and lichen in the hollows leading to the highest CO2 uptake at the drained hummocks and significant losses in the hollows. The tree NPP (including above- and below-ground growth and litter fall) in 2011 and 2012 was significantly higher at the drained site (92 and 83 g C m-2) than at the experimental (58 and 55 g C m-2) and control (52 and 46 g C m-2) sites. We also quantified the impact of climatic warming at all water table treatments by equipping additional plots with open-top chambers (OTCs) that caused a passive warming on average of ~ 1 °C and differential air warming of ~ 6 °C at midday full sun over the study years. Warming significantly enhanced shrub growth and the CO2 sink function of the drained

  19. A Statistical Approach to Predict the Failure Enthalpy and Reliability of Irradiated PWR Fuel Rods During Reactivity-Initiated Accidents

    SciTech Connect

    Nam, Cheol; Jeong, Yong-Hwan; Jung, Youn-Ho

    2001-11-15

    During the last decade, the failure behavior of high-burnup fuel rods under a reactivity-initiated accident (RIA) condition has been a serious concern since fuel rod failures at low enthalpy have been observed. This has resulted in the reassessment of existing licensing criteria and failure-mode study. To address the issue, a statistics-based methodology is suggested to predict failure probability of irradiated fuel rods under an RIA. Based on RIA simulation results in the literature, a failure enthalpy correlation for an irradiated fuel rod is constructed as a function of oxide thickness, fuel burnup, and pulse width. Using the failure enthalpy correlation, a new concept of ''equivalent enthalpy'' is introduced to reflect the effects of the three primary factors as well as peak fuel enthalpy into a single damage parameter. Moreover, the failure distribution function with equivalent enthalpy is derived, applying a two-parameter Weibull statistical model. Finally, the sensitivity analysis is carried out to estimate the effects of burnup, corrosion, peak fuel enthalpy, pulse width, and cladding materials used.

  20. Effects of Tube Rupture Modeling and Parameters on Analysis of MSGTR Event Progression in PWR

    SciTech Connect

    Jeong, Ji Hwan; Choi, Ki Yong; Chang, Keun Sun; Kweon, Young Chel

    2002-07-01

    A multiple steam generator tube rupture (MSGTR) event in APR1400 has been investigated using the best estimate thermal hydraulic system code, MARS1.4. The effects of parameters such as the number of ruptured tubes, rupture location, affected steam generator on analysis of the MSGTR event in APR1400 is examined. In particular, tube rupture modeling methods, single tube modeling (STM) and double tube modeling (DTM), are compared. When five tubes are ruptured, the STM predicts the operator response time of 2085 seconds before main steam safety valves (MSSVs) are lifted. The effects of rupture location on the MSSV lift time is not significant in case of STM, but the MSSV lift time for tube-top rupture is found to be 25.3% larger than that for rupture at hog-leg side tube sheet in case of DTM. The MSSV lift time for the cases that both steam generators are affected (4C5x, 4C23x) are found to be larger than that of the single steam generator cases (4A5x, 4B5x) due to a bifurcation of the primary leak flow. The discharge coefficient of Cd is found to affect the MSSV lift time only for smaller value of 0.5. It is found that the most dominant parameter governing the MSSV lift time is the leak flow rate. Whether any modeling method is used, it gives the similar MSSV lift time if the leak flow rate is close, except the case of both steam generators are affected. Therefore, the system performance and the MSSV lift time of the APR1400 are strongly dependent on the break flow model used in the best estimate system code. (authors)

  1. Geochemistry Model Abstraction and Sensitivity Studies for the 21 PWR CSNF Waste Package

    SciTech Connect

    P. Bernot; S. LeStrange; E. Thomas; K. Zarrabi; S. Arthur

    2002-10-29

    The CSNF geochemistry model abstraction, as directed by the TWP (BSC 2002b), was developed to provide regression analysis of EQ6 cases to obtain abstracted values of pH (and in some cases HCO{sub 3}{sup -} concentration) for use in the Configuration Generator Model. The pH of the system is the controlling factor over U mineralization, CSNF degradation rate, and HCO{sub 3}{sup -} concentration in solution. The abstraction encompasses a large variety of combinations for the degradation rates of materials. The ''base case'' used EQ6 simulations looking at differing steel/alloy corrosion rates, drip rates, and percent fuel exposure. Other values such as the pH/HCO{sub 3}{sup -} dependent fuel corrosion rate and the corrosion rate of A516 were kept constant. Relationships were developed for pH as a function of these differing rates to be used in the calculation of total C and subsequently, the fuel rate. An additional refinement to the abstraction was the addition of abstracted pH values for cases where there was limited O{sub 2} for waste package corrosion and a flushing fluid other than J-13, which has been used in all EQ6 calculation up to this point. These abstractions also used EQ6 simulations with varying combinations of corrosion rates of materials to abstract the pH (and HCO{sub 3}{sup -} in the case of the limiting O{sub 2} cases) as a function of WP materials corrosion rates. The goodness of fit for most of the abstracted values was above an R{sup 2} of 0.9. Those below this value occurred during the time at the very beginning of WP corrosion when large variations in the system pH are observed. However, the significance of F-statistic for all the abstractions showed that the variable relationships are significant. For the abstraction, an analysis of the minerals that may form the ''sludge'' in the waste package was also presented. This analysis indicates that a number a different iron and aluminum minerals may form in the waste package other than those

  2. [Primary aldosteronism].

    PubMed

    Amar, Laurence

    2015-06-01

    Primary aldosteronism affects 6% of hypertensive patients. The diagnosis should be suspected in any patient with severe or resistant hypertension or hypertension associated with hypokalemia. The screening test consists on the assessment of the aldosterone to renin ratio. In case of an elevated ratio, the diagnosis of primary aldosteronism is confirmed by either elevated concentrations of basal plasma and/or urinary aldosterone or absence of suppression of aldosterone during dynamic test (including the saline infusion test). CT aims to ensure the absence of adrenal carcinoma and to study the morphology of the adrenals. The unilateral or bilateral type of aldosterone secretion is based on the realization of an adrenal venous sampling. When the hypersecretion is unilateral, the treatment consists of adrenalectomy leading to cure of hypertension in 42% of cases, improvement in 40% of cases. For patient with bilateral disease or who don't want to undergo surgery, treatment is based on spironolactone usually at doses of 25 or 50 mg in combination with other antihypertensives drugs such as diuretics or calcium channel blockers.

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

    SciTech Connect

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

    1983-01-01

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

  4. On-line measurements of RuO{sub 4} during a PWR severe accident

    SciTech Connect

    Reymond-Laruinaz, S.; Doizi, D.; Boudon, V.; Ducros, G.

    2015-07-01

    After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

  5. Intergranular stress-corrosion cracking of austenitic stainless steels in PWR boric-acid storage systems

    SciTech Connect

    Macdonald, D.D.; Cragnolino, G.A.; Olemacher, J.; Chen, T.Y.; Dhawale, S.

    1982-08-01

    A review is presented of the available literature on the intergranular stress corrosion cracking (IGSCC) of austenitic stainless steels at temperatures below 100/sup 0/C, as well as the results of an experimental investigation of the IGSCC of Types 304, 304L, and 316L stainless steels conducted in boric acid environments of the type employed in pressurized nuclear reactors (PWRs) for nuclear shim control. The susceptibility of furnace sensitized Type 304SS to IGSCC was studied using slow strain rate tests as a function of pH, temperature, potential, and concentration of suspected contaminants: chloride, thiosulfate, and tetrathionate. Possible alternate alloys, such as Types 304L and 316L stainless steels, were also tested under those specific conditions that render Type 304SS susceptible to cracking. Corrosion potentials that can be attained in air-saturated boric acid solutions in the presence of the above mentioned species were measured in order to evaluate the propensity towards intergranular cracking under conditions simulating those that prevail in service.

  6. 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)

  7. Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

    SciTech Connect

    Gerard, R.; Malekian, C.; Meessen, O.

    1997-04-01

    The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

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

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

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

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

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

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

  10. 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

  11. 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)

  12. Primary hyperparathyroidism

    PubMed Central

    Bilezikian, John P; Cusano, Natalie E.; Khan, Aliya A.; Liu, Jian-Min; Marcocci, Claudio; Bandeira, Francisco

    2017-01-01

    Primary hyperparathyroidism (PHPT) is a common disorder in which parathyroid hormone (PTH) is excessively secreted from one or more of the four parathyroid glands. A single benign parathyroid adenoma is the cause in most people. However, multiglandular disease is not rare and is typically seen in familial PHPT syndromes. The genetics of PHPT is usually monoclonal when a single gland is involved and polyclonal when multiglandular disease is present. The genes that have been implicated in PHPT include proto-oncogenes and tumour-suppressor genes. Hypercalcaemia is the biochemical hallmark of PHPT. Usually, the concentration of PTH is frankly increased but can remain within the normal range, which is abnormal in the setting of hypercalcaemia. Normocalcaemic PHPT, a variant in which the serum calcium level is persistently normal but PTH levels are increased in the absence of an obvious inciting stimulus, is now recognized. The clinical presentation of PHPT varies from asymptomatic disease (seen in countries where biochemical screening is routine) to classic symptomatic disease in which renal and/or skeletal complications are observed. Management guidelines have recently been revised to help the clinician to decide on the merits of a parathyroidectomy or a non-surgical course. This Primer covers these areas with particular attention to the epidemiology, clinical presentations, genetics, evaluation and guidelines for the management of PHPT. PMID:27194212

  13. Primary Hyperparathyroidism

    PubMed Central

    Bandeira, Leonardo; Bilezikian, John

    2016-01-01

    Over the past several generations, primary hyperparathyroidism (PHTP) has undergone a change in its clinical presentation in many countries from a symptomatic disease to an asymptomatic one. The reasons for this change in clinical presentation are related to the widespread use of biochemical screening tests, to the measurement of PTH more routinely in the evaluation of metabolic bone disease and to the status of vitamin D sufficiency in the population. Along with recognition of a broader clinical spectrum of disease, including a more recently recognized normocalcemic variant, has come an appreciation that the evaluation of classic target organs that can be affected in PHPT, such as the skeleton and the kidneys, require more advanced imaging technology for complete evaluation. It is clear that even in asymptomatic patients, evidence for microstructural disease in the skeleton and calcifications in the kidneys can be demonstrated often. Potential non-classical manifestations of PHPT related to neurocognition and the cardiovascular system continue to be of interest. As a result of these advances, revised guidelines for the management of asymptomatic PHPT have been recently published to help the clinician determine whether surgery is appropriate or whether a more conservative approach is acceptable. PMID:27508075

  14. Behavior of stainless steels in pressurized water reactor primary circuits

    NASA Astrophysics Data System (ADS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-08-01

    Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

  15. Simulation of primary-slag melting behavior in the cohesive zone of a blast furnace, considering the effect of Al{sub 2}O{sub 3}, Fe{sub t}O, and basicity in the sinter ore

    SciTech Connect

    Hino, Mitsutaka; Nagasaka, Tetsuya; Katsumata, Akitoshi; Higuchi, Kenichi; Yamaguchi, Kazuyoshi; Kon-No, Norimitsu

    1999-08-01

    The alumina content in the iron ore imported to Japan is increasing year by year, and some problems in blast furnace operation, due to the use of the high-alumina-containing sinter, have already been reported. In order to clarify the mechanism of the harmful effect of alumina on the blast furnace operation, the behavior of the primary melt, which is formed in the sinter at the cohesive zone of the blast furnace, has been simulated by dripping slag through an iron or oxide funnel. The effects of basicity, Al{sub 2}O{sub 3}, and Fe{sub t}O contents in the five slag systems on the dripping temperature and weight of slag remaining on the funnel have been discussed. It was found that the eutectic melt formed in the sinter would play an important role in the dripping behavior of the slag in the blast furnace through the fine porosity of the reduced iron and ore particles. Al{sub 2}O{sub 3} increased the weight of the slag remaining on the funnel, and its effect became very significant in the acidic and low-Fe{sub t}O-containing slag. It was estimated that the increase of the weight of the slag remaining on the funnel by Al{sub 2}O{sub 3} in the ore could result in a harmful effect on the permeability resistance and an indirect reduction rate of the sinter in the blast furnace.

  16. Primary hyperoxaluria.

    PubMed

    Lorenzo, Víctor; Torres, Armando; Salido, Eduardo

    2014-05-21

    Primary hyperoxaluria (PH) occurs due to an autosomal recessive hereditary disorder of the metabolism of glyoxylate, which causes excessive oxalate production. The most frequent and serious disorder is due to enzyme deficit of alanine-glyoxylate aminotransferase (PH type I) specific to hepatic peroxisome. As oxalate is not metabolised in humans and is excreted through the kidneys, the kidney is the first organ affected, causing recurrent lithiasis, nephrocalcinosis and early renal failure. With advance of renal failure, particularly in patients on haemodialysis (HD), calcium oxalate is massively deposited in tissues, which is known as oxalosis. Diagnosis is based on family history, the presence of urolithiasis and/or nephrocalcinosis, hyperoxaluria, oxalate deposits in tissue forming granulomas, molecular analysis of DNA and enzyme analysis if applicable. High diagnostic suspicion is required; therefore, unfortunately, in many cases it is diagnosed after its recurrence following kidney transplantation. Conservative management of this disease (high liquid intake, pyridoxine and crystallisation inhibitors) needs to be adopted early in order to delay kidney damage. Treatment by dialysis is ineffective in treating excess oxalate. After the kidney transplant, we normally observe a rapid appearance of oxalate deposits in the graft and the results of this technique are discouraging, with very few exceptions. Pre-emptive liver transplantation, or simultaneous liver and kidney transplants when there is already irreversible damage to the kidney, is the treatment of choice to treat the underlying disease and suppress oxalate overproduction. Given its condition as a rare disease and its genetic and clinical heterogeneity, it is not possible to gain evidence through randomised clinical trials. As a result, the recommendations are established by groups of experts based on publications of renowned scientific rigour. In this regard, a group of European experts (OxalEurope) has

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

    SciTech Connect

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

    2000-09-01

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

  18. [Primary lipodystrophies].

    PubMed

    Capeau, J; Magré, J; Lascols, O; Caron, M; Béréziat, V; Vigouroux, C

    2007-02-01

    Primary lipodystrophies represent a heterogeneous group of very rare diseases with a prevalence of less than 1 case for 100.000, inherited or acquired, caracterized by a loss of body fat either generalized or localized (lipoatrophy). In some forms, lipoatrophy is associated with a selective hypertrophy of other fat depots. Clinical signs of insulin resistance are often present: acanthosis nigricans, signs of hyperandrogenism. All lipodystrophies are associated with dysmetabolic alterations with insulin resistance, altered glucose tolerance or diabetes and hypertriglyceridemia leading to a risk of acute pancreatitis. Chronic complications are those resulting from diabetes involving the retina, kidney and nerves, cardiovascular complications and steatotic liver lesions that could result in cirrhosis. Genetic forms of generalized lipodystrophy (or Berardinelli-Seip syndrome) result, in most cases, from recessive mutations in one of two genes: either BSCL2 coding seipin or BSCL1 coding AGPAT2, an acyl-transferase involved in triglyceride synthesis. Acquired generalized lipodystrophy (Lawrence syndrome) is of unknown origin but is sometimes associated with signs of autoimmunity. Partial lipodystrophies can be familial with dominant transmission. Heterozygous mutations have been identified in the LMNA gene encoding nuclear lamin A/C belonging to the nuclear lamina, or in PPARG encoding the adipogenic transcription factor PPARgamma. Some less typical lipodystrophies, associated with signs of premature aging, have been linked to mutations in LMNA or in the ZMPSTE24 gene encoding the protease responsible for the maturation of prelamin A into lamin A. Acquired partial lipodystrophy (Barraquer-Simons syndrome) is characterized by cephalothoracic fat loss. Its aetiology is unknown but mutations in LMNB2, encoding the lamina protein lamin B2, could represent susceptibility factors. Highly active antiretroviral treatments for HIV infection are currently the most frequent cause

  19. Development of a dynamic simulation mode in Serpent 2 Monte Carlo code

    SciTech Connect

    Leppaenen, J.

    2013-07-01

    This paper presents a dynamic neutron transport mode, currently being implemented in the Serpent 2 Monte Carlo code for the purpose of simulating short reactivity transients with temperature feedback. The transport routine is introduced and validated by comparison to MCNP5 calculations. The method is also tested in combination with an internal temperature feedback module, which forms the inner part of a multi-physics coupling scheme in Serpent 2. The demo case for the coupled calculation is a reactivity-initiated accident (RIA) in PWR fuel. (authors)

  20. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    SciTech Connect

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  1. 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.

  2. 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.

  3. 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.

  4. 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)

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

    SciTech Connect

    Ade, Brian J; Gauld, Ian C

    2011-10-01

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

  6. Simulation of Accident Sequences Including Emergency Operating Procedures

    SciTech Connect

    Queral, Cesar; Exposito, Antonio; Hortal, Javier

    2004-07-01

    Operator actions play an important role in accident sequences. However, design analysis (Safety Analysis Report, SAR) seldom includes consideration of operator actions, although they are required by compulsory Emergency Operating Procedures (EOP) to perform some checks and actions from the very beginning of the accident. The basic aim of the project is to develop a procedure validation system which consists of the combination of three elements: a plant transient simulation code TRETA (a C based modular program) developed by the CSN, a computerized procedure system COPMA-III (Java technology based program) developed by the OECD-Halden Reactor Project and adapted for simulation with the contribution of our group and a software interface that provides the communication between COPMA-III and TRETA. The new combined system is going to be applied in a pilot study in order to analyze sequences initiated by secondary side breaks in a Pressurized Water Reactors (PWR) plant. (authors)

  7. 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.

  8. 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.

  9. 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

  10. Primary Progressive Aphasia

    MedlinePlus

    Primary progressive aphasia Overview By Mayo Clinic Staff Primary progressive aphasia (uh-FAY-zhuh) is a rare nervous system (neurological) syndrome ... your ability to communicate. People with primary progressive aphasia can have trouble expressing their thoughts and understanding ...

  11. TANGO Array.. 2. Simulations

    NASA Astrophysics Data System (ADS)

    Bauleo, P.; Bonifazi, C.; Filevich, A.

    2004-01-01

    The angular and energy resolutions of the TANGO Array were obtained using extensive Monte Carlo simulations performed with a double purpose: (1) to determine the appropriate parameters for the array fitting to the desired range of sensitivity (the knee energy region), and (2) to construct a reliable shower database required for reference in the analysis of experimental data. The AIRES code, with the SIBYLL hadronic collision package, was used to simulate Extended Air Showers produced by primary cosmic rays (assuming protons and iron nuclei), with energies ranging from 10 14 to 10 18 eV. These data were fed into a realistic code which simulates the response of the detectors (water Cherenkov detectors), including the electronics, pickup noise, and the signal attenuation in the connecting cables. The trigger stage was considered in the simulations in order to estimate the trigger efficiency of the array and to verify the accuracy of the reconstruction codes. This paper delineates the simulations performed to obtain the expected behavior of the array, and describes the simulated data. The results of these simulations suggest that we can expect an error in the energy of the primary cosmic-ray of ˜60% of the estimated value and that the error in the measurement of the direction of arrival can be estimated as ˜4°. The present simulations also indicate that unambiguous assignments of the primary energy cannot be obtained because of the uncertainty in the nature of the primary cosmic ray.

  12. 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

  13. Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system. [Pressurized Water Reactor

    NASA Technical Reports Server (NTRS)

    Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

    1986-01-01

    The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

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

    SciTech Connect

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

    2006-10-20

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

  15. The effect of hydrazine dosing on high temperature pH{sub T} and redox potentials under PWR environments

    SciTech Connect

    Maekelae, K.; Aaltonen, P.; Buddas, T.

    1995-10-01

    The release and deposit of corrosion products, which play a key role in activity transport, are controlled by the properties of the primary water and oxides present on component surfaces. Some of the VVER 440 type reactors have started to use hydrazine dosing to primary coolant instead of ammonia, because it has been shown to be efficient in reducing activity transport. On the other hand, some other studies have shown that there is no significant difference between new VVER units using hydrazine dosing and the ones operating with standard potassium/ammonia water chemistry. In this paper the results are presented concerning the out-of-core high temperature water chemistry and incore redox potential measurements at Rez research reactor in Czech Republic during hydrazine/ammonia water chemistries.

  16. 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.

  17. Simple interphase drag model for numerical two-fluid modeling of two-phase flow systems. [PWR; BWR

    SciTech Connect

    Chow, H.; Ransom, V.H.

    1984-06-08

    The interphase drag model that has been developed for RELAP5/MOD2 is based on a simple formulation having flow regime maps for both horizontal and vertical flows. The model is based on a conventional semi-empirical formulation that includes the product of drag coefficient, interfacial area, and relative dynamic pressure. The interphase drag model is implemented in the RELAP5/MOD2 light water reactor transient analysis code and has been used to simulate a variety of separate effects experiments to assess the model accuracy. The results from three of these simulations, the General Electric Company small vessel blowdown experiment, Dukler and Smith's counter-current flow experiment, and a Westinghouse Electric Company FLECHT-SEASET forced reflood experiment, are presented and discussed.

  18. Primary periosteal lymphoma--rare and unusual.

    PubMed

    Abdelwahab, Ibrahim Fikry; Hoch, Benjamin; Hermann, George; Bianchi, Stefano; Klein, Michael J; Springfield, Dempsey S

    2007-04-01

    We describe a primary periosteal lymphoma that involved only the periosteum without affecting the adjacent medulla or the regional lymph nodes. No other lymphomatous foci were found in either the distant lymph nodes or viscera. This unusual presentation simulates the imaging appearance of surface lesions of bone, namely benign and malignant tumors, and departs from the typical appearance of primary lymphoma of bone. Therefore, this rare type of lymphoma should be considered in the differential diagnosis of surface bone lesions.

  19. VERA Core Simulator methodology for pressurized water reactor cycle depletion

    DOE PAGES

    Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

    2017-01-12

    This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

  20. Brain tumor - primary - adults

    MedlinePlus

    ... Vestibular schwannoma (acoustic neuroma) - adults; Meningioma - adults; Cancer - brain tumor (adults) ... Primary brain tumors include any tumor that starts in the brain. Primary brain tumors can start from brain cells, ...

  1. Primary renal carcinoid tumor.

    PubMed

    Kanodia, K V; Vanikar, A V; Patel, R D; Suthar, K S; Kute, V B; Modi, P R; Trivedi, H L

    2013-09-01

    Primary renal carcinoid tumor is extremely rare and, therefore, its pathogenesis and prognosis is not well known. We report a primary renal carcinoid in a 26-year-old man treated by radical nephrectomy.

  2. Primary enzyme quantitation

    DOEpatents

    Saunders, G.C.

    1982-03-04

    The disclosure relates to the quantitation of a primary enzyme concentration by utilizing a substrate for the primary enzyme labeled with a second enzyme which is an indicator enzyme. Enzyme catalysis of the substrate occurs and results in release of the indicator enzyme in an amount directly proportional to the amount of primary enzyme present. By quantifying the free indicator enzyme one determines the amount of primary enzyme present.

  3. Investigating Primary Source Literacy

    ERIC Educational Resources Information Center

    Archer, Joanne; Hanlon, Ann M.; Levine, Jennie A.

    2009-01-01

    Primary source research requires students to acquire specialized research skills. This paper presents results from a user study testing the effectiveness of a Web guide designed to convey the concepts behind "primary source literacy". The study also evaluated students' strengths and weaknesses when conducting primary source research. (Contains 3…

  4. Primary Care's Dim Prognosis

    ERIC Educational Resources Information Center

    Alper, Philip R.

    2010-01-01

    Given the chorus of approval for primary care emanating from every party to the health reform debate, one might suppose that the future for primary physicians is bright. Yet this is far from certain. And when one looks to history and recognizes that primary care medicine has failed virtually every conceivable market test in recent years, its…

  5. Multimuons events and primary composition

    NASA Technical Reports Server (NTRS)

    Acharya, B. S.; Capdevielle, J. N.

    1985-01-01

    Nucleon decay detectors at large depths offers now a total area larger than 1000 sq m to registrate muons of energy exceeding 1 TeV. Near complete high energy muon families are detected in those arrays. An extensive 3D Monte-Carlo simulation was conducted in view to understand the spatial distribution of those events and the possible link with elementary act or primary composition. As pion or kaon parents have a very small decay probability at so high energy, multimuon phenomena occurs at high altitude where the atmospheric density is small after the most energetic collisions.

  6. 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.

  7. Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

    SciTech Connect

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Svard, Staffan Jacobsson; Jansson, Peter; Swinhoe, Martyn T.; Tobin, Stephen J.

    2015-03-04

    Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.

  8. The Shuttle Mission Simulator computer generated imagery

    NASA Technical Reports Server (NTRS)

    Henderson, T. H.

    1984-01-01

    Equipment available in the primary training facility for the Space Transportation System (STS) flight crews includes the Fixed Base Simulator, the Motion Base Simulator, the Spacelab Simulator, and the Guidance and Navigation Simulator. The Shuttle Mission Simulator (SMS) consists of the Fixed Base Simulator and the Motion Base Simulator. The SMS utilizes four visual Computer Generated Image (CGI) systems. The Motion Base Simulator has a forward crew station with six-degrees of freedom motion simulation. Operation of the Spacelab Simulator is planned for the spring of 1983. The Guidance and Navigation Simulator went into operation in 1982. Aspects of orbital visual simulation are discussed, taking into account the earth scene, payload simulation, the generation and display of 1079 stars, the simulation of sun glare, and Reaction Control System jet firing plumes. Attention is also given to landing site visual simulation, and night launch and landing simulation.

  9. 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

  10. Gross Primary Productivity

    NASA Technical Reports Server (NTRS)

    2002-01-01

    NASA's new Moderate-resolution Imaging Spectroradiometer (MODIS) allows scientists to gauge our planet's metabolism on an almost daily basis. GPP, gross primary production, is the technical term for plant photosynthesis. This composite image over the continental United States, acquired during the period March 26-April 10, 2000, shows regions where plants were more or less productive-i.e., where they 'inhaled' carbon dioxide and then used the carbon from photosynthesis to build new plant structures. This false-color image provides a map of how much carbon was absorbed out of the atmosphere and fixed within land vegetation. Areas colored blue show where plants used as much as 60 grams of carbon per square meter. Areas colored green and yellow indicate a range of anywhere from 40 to 20 grams of carbon absorbed per square meter. Red pixels show an absorption of less than 10 grams of carbon per square meter and white pixels (often areas covered by snow or masked as urban) show little or no absorption. This is one of a number of new measurements that MODIS provides to help scientists understand how the Earth's landscapes are changing over time. Scientists' goal is use of these GPP measurements to refine computer models to simulate how the land biosphere influences the natural cycles of water, carbon, and energy throughout the Earth system. The GPP will be an integral part of global carbon cycle source and sink analysis, an important aspect of Kyoto Protocol assessments. This image is the first of its kind from the MODIS instrument, which launched in December 1999 aboard the Terra spacecraft. MODIS began acquiring scientific data on February 24, 2000, when it first opened its aperture door. The MODIS instrument and Terra spacecraft are both managed by NASA's Goddard Space Flight Center, Greenbelt, MD. Image courtesy Steven Running, MODIS Land Group Member, University of Montana

  11. Experimental Investigation of Coolant Mixing in the RPV of PWR in the Late Phase of a SBLOCA Event

    SciTech Connect

    Kliem, Soren; Prasser, Horst-Michael; Suehnel, Tobias; Weiss, Frank-Peter; Hansen, Asmus

    2006-07-01

    Partial depletion of the primary circuit of a pressurized water reactor during a postulated small break loss of coolant accident can lead to interruption of one-phase flow natural circulation. In this case, the decay heat is removed from the core in the reflux-condenser mode. In this operation mode, slugs of lower borated water can accumulate in the cold legs. After refilling of the primary circuit, the natural circulation in the two loops not receiving emergency core cooling injection (ECC) re-establishes and the lower borated slugs are shifted towards the reactor pressure vessel (RPV). Entering the core, the lower borated water causes a reactivity insertion. Mixing inside the RPV is an important phenomenon limiting the reactivity insertion and preventing a re-criticality. The mixing of these lower borated slugs with the ambient coolant in the RPV was investigated at the 1:5 scaled coolant mixing test facility ROCOM. Wire mesh sensors based on electrical conductivity measurement are used in ROCOM to measure in detail the spreading of a tracer solution in the facility. The mixing in the downcomer was observed with a sensor which spans a measuring grid of 64 azimuthal and 32 positions over the height. The resulting distribution of the boron concentration at the core inlet was measured with a sensor integrated into the lower core support plate providing one measurement position at the entry into each fuel assembly. The boundary conditions for the mixing experiment were taken from an experiment at the thermal-hydraulic test facility PKL operated by FANP Germany. The slugs, which have a lower density, accumulate in the upper part of the downcomer after shifting into the RPV. The ECC-water injected into the RPV falls almost straight down through the lower borated water and accelerates. On the outer sides of the ECC-streak, lower borated coolant admixes and flows together with the ECC-water downwards. This is the only mechanism of transporting the lower borated water

  12. Balancing outage performance of primary user and secondary user by relay-assisted primary transmission

    NASA Astrophysics Data System (ADS)

    Zhao, Feng; Sun, Xiangqi; Chen, Hongbin; Bie, Rongfang

    2014-12-01

    In this paper, a cooperative transmission protocol for cognitive radio systems is proposed. In this protocol, the primary system comprises a transmitter (PT), a receiver (PR), and a decode-and-forward relay (Relay), while the secondary system comprises a transmitter (ST) and a receiver (SR). Both the ST and the Relay assist the transmissions of the primary users together. The outage probabilities of the primary system and the secondary system are analyzed and verified through simulations. In order to decrease outage probability of the secondary system, power allocation is performed at the ST. However, it will lead to deterioration of outage performance of the primary system. In order to guarantee outage performance of the primary system, a Relay is employed. Compared with two existing protocols, one without cooperation and the other with cooperation of the secondary system only, the proposed protocol is able to better balance outage performances of the primary system and the secondary system.

  13. Feasibility study on the development of a non-invasive liquid-level gauge for nuclear power reactors. [PWR

    SciTech Connect

    Baratta, A.J.; Jester, W.A.; Imel, G.R.; Okyere, E.W.; Foderaro, A.H.; Kenney, E.S.; McMaster, I.B.; Gundy, M.L.

    1983-05-01

    During the TMI-2 accident the source range detector exhibited anomalous behavior. Analysis of the detector output by the authors and others attributed this behavior to the density and level changes that occurred in the reactor pressure vessel during the TMI-2 accident. As a result of this analysis, a pressure vessel level and density gauge was proposed which uses a string of neutron detectors external to the vessel. This project investigates the feasibility of such a system. Experiments were conducted using an experimental apparatus and The Pennsylvania State University's Breazeale Nuclear Reactor to simulate a reactor during a LOCA. In addition, analytical studies were performed to explain these experiments and aid in further understanding the TMI-2 detector reading. An analysis of recent LOFT data was also conducted. Each of these confirms the ability of an external string of neutron detectors to sense and unambiguously measure water level and density variations in a reactor pressure vessel.

  14. A study of the mechanism of primary water stress corrosion cracking of Alloy 600

    SciTech Connect

    Gourgues, A.F.; Andrieu, E.; Scott, P.M.

    1995-12-31

    Two aspects of the mechanism of stress corrosion cracking of Alloy 600 in pressurized water reactors (PWR) primary water have been studied in detail. Results are presented showing that grain boundaries of Alloy 600 are embrittled to a depth of several microns by exposure to primary water in an unstressed condition. It has been established that this embrittlement is not reversible by high temperature degassing and cannot be directly due to hydrogen. The results seem to support the hypothesis that oxygen atom penetration of grain boundaries is possible. However, no evidence of formation of grain boundary gas bubbles or oxides has been found. It is envisaged that this embrittlement process could sequentially act at the tip of a growing stress corrosion crack. The second phenomenon under study has been the plastic deformation behavior of Alloy 600 since it is known that cold work and stress have an important effect on stress corrosion cracking sensitivity. Results of plastic deformation during cyclic straining at various controlled strain rates are presented showing that Alloy 600 is not very sensitive to loading history and that cold work is of an essentially kinematic nature.

  15. Thermally activated dislocation creep model for primary water stress corrosion cracking of NiCrFe alloys

    SciTech Connect

    Hall, M.M., Jr

    1995-12-31

    There is a growing awareness that awareness that environmentally assisted creep plays an important role in integranular stress corrosion cracking (IGSCC) of NiCrFe alloys in the primary coolant water environment of a pressurized water reactor (PWR). The expected creep mechanism is the thermally activated glide of dislocations. This mode of deformation is favored by the relatively low temperature of PWR operation combined with the large residual stresses that are most often identified as responsible for the SCC failure of plant components. Stress corrosion crack growth rate (CGR) equations that properly reflect the influence of this mechanism of crack tip deformation are required for accurate component life predictions. A phenomenological IGSCC-CGR model, which is based on an apriori assumption that the IGSCC-CGR is controlled by a low temperature dislocation creep mechanism, is developed in this report. Obstacles to dislocation creep include solute atoms such as carbon, which increase the lattice friction force, and forest dislocations, which can be introduced by cold prestrain. Dislocation creep also may be environmentally assisted due to hydrogen absorption at the crack tip. The IGSCC-CGR model developed here is based on an assumption that crack growth occurs by repeated fracture events occurring within an advancing crack-tip creep-fracture zone. Thermal activation parameters for stress corrosion cracking are obtained by fitting the CGR model to IGSCC-CGR data obtained on NiCrFe alloys, Alloy X-750 and Alloy 600. These IGSCC-CGR activation parameters are compared to activation parameters obtained from creep and stress relaxation tests. Recently reported CGR data, which exhibit an activation energy that depends on yield stress and the applied stress intensity factor, are used to benchmark the model. Finally, the effects of matrix carbon concentration, grain boundary carbides and absorbed hydrogen concentration are discussed within context of the model.

  16. Primary Intraosseous Meningioma.

    PubMed

    Chen, Thomas C

    2016-04-01

    Primary intraosseous meningiomas are a subtype of primary extradural meningiomas. They represent approximately two-thirds of extradural meningiomas and fewer than 2% of meningiomas overall. These tumors originate within the bones of the skull and can have a clinical presentation and radiographic differential diagnosis different from those for intradural meningiomas. Primary intraosseous meningiomas are classified based on location and histopathologic characteristics. Treatment is primarily surgical resection with wide margins if possible. Sparse literature exists regarding the use of adjuvant therapies. The literature regarding primary intraosseous meningiomas consists primarily of clinical case reports and case series. This literature is reviewed and summarized in this article.

  17. Thermal-hydraulic code qualification: ATHOS2 and data from Bugey 4 and Tricastin 1. Final report. [PWR

    SciTech Connect

    Masiello, P.J.

    1983-02-01

    Measured data from steam generators at the Bugey 4 and Tricastin 1 nuclear power plants operated by Electricite de France (EdF) have been used in the qualification of the ATHOS2 computer code. ATHOS2 is a three-dimensional, two-phase thermal-hydraulic code for the steady-state and transient analysis of recirculating-type steam generators. Predicted data for circulation ratio and secondary fluid temperature just above the tube sheet have been compared with corresponding data measured by EdF during on-site testing of Westinghouse Model 51A (Bugey 4) and 51M (Tricastin 1) steam generators. Comparative analyses have been performed for steady-state operating conditions at five power levels for each plant installation. The transient capabilities of the ATHOS2 code were examined in the simulation of an open-grid (load reject from 100% power) test conducted at Bugey 4. Results show that predicted data for secondary fluid temperature at eight locations just above the tube sheet are typically within 1.5/sup 0/C of measured data.

  18. EMR Curriculum Guide: Primary.

    ERIC Educational Resources Information Center

    Ruschmeier, Veronica M., Ed.; Rockwell, Linda, Ed.

    Presented is a curriculum guide for educable mentally retarded children in primary and intermediate grades which specifies behavioral and interim objectives in the areas of basic verbal and arithmetic skills, vocational competencies, social competencies, and physical skills. Objectives such as the following are identified at the primary level:…

  19. Using Primary Source Documents.

    ERIC Educational Resources Information Center

    Mintz, Steven

    2003-01-01

    Explores the use of primary sources when teaching about U.S. slavery. Includes primary sources from the Gilder Lehrman Documents Collection (New York Historical Society) to teach about the role of slaves in the Revolutionary War, such as a proclamation from Lord Dunmore offering freedom to slaves who joined his army. (CMK)

  20. Medics in Primary School

    ERIC Educational Resources Information Center

    Press, Colin

    2003-01-01

    Some time ago a flyer on "Medics in Primary School" came the author's way. It described a programme for making placements in primary schools available to medical students. The benefits of the program to medical students and participating schools were highlighted, including opportunities to develop communication skills and demystify…

  1. Transforming Primary Mathematics

    ERIC Educational Resources Information Center

    Askew, Mike

    2011-01-01

    What is good mathematics teaching? What is mathematics teaching good for? Who is mathematics teaching for? These are just some of the questions addressed in "Transforming Primary Mathematics", a highly timely new resource for teachers which accessibly sets out the key theories and latest research in primary maths today. Under-pinned by findings…

  2. [The primary healthcare centres].

    PubMed

    Brambilla, Antonio; Maciocco, Gavino

    2014-04-01

    The central attributes of primary care are: first contact (accessibility), longitudinality (person- focused preventive and curative care overtime), patient-oriented comprehensiveness and coordination (including navigation towards secondary and tertiary care). Besides taking care of the needs of the individuals, primary health care teams are also looking at the community, especially when addressing social determinants of health. The rationale for the benefits for primary care for health has been found in: 1) greater access to needed services; 2) better quality of care; 3) a greater focus on prevention; 4) early management of health problems; 5) organizing and delivering high quality care for chronic non-communicable diseases. This paper describes the role of primary healthcare centres in strengthening community primary services and in reducing health inequalities. Furthemore, the experiences of Regional Health Services from Tuscany and Emilia-Romagna are discussed, with a brief overview of the literature.

  3. Primary Intraocular Lymphoma

    PubMed Central

    Faia, Lisa J.; Chan, Chi-Chao

    2009-01-01

    Primary intraocular lymphoma, recently suggested to be renamed primary retinal lymphoma, is a subset of primary central nervous system lymphoma and is usually an aggressive diffuse large B-cell lymphoma. Between 56% and 85% of patients who initially present with primary intraocular lymphoma alone will develop cerebral lesions. Patients typically complain of decreased vision and floaters, most likely secondary to the chronic vitritis and subretinal lesions. The diagnosis of primary intraocular lymphoma can be difficult to make and requires tissue for diagnosis. The atypical lymphoid cells are large and display a high nuclear to cytoplasmic ratio, prominent nucleoli, and basophilic cytoplasm. Flow cytometry, immunohistochemistry, cytokine analysis, and gene rearrangements also aid in the diagnosis. Local and systemic treatments, such as chemotherapy and radiation, are employed, although the relapse rate remains high. PMID:19653715

  4. A manifestation of the Ostwald step rule: Molecular-dynamics simulations and free-energy landscape of the primary nucleation and melting of single-molecule polyethylene in dilute solution

    NASA Astrophysics Data System (ADS)

    Larini, L.; Leporini, D.

    2005-10-01

    The paper presents numerical results from extensive molecular-dynamics simulations of the crystallization process of a single polyethylene chain with N =500 monomers. The development of the ordered structure is seen to proceed along different routes involving either the global reorganization of the chain or, alternatively, well-separated connected nuclei. No dependence on the thermal history was observed at the late stages of the crystallization. The folding process involves several intermediate ordered metastable states, in strong analogy with the experiments, and ends up in a well-defined long-lived lamella with ten stems of approximately equal length, arranged into a regular, hexagonal pattern. This behavior may be seen as a microscopic manifestation of the Ostwald step rule. Both the metastable states and the long-lived one are evidenced as the local minima and the global one of the free-energy landscape, respectively. The study of the microscopic organization of the lamella evidenced that the two caps are rather flat, i.e., the loops connecting the stems are short. Interestingly, annealing the chain through the different metastable states leaves the average number of monomers per loop nearly unchanged. It is also seen that the chain ends, the so-called cilia, are localized on the surface of the lamella, in agreement with the experiments, and that structural fluctuations take place on the lamella surface, as noted by recent Monte Carlo simulations. The study of the melting process evidences that the degree of hysteresis is small.

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

    SciTech Connect

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

    2001-06-21

    effect. Kim et al. (Reference (p)) and Kim and Kim (Reference (q)) more recently investigated the influence that an applied hoop stress has on the corrosion resistance of Zircaloy tubes in a 400 C steam and in a 350 C concentrated lithia water environment. Both of these studies found the applied tensile hoop stress to have no effect on cladding corrosion rates in the 400 C steam environment but to have accelerated corrosion in the lithiated water environment. In both cases, the corrosion acceleration in the lithiated water environment was attributed to the accumulation of the increased hydrogen picked up in the lithiated environment into the tensile regions of the test specimen. Dense hydride rims have been shown, independent of clad strain, to accelerate the corrosion of Zirconium alloys (References (r) and (s)), suggesting that the primary effect of applied stresses on the corrosion of Zircaloy in the above studies is through the accumulation of hydrogen at the oxide-to-metal interface and not through a direct mechanical breakdown of the passivating boundary layer. To further investigate the potential role of in-situ clad straining (or stress) on Zircaloy corrosion rates, two experimental studies were performed. First, several samples that were irradiated with and without an applied stress were destructively examined for the extent of corrosion occurring in strained and nonstrained regions of the test samples. The extent of corrosion was determined, posttest, by metallographic examination. Second, the corrosion process was monitored in-situ using electrochemical impedance spectroscopy on samples exposed out-of-reactor with and without an applied stress. Post test, these autoclave samples were also metallographically examined.

  6. Aircraft Simulators and Pilot Training.

    ERIC Educational Resources Information Center

    Caro, Paul W.

    Flight simulators are built as realistically as possible, presumably to enhance their training value. Yet, their training value is determined by the way they are used. Traditionally, simulators have been less important for training than have aircraft, but they are currently emerging as primary pilot training vehicles. This new emphasis is an…

  7. 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.

  8. Visualizing renal primary cilia.

    PubMed

    Deane, James A; Verghese, Elizabeth; Martelotto, Luciano G; Cain, Jason E; Galtseva, Alya; Rosenblum, Norman D; Watkins, D Neil; Ricardo, Sharon D

    2013-03-01

    Renal primary cilia are microscopic sensory organelles found on the apical surface of epithelial cells of the nephron and collecting duct. They are based upon a microtubular cytoskeleton, bounded by a specialized membrane, and contain an array of proteins that facilitate their assembly, maintenance and function. Cilium-based signalling is important for the control of epithelial differentiation and has been implicated in the pathogenesis of various cystic kidney diseases and in renal repair. As such, visualizing renal primary cilia and understanding their composition has become an essential component of many studies of inherited kidney disease and mechanisms of epithelial regeneration. Primary cilia were initially identified in the kidney using electron microscopy and this remains a useful technique for the high resolution examination of these organelles. New reagents and techniques now also allow the structure and composition of primary cilia to be analysed in detail using fluorescence microscopy. Primary cilia can be imaged in situ in sections of kidney, and many renal-derived cell lines produce primary cilia in culture providing a simplified and accessible system in which to investigate these organelles. Here we outline microscopy-based techniques commonly used for studying renal primary cilia.

  9. Lunar Regolith Simulant User's Guide

    NASA Technical Reports Server (NTRS)

    Schrader, C. M.; Rickman, D. L.; McLemore, C. A.; Fikes, J. C.

    2010-01-01

    Based on primary characteristics, currently or recently available lunar regolith simulants are discussed from the perspective of potential experimental uses. The characteristics used are inherent properties of the material rather than their responses to behavioral (geomechanical, physiochemical, etc.) tests. We define these inherent or primary properties to be particle composition, particle size distribution, particle shape distribution, and bulk density. Comparable information about lunar materials is also provided. It is strongly emphasized that anyone considering either choosing or using a simulant should contact one of the members of the simulant program listed at the end of this document.

  10. The promise of quantum simulation

    SciTech Connect

    Muller, Richard P.; Blume-Kohout, Robin

    2015-07-21

    In this study, quantum simulations promise to be one of the primary applications of quantum computers, should one be constructed. This article briefly summarizes the history of quantum simulation in light of the recent result of Wang and co-workers, demonstrating calculation of the ground and excited states for a HeH+ molecule, and concludes with a discussion of why this and other recent progress in the field suggest that quantum simulations of quantum chemistry have a bright future.

  11. The promise of quantum simulation

    DOE PAGES

    Muller, Richard P.; Blume-Kohout, Robin

    2015-07-21

    In this study, quantum simulations promise to be one of the primary applications of quantum computers, should one be constructed. This article briefly summarizes the history of quantum simulation in light of the recent result of Wang and co-workers, demonstrating calculation of the ground and excited states for a HeH+ molecule, and concludes with a discussion of why this and other recent progress in the field suggest that quantum simulations of quantum chemistry have a bright future.

  12. TANGO ARRAY II: Simulations

    NASA Astrophysics Data System (ADS)

    Bauleo, P.; Bonifazi, C.; Filevich, A.

    The angular and energy resolution of the TANGO Array has been obtained using Monte Carlo simulations. The AIRES code, with the SYBILL hadronic collision package, was used to simulate Extended Air Showers produced by primary cosmic rays (protons and iron nuclei), with energies ranging from 1014 eV to 1018 eV. These data were fed into a realistic code which simulates the response of the detector stations (water ˇCerenkov detectors), including the electronics, pick up noise, and the signal attenuation in the connecting cabling. The trigger stage is taken into account in order to produce estimates of the trigger efficiency of the array and to check the accuracy of the reconstruction codes. This paper describes the simulations performed to obtain the expected behavior of the array, and presents the simulated data. These simulations indicate that the accuracy of the cosmic ray primary energy determination is expected to be ˜ 60 % and the precision in the measurement of the direction of arrival can be estimated as ˜ 4 degrees.

  13. Primary vascular access.

    PubMed

    Gibbons, C P

    2006-05-01

    Primary vascular access is usually achievable by a distal autogenous arterio-venous fistula (AVF). This article describes the approach to vascular access planning, the usual surgical options and the factors affecting patency.

  14. Primary infertility (image)

    MedlinePlus

    Primary infertility is a term used to describe a couple that has never been able to conceive a pregnancy ... to do so through unprotected intercourse. Causes of infertility include a wide range of physical as well ...

  15. Parenthood after Primary Infertility.

    ERIC Educational Resources Information Center

    Frances-Fischer, Jana E.; Lightsey, Owen Richard, Jr.

    2003-01-01

    Reviews the literature on the experience of parenting after primary infertility and describes construction and initial testing of an instrument for assessing characteristics of this understudied population. (Contains 52 references and 4 tables.) (GCP)

  16. Primary Nurse - Role Evolution

    ERIC Educational Resources Information Center

    Mundinger, Mary O'Neil

    1973-01-01

    Primary nursing means that each patient has an individual nurse who is responsible for assessing his nursing needs and planning and evaluating his nursing care. The article describes the advantages and problems connected with this approach to patient care. (AG)

  17. Inside the Primary Classroom.

    ERIC Educational Resources Information Center

    Simon, Brian

    1980-01-01

    Presents some of the findings of the ORACLE research program (Observational Research and Classroom Learning Evaluation), a detailed observational study of teacher-student interaction, teaching styles, and management methods within a sample of primary classrooms. (Editor/SJL)

  18. Primary aldosteronism and pregnancy.

    PubMed

    Morton, Adam

    2015-10-01

    Primary aldosteronism is the most common cause of secondary hypertension. Less than 50 cases of pregnancy in women with primary aldosteronism have been reported, suggesting the disorder is significantly underdiagnosed in confinement. Accurate diagnosis is complicated by physiological changes in the renin-angiotensin-aldosterone axis in pregnancy, leading to a risk of false negative results on screening tests. The course of primary aldosteronism during pregnancy is highly variable, although overall it is associated with a very high risk of fetal and maternal morbidity and mortality. The optimal management of primary aldosteronism during pregnancy is unclear, with uncertainty regarding the safety of mineralocorticoid antagonists and amiloride, their relative efficacy compared with the antihypertensive medications commonly used during pregnancy, and as to whether prognosis is improved by laparoscopic adrenalectomy where an adrenal adenoma can be demonstrated.

  19. Primary biliary cirrhosis

    MedlinePlus

    ... medlineplus.gov/ency/article/000282.htm Primary biliary cirrhosis To use the sharing features on this page, ... and leads to scarring of the liver called cirrhosis. This is called biliary cirrhosis. Causes The cause ...

  20. Primary actinomycosis of hand

    PubMed Central

    Padhi, Sanghamitra; Dash, Muktikesh; Turuk, Jyotirmayee; Sahu, Rani; Panda, Pritilata

    2014-01-01

    Actinomycosis is a chronic granulomatous suppurative disease having the propensity for extension to the contagious tissue with the formation of multiple discharging sinus tracts. Primary actinomycosis of extremity is a very uncommon clinical entity and is commonly considered as a soft-tissue infection. We report here, a case of primary actinomycosis of the upper extremity in a 24-year-old male who was treated successfully with surgical excision and extended period of antimicrobial treatment. PMID:25538911

  1. Primary headache disorders.

    PubMed

    Benoliel, Rafael; Eliav, Eli

    2013-07-01

    Primary headache disorders include migraine, tension-type headaches, and the trigeminal autonomic cephalgias (TACs). "Primary" refers to a lack of clear underlying causative pathology, trauma, or systemic disease. The TACs include cluster headache, paroxysmal hemicrania, and short-lasting neuralgiform headache attacks with conjunctival injection and tearing; hemicrania continua, although classified separately by the International Headache Society, shares many features of both migraine and the TACs. This article describes the features and treatment of these disorders.

  2. Primary vitreoretinal lymphoma

    PubMed Central

    Mulay, Kaustubh; Narula, Ritesh; Honavar, Santosh G

    2015-01-01

    Primary vitreoretinal lymphoma (PVRL) is an uncommon, but potentially fatal intraocular malignancy, which may occur with or without primary central nervous system lymphoma (PCNSL). Considered to be a subset of PCNSL, it is mostly of diffuse large B-cell type. The diagnosis of PVRL poses a challenge not only to the clinician, but also to the pathologist. Despite aggressive treatment with chemotherapy and/or radiotherapy, relapses or CNS involvement are common. PMID:25971162

  3. Primary care research ethics.

    PubMed Central

    Jones, R; Murphy, E; Crosland, A

    1995-01-01

    Research activity in primary care is increasing rapidly, and raises a range of specific ethical issues. Many of these relate to the involvement of individuals in the community who are not seeking medical care and to the impact of research participation on relationships between general practitioners and their patients. The ethical issues pertinent to a range of quantitative and qualitative research methodologies in primary care are identified and considered. PMID:8554844

  4. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

    DOE PAGES

    Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; ...

    2016-08-25

    We derived a consistent “2D/1D” neutron transport method from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. Our paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. We also performed several applications on both leadership-class and industry-classmore » computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.« less

  5. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

    NASA Astrophysics Data System (ADS)

    Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T. H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew

    2016-12-01

    A consistent "2D/1D" neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

  6. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

    SciTech Connect

    Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T. H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew

    2016-08-25

    We derived a consistent “2D/1D” neutron transport method from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. Our paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. We also performed several applications on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

  7. Primary biliary cirrhosis.

    PubMed

    Carey, Elizabeth J; Ali, Ahmad H; Lindor, Keith D

    2015-10-17

    Primary biliary cirrhosis is a chronic cholestatic liver disease characterised by destruction of small intrahepatic bile ducts, leading to fibrosis and potential cirrhosis through resulting complications. The serological hallmark of primary biliary cirrhosis is the antimitochondrial antibody, a highly disease-specific antibody identified in about 95% of patients with primary biliary cirrhosis. These patients usually have fatigue and pruritus, both of which occur independently of disease severity. The typical course of primary biliary cirrhosis has changed substantially with the introduöction of ursodeoxycholic acid (UDCA). Several randomised placebo-controlled studies have shown that UDCA improves transplant-free survival in primary biliary cirrhosis. However, about 40% of patients do not have a biochemical response to UDCA and would benefit from new therapies. Liver transplantation is a life-saving surgery with excellent outcomes for those with decompensated cirrhosis. Meanwhile, research on nuclear receptor hormones has led to the development of exciting new potential treatments. This Seminar will review the current understanding of the epidemiology, pathogenesis, and natural history of primary biliary cirrhosis, discuss management of the disease and its sequelae, and introduce research on new therapeutic options.

  8. Primary cerebral malignant melanoma

    PubMed Central

    Tang, Kai; Kong, Xiangyi; Mao, Gengsheng; Qiu, Ming; Zhu, Haibo; Zhou, Lei; Nie, Qingbin; Xu, Yi; Du, Shiwei

    2017-01-01

    Abstract Primary intracranial melanomas are uncommon and constitute approximately 1% of all melanoma cases and 0.07% of all brain tumors. In nature, these primary melanomas are very aggressive and can spread to other organs. We report an uncommon case of primary cerebral malignant melanoma—a challenging diagnosis guided by clinical presentations, radiological features, and surgical biopsy results, aiming to emphasize the importance of considering primary melanoma when making differential diagnoses of intracranial lesions. We present a rare case of a primary cerebral melanoma in the left temporal lobe. The mass appeared iso-hypodense on brain computed tomography (CT), short signal on T1-weighted magnetic resonance images (T1WI) and long signal on T2WI. It was not easy to make an accurate diagnosis before surgery. We showed the patient's disease course and reviewed related literatures, for readers’ reference. Written informed consent was obtained from the patient for publication of this case report and any accompanying images. Because of this, there is no need to conduct special ethic review and the ethical approval is not necessary. After surgery, the pathological examination confirmed the diagnosis of melanoma. The patient was discharged without any complications and went on to receive adjuvant radiochemotherapy. It is difficult to diagnose primary cerebral melanoma in the absence of any cutaneous melanosis. A high index of clinical suspicion along with good pathology reporting is the key in diagnosing these extremely rare tumors. PMID:28121927

  9. Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

    SciTech Connect

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2004-07-01

    To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

  10. LOCA simulation in NRU program: data report for the fourth materials experiment (MT-4)

    SciTech Connect

    Wilson, C.L.; Mohr, C.L.; Hesson, G.M.; Wildung, N.J.; Russcher, G.E.; Webb, B.J.; Freshley, M.D.

    1983-07-01

    A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program by Pacific Northwest Laboratory (PNL). This experiment (MT-4) was funded by the US Nuclear Regulatory Commission (NRC) to evaluate ballooning and rupture during adiabatic heatup in the temperature range of 1033 to 1200K (1400 to 1700/sup 0/F). The 12 rest rods in the center of the 32-rod bundle were initially pressurized to 4.62 MPa (670 psia) to insure rupture in the correct temperature range. All 12 test rods ruptured with an average strain of 43.7% at the maximum flow blockage elevation of 2.68 m (105.4 in.). Experimental data for the MT-4 transient experiment and post-test measurements and photographs of the fuel are presented in this report.

  11. Other primary headaches

    PubMed Central

    Bahra, Anish

    2012-01-01

    The ‘Other Primary Headaches’ include eight recognised benign headache disorders. Primary stabbing headache is a generally benign disorder which often co-exists with other primary headache disorders such as migraine and cluster headache. Primary cough headache is headache precipitated by valsalva; secondary cough has been reported particularly in association with posterior fossa pathology. Primary exertional headache can occur with sudden or gradual onset during, or immediately after, exercise. Similarly headache associated with sexual activity can occur with gradual evolution or sudden onset. Secondary headache is more likely with both exertional and sexual headache of sudden onset. Sudden onset headache, with maximum intensity reached within a minute, is termed thunderclap headache. A benign form of thunderclap headache exists. However, isolated primary and secondary thunderclap headache cannot be clinically differentiated. Therefore all headache of thunderclap onset should be investigated. The primary forms of the aforementioned paroxysmal headaches appear to be Indomethacin sensitive disorders. Hypnic headache is a rare disorder which is termed ‘alarm clock headache’, exclusively waking patients from sleep. The disorder can be Indomethacin responsive, but can also respond to Lithium and caffeine. New daily persistent headache is a rare and often intractable headache which starts one day and persists daily thereafter for at least 3 months. The clinical syndrome more often has migrainous features or is otherwise has a chronic tension-type headache phenotype. Management is that of the clinical syndrome. Hemicrania continua straddles the disorders of migraine and the trigeminal autonomic cephalalgias and is not dealt with in this review. PMID:23024566

  12. Primary care development zones.

    PubMed

    Beardshaw, V; Gordon, P; Plamping, D

    1993-01-30

    Most commentators on the Tomlinson report have agreed with its emphasis on improving primary and community care. The three elements of such a strategy are a remedial programme to bring primary care up to national standards, a programme to provide such services to people with non-standard needs such as mobile Londoners, ethnic minorities, and homeless people, and the development of an expanded model of primary care. No one model will be appropriate across all of London. The process should start with an audit of existing resources and services within each community, together with an analysis of needs. From this would develop a local programme with specific plans for investment in premises, staffing, training, and management. New contractual mechanisms may be needed to attract practitioners, improve their premises, secure out of hours services, and provide medical cover for community beds. There should also be incentives for closer working between primary and secondary services. No developments on the scale needed for London have been carried out in primary care within the lifetime of the NHS--but their success will be critical to the calibre of health services for Londoners into the next century.

  13. Primary lymphoma of the brain

    MedlinePlus

    Brain lymphoma; Cerebral lymphoma; Primary lymphoma of the central nervous system; Lymphoma - brain ... The cause of primary brain lymphoma is not known. People with a weakened immune system are at high risk for primary lymphoma of the brain. ...

  14. Primary hepatic carcinoid tumor.

    PubMed

    Gao, Jinbo; Hu, Zhijian; Wu, Junwei; Bai, Lishan; Chai, Xinqun

    2011-11-19

    Primary hepatic carcinoid tumor is rare and poses a challenge for diagnosis and management. We presented a case of primary hepatic carcinoid tumor in a 53-year-old female with a complaint of right upper abdominal pain. Computer tomography scans revealed a hypervascular mass in segment 4 of the liver. An ultrasonography-guided biopsy showed a carcinoid tumor. No other lesions were found by the radiological investigations. Surgery resection was performed and histopathological examination revealed a primary hepatic carcinoid tumor. Three years later, recurrence was found and transcatheter arterial chemoembolization was performed. After transcatheter arterial chemoembolization, the patient has been free of symptom and had no radiological disease progression for over 6 months. Surgical resection combination with transcatheter arterial chemoembolization is effective to offer excellent palliation.

  15. Plume primary smoke

    NASA Astrophysics Data System (ADS)

    Chastenet, J. C.

    1993-06-01

    The exhaust from a solid propellant rocket motor usually contains condensed species. These particles, also called 'Primary Smoke', are often prejudicial to missile detectability and to the guidance system. To avoid operational problems it is necessary to know and quantify the effects of particles on all aspects of missile deployment. A brief description of the origin of the primary smoke is given. It continues with details of the interaction between particles and light as function of both particles and light properties (nature, size, wavelength, etc). The effects of particles on plume visibility, attenuation of an optical beam propagated through the plume and the contribution of particles on optical signatures of the plume are also described. Finally, various methods used in NATO countries to quantify the primary smoke effects are discussed.

  16. SALT segmented primary mirror: inductive edge sensors

    NASA Astrophysics Data System (ADS)

    Gajjar, Hitesh; Menzies, John; Buckley, David; Neel, Christian; Parbaud, Philippe; Royet, Stéphane

    2014-07-01

    The development of an inductive edge sensor is in process for the control of the Southern African Large Telescope's (SALT)1 segmented mirror primary. The original capacitive edge sensing system was not capable of maintaining the figure of the primary mirror due to excessive noise and a severe sensitivity to humidity despite exhaustive attempts at characterisation1. The prototype of the inductive edge sensor has progressed to a mature industrialised version that is in the process of being installed and commissioned on SALT. The performance of the sensor in response to temperature and RH is very good with a maximum error of 10nm typical after temperature compensation. The noise and control characteristics of the array have been simulated in order to establish the maximum cumulative error and error rate tolerable for the SALT specific case. It has been established through simulation that over the expected 5 day alignment cycle, a maximum cumulative error of 30nm can be tolerated.

  17. [Primary care in Ireland].

    PubMed

    Sánchez-Sagrado, T

    2017-03-27

    Spanish doctors are still leaving the country to look for quality work. Ireland is not a country with many Spanish professionals but it is interesting to know its particular Health care system. Ireland is one of the countries with a national health care system, although it has a mixture of private health care insurance schemes. People have a right to health care if they have been living in Ireland at least for a year. Access to the primary care health system depends on age and income: free of charge for Category 1 and co-payments for the rest. This division generates great inequalities among the population. Primary Care doctors are self-employed, and they work independently. However, since 2001 they have tended to work in multidisciplinary teams in order to strengthen the Primary Care practice. Salary is gained from a combination of public and private incomes which are not differentiated. The role of the General Practitioner consists in the treatment of acute and chronic diseases, minor surgery, child care, etc. There is no coordination between Primary and Secondary care. Access to specialised medicine is regulated by the price of consultation. Primary Care doctors are not gatekeepers. To be able to work here, doctors must have three years of training after medical school. After that, Continuing Medical Education is compulsory, and the college of general practitioners monitors it annually. The Irish health care system does not fit into the European model. Lack of a clear separation between public and private health care generates great inequalities. The non-existence of coordination between primary and specialised care leads to inefficiencies, which Ireland cannot allow itself after a decade of economic crisis.

  18. Primary appendiceal mucinous adenocarcinoma.

    PubMed

    Behera, Prativa Kumari; Rath, Pramod Kumar; Panda, Rabiratna; Satpathi, Sanghamitra; Behera, Rajan

    2011-04-01

    Primary Adenocarcinomas of the appendix are extremely rare tumor. We report a case of primary mucinous adenocarcinoma in a 40 year old lady misdiagnosed as having acute appendicitis. All the routine investigations were within normal limit. USG of abdomen showed dilated appendix with little fluid collection adjacent to it and no other abnormality was seen which suggested acute appendicitis. Appendicectomy was done and excised appendix was sent for histopathological examination. Mucinous Adenocarcinoma of the appendix was confirmed after histopathological examination. Right hemicolectomy was done as a second stage procedure. As some cases are incidentally discovered, this case emphasizes that histological examination of all appendicectomy specimens is mandatory.

  19. Does primary fibromyalgia exist?

    PubMed

    Forslind, K; Fredriksson, E; Nived, O

    1990-10-01

    Twenty-one of 25 consecutive primary fibromyalgia or fibrositis patients, identified during a 5-year period in a tertiary care day-ward for pain syndromes, were re-examined. Fifteen fulfilled criteria for fibromyalgia but unexpectedly, all cases had either psychiatric disturbance or thyroid dysfunction. Of the four patients not seen at follow-up, two had developed neurological diseases, another rheumatoid arthritis and one other hypothyroidism. Thus, after 5 years no patient fulfilled the criteria for primary fibromyalgia. Women occupied as manual workers were over-represented. Most patients reported beneficial effects of physiotherapy. None of the patients has been able to return to full time work.

  20. Fibrositis and primary hypothyroidism.

    PubMed

    Carette, S; Lefrançois, L

    1988-09-01

    The prevalence of fibrositis was determined in 100 patients with subclinical or biochemical primary hypothyroidism. Nineteen patients reported symptoms of joint and/or muscle pain with stiffness. Five of these patients presented 7 or more tender points on examination, thus allowing a diagnosis of fibrositis to be made in only 5% of the total group. Symptomatic improvement after thyroid hormone replacement occurred in 10 of the 19 patients, including 3 of those with fibrositis. There were no significant changes in tender points. Our data indicate that fibrositis is uncommon in patients with primary hypothyroidism despite the frequent occurrence of symptoms suggestive of this syndrome.

  1. Melatonin for primary insomnia?

    PubMed

    2009-07-01

    Melatonin, a hormone produced by the pineal gland, has a key role in regulating circadian rhythms, most importantly, the sleep-wake cycle. Melatonin's action has led to its being tried as a treatment for a wide range of sleep disorders, such as jet lag, primary insomnia, sleep-wake cycle disruption and sleep problems in children with neuro-developmental disorders. Until recently, it had not been licensed in the UK for any indication. Prolonged-release melatonin (Circadin - Lundbeck) has now been licensed as a treatment for primary insomnia. Here we consider whether this product has a place in the management of people with this condition.

  2. Primary Creative Writing.

    ERIC Educational Resources Information Center

    Wooten, Vida Jo

    1968-01-01

    Children will enjoy creative writing in the primary grades if they are given inspiration, time to write, and the opportunity to share their work with classmates. A second-grade class began a creative writing project by listening to poetry and selecting poems to memorize and recite. This stimulated and encouraged them to evaluate and to write…

  3. Healthcare is primary.

    PubMed

    Kumar, Raman

    2015-01-01

    India is undergoing a rapid transformation in terms of governance, administrative reforms, newer policy develoment, and social movements. India is also considered one of the most vibrant economies in the world. The current discourse in public space is dominated by issues such as economic development, security, corruption free governance, gender equity, and women safety. Healthcare though remains a pressing need of population; seems to have taken a backseat. In the era of decreasing subsidies and cautious investment in social sectors, the 2(nd) National Conference on Family Medicine and Primary Care 2015 (FMPC) brought a focus on "healthcare" in India. The theme of this conference was "Healthcare is Primary." The conference participants discussed on the theme of why healthcare should be a national priority and why strong primary care should remain at the center of healthcare delivery system. The experts recommended that India needs to strengthen the "general health system" instead of focusing on disease based vertical programs. Public health system should have capacity and skill pool to be able to deliver person centered comprehensive health services to the community. Proactive implementation of policies towards human resource in health is the need of the hour. As the draft National Health Policy 2015 is being debated, "family medicine" (academic primary care), the unfinished agenda of National Health Policy 2002, remains a priority area of implementation.

  4. Beginning Primary Teaching

    ERIC Educational Resources Information Center

    Jacklin, Angela; Griffiths, Vivienne; Robinson, Carol

    2006-01-01

    This book supports primary teachers' early professional development and learning, tackling key questions and concerns that new teachers might face in their early careers, such as: How will I get through the first term? When will I feel like a "real" teacher? What can I expect from my first years in teaching? Drawing on the experiences of beginning…

  5. New Primary School Syllabus.

    ERIC Educational Resources Information Center

    Ministry of Education and Culture (Trinidad and Tobago).

    This official syllabus of Trinidad and Tobago's primary schools gives detailed guidelines on the teaching objectives of each curriculum area and how these can best be realized, as well as descriptions of the subject matter. The curriculum is divided into three levels: Level I (5- to 7-year-olds), Level II (7- to 9-year-olds) and Level III (10+- to…

  6. Primary diffuse leptomeningeal gliosarcomatosis.

    PubMed

    Moon, Ju Hyung; Kim, Se Hoon; Kim, Eui Hyun; Kang, Seok-Gu; Chang, Jong Hee

    2015-04-01

    Primary diffuse leptomeningeal gliomatosis (PDLG) is a rare condition with a fatal outcome, characterized by diffuse infiltration of the leptomeninges by neoplastic glial cells without evidence of primary tumor in the brain or spinal cord parenchyma. In particular, PDLG histologically diagnosed as gliosarcoma is extremely rare, with only 2 cases reported to date. We report a case of primary diffuse leptomeningeal gliosarcomatosis. A 68-year-old man presented with fever, chilling, headache, and a brief episode of mental deterioration. Initial T1-weighted post-contrast brain magnetic resonance imaging (MRI) showed diffuse leptomeningeal enhancement without a definite intraparenchymal lesion. Based on clinical and imaging findings, antiviral treatment was initiated. Despite the treatment, the patient's neurologic symptoms and mental status progressively deteriorated and follow-up MRI showed rapid progression of the disease. A meningeal biopsy revealed gliosarcoma and was conclusive for the diagnosis of primary diffuse leptomeningeal gliosarcomatosis. We suggest the inclusion of PDLG in the potential differential diagnosis of patients who present with nonspecific neurologic symptoms in the presence of leptomeningeal involvement on MRI.

  7. Pediatric primary gastric lymphoma.

    PubMed

    Harris, G J; Laszewski, M J

    1992-04-01

    Primary gastric lymphoma in the pediatric population is rare. We have described a case of non-Hodgkin's lymphoma (Burkitt's type) manifested as a gastric mass. Despite its rarity in children, this tumor should be treated aggressively, since long-term survival has been reported.

  8. Primary Biliary Cirrhosis

    MedlinePlus

    ... of liver cancer every 6 to 12 months. Health care providers use blood tests, ultrasound, or both to check for signs of ... make the diagnosis of primary biliary cirrhosis. A health care provider uses the test selectively when he or she is concerned that ...

  9. Primary Standards Laboratory report

    SciTech Connect

    Not Available

    1990-12-01

    Sandia National Laboratories operates the Primary Standards Laboratory (PSL) for the Department of Energy, Albuquerque Operations Office (DOE/AL). This report summarizes metrology activities that received emphasis in the first half of 1990 and provides information pertinent to the operation of the DOE/AL system-wide Standards and Calibration Program.

  10. Primary Premier for Belfast

    ERIC Educational Resources Information Center

    McAlister, Peter

    2009-01-01

    The author talks about the Association for Science Education (ASE) Primary Science Committee's (PSC) March 2009 meeting which was held in Belfast as guests of ASE Northern Ireland. To mark the auspicious occasion of a body that usually meets four times a year in the Hatfield HQ crossing the Irish Sea to be hosted by its Celtic cousins, a Lord…

  11. From Primary to Secondary

    ERIC Educational Resources Information Center

    Anderson, Lyn

    2011-01-01

    This author discusses her decision to move to secondary school to teach mathematics, after having taught and been a mathematics manager in primary schools for six years. She states that this was a valuable experience and sparked her interest in the transition experiences of students, particularly in mathematics. Research (Evangelou et al, 2008,…

  12. Primary Art Resource Guide.

    ERIC Educational Resources Information Center

    Newton Unified School District 373, KS.

    GRADES OR AGES: Primary Grades. SUBJECT MATTER: Art. ORGANIZATION AND PHYSICAL APPEARANCE: The guide begins with a list of topics for art expression. The main body of the guide contains 15 color-coded sections on the following subjects: 1) mobiles and folded paper; 2) collage and photo montage; 3) square paper and mosaics; 4) wax paper and…

  13. Primary ectopic frontotemporal craniopharyngioma

    PubMed Central

    Ortega-Porcayo, Luis Alberto; Ponce-Gómez, Juan Antonio; Martínez-Moreno, Mauricio; Portocarrero-Ortíz, Lesly; Tena-Suck, Martha Lilia; Gómez-Amador, Juan Luis

    2015-01-01

    Introduction Primary ectopic craniopharyngiomas have only rarely been reported. Craniopharyngiomas involve usually the sellar and suprasellar region, but can be originated from cell remnants of the obliterated craniopharyngeal duct or metaplastic change of andenohypophyseal cells. We present the first case of a primary ectopic frontotemporal craniopharyngioma. Presentation of case A 35-year old woman presented with a one-year history of headache and diplopia. MRI showed a large frontotemporal cystic lesion. Tumor resection was performed with a keyhole endoscopic frontal lateral approach. The pathological features showed an adamantinomatous craniopharyngioma with a cholesterol granuloma reaction. Discussion There have been reported different localizations for primary ectopic craniopharyngioma. Our case presented a lobulated frontotemporal cystic mass formed by a dense eosinophilic proteinaceous material dystrophic calcifications and cholesterol crystals, with epithelial remnants. No tumor regrowth was observed in the magnetic resonance image 27 months postoperatively. Conclusion Primary ectopic craniopharyngioma is a rare entity with a pathogenesis that remains uncertain. This is an unusual anatomic location associated with unique clinical findings. PMID:25725331

  14. Philosophy in Primary Schools?

    ERIC Educational Resources Information Center

    White, John

    2012-01-01

    The article is a critical discussion of the aims behind the teaching of philosophy in British primary schools. It begins by reviewing the recent Special Issue of the "Journal of Philosophy of Education" Vol 45 Issue 2 2011 on "Philosophy for Children in Transition", so as to see what light this might throw on the topic just…

  15. Multiple Primary Cancer Monograph

    Cancer.gov

    To identify groups of cancer survivors that are at increased risk for multiple primary cancers, investigators led an effort to provide the first comprehensive population-based analysis of the risk of subsequent cancer in the U.S., resulting in a monograph.

  16. Restoring primary anterior teeth.

    PubMed

    Waggoner, William F

    2002-01-01

    A variety of esthetic restorative materials are available for restoring primary incisors. Knowledge of the specific strengths, weakness, and properties of each material will enhance the clinician's ability to make the best choice of selection for each individual situation. Intracoronal restorations of primary teeth may utilize resin composites, glass ionomer cements, resin-modified ionomers, or polyacid-modified resins. Each has distinct advantages and disadvantages and the clinical conditions of placement may be a strong determining factor as to which material is utilized. Full coronal restoration of primary incisors may be indicated for a number of reasons. Crowns available for restoration of primary incisors include those that are directly bonded onto the tooth, which generally are a resin material, and those crowns that are luted onto the tooth and are some type of stainless steel crown. However, due to lack of supporting clinical data, none of the crowns can be said to be superior to the others under all circumstances. Though caries in the mandibular region is rare, restorative solutions for mandibular incisors are needed. Neither stainless steel crowns nor celluloid crown forms are made specifically for mandibular incisors. Many options exist to repair carious primary incisors, but there is insufficient controlled, clinical data to suggest that one type of restoration is superior to another. This does not discount the fact that dentists have been using many of these crowns for years with much success. Operator preferences, esthetic demands by parents, the child's behavior, and moisture and hemorrhage control are all variables which affect the decision and ultimate outcome of whatever restorative treatment is chosen.

  17. Primary gastrointestinal lymphoma

    PubMed Central

    Aledavood, Amir; Nasiri, Mohammad Reza Ghavam; Memar, Bahram; Shahidsales, Soodabeh; Raziee, Hamid Reza; Ghafarzadegan, Kamran; Mohtashami, Samira

    2012-01-01

    Background: Extranodal lymphoma may arise anywhere outside lymph nodes mostly in the gastrointestinal (GI) tract as non-Hodgkin's disease. We reviewed the clinicopathological features and treatment results of patients with primary GI lymphoma. Materials and Methods: A total number of 30 cases with primary GI lymphoma were included in this study. Patients referred to the Radiation Oncology Department of Omid Hospital (Mashhad, Iran) during a 5-year period (2006-11). Clinical, paraclinical, and radiological data was collected from medical records of the patients. Results: Out of the 30 patients with primary GI lymphoma in the study, 12 were female (40%) and 18 were male (60%) (male to female ratio: 3/2). B symptoms were present in 27 patients (90%). Antidiuretic hormone (LDH) levels were elevated in 9 patients (32.1%). The most common primary site was stomach in 14 cases (46.7%). Other common sites included small intestine and colon each in 8 patients (26.7%). All patients had histopathologically proven non-Hodgkin's lymphoma. The most common histologic subtype was diffuse large B-cell lymphoma (DLBL) in 16 patients (53.3%). In addition, 28 patients (93.3%) received chemotherapy with cyclophosphamide, vincristine, doxorubicin, prednisolone (CHOP regimen). The median course of chemotherapy was 6 cources. Moreover, 8 patients (26.7%) received radiotherapy with cobalt 60. The median follow-up time was 26 months. The overall 5-year survival rate was 53% and the median survival time was 60 months. Conclusion: Primary GI lymphoma is commonly seen in stomach and small intestine and mostly is DLBCL or mucosa-associated lymphoid tissue (MALT) lymphoma. PMID:23626617

  18. JASMINE simulator

    NASA Astrophysics Data System (ADS)

    Ueda, Seiji; Yamada, Yoshiyuki; Kuwabara, Takashi; Gouda, Naoteru; Tsujimoto, Takuji; Kobayashi, Yukiyasu; Nakajima, Tadashi; Matsuhara, Hideo; Yano, Taihei; Suganuma, Masahiro; Jasmine Working Group

    2005-04-01

    We explain simulation tools in the JASMINE project (JASMINE simulator). The JASMINE project stands at the stage where its basic design will be determined in a few years. Therefore it is very important to simulate the data stream generated by astrometric fields at JASMINE in order to support investigations of error budgets, sampling strategy, data compression, data analysis, scientific performances, etc. We find that new software technologies, such as Object Oriented (OO) methodologies, are ideal tools for the simulation system of JASMINE (the JASMINE simulator). In this article, we explain the framework of the JASMINE simulator.

  19. JASMINE simulator

    NASA Astrophysics Data System (ADS)

    Yamada, Yoshiyuki; Gouda, Naoteru; Yano, Taihei; Kobayashi, Yukiyasu; Tsujimoto, Takuji; Suganuma, Masahiro; Niwa, Yoshito; Sako, Nobutada; Hatsutori, Yoichi; Tanaka, Takashi

    2006-06-01

    We explain simulation tools in JASMINE project (JASMINE simulator). The JASMINE project stands at the stage where its basic design will be determined in a few years. Then it is very important to simulate the data stream generated by astrometric fields at JASMINE in order to support investigations into error budgets, sampling strategy, data compression, data analysis, scientific performances, etc. Of course, component simulations are needed, but total simulations which include all components from observation target to satellite system are also very important. We find that new software technologies, such as Object Oriented(OO) methodologies are ideal tools for the simulation system of JASMINE(the JASMINE simulator). In this article, we explain the framework of the JASMINE simulator.

  20. Primary health care is viable.

    PubMed

    Segall, M

    1987-01-01

    'Selective primary health care' and other recent vertical health strategies have been justified on the grounds that the broad primary health care (PHC) approach cannot be afforded by developing countries in the present constrained economic circumstances. This judgement is too sweeping. A simulated case example is presented, starting with baseline health expenditure data that are representative of the situation in many developing countries. It is assumed that real economic growth occurs and that government funding of health care is allowed to grow in parallel. Two annual growth rates are considered: 2 and 5 per cent. Two restrictive conditions are applied: none of the main health services is subjected to absolute cuts; and, additional funds from existing or new sources of finance are not considered. It is shown that, even with slow growth rates, substantial increases in the funding of priority (rural and PHC) services can be achieved if the growth in expenditures of lower-priority services is curtailed. Also, savings from improved health service efficiency can be channelled to priority services. The message is that the PHC approach is viable even with slow economic growth. What is required is the technical capacity to identify and plan resource flows in the health sector, and the political will to effect resource allocations according to PHC priorities. A strategic policy like PHC should not be 'adjusted' out of effective existence because of reversible economic problems. Rather, actions should be taken to reverse the adverse economic environment. International health-related agencies should continue to support countries to develop national health systems based on PHC, and should campaign for reforms in the world economy to create at least the minimum economic conditions necessary for PHC implementation.

  1. Spirometry in primary care

    PubMed Central

    Coates, Allan L; Graham, Brian L; McFadden, Robin G; McParland, Colm; Moosa, Dilshad; Provencher, Steeve; Road, Jeremy

    2013-01-01

    Canadian Thoracic Society (CTS) clinical guidelines for asthma and chronic obstructive pulmonary disease (COPD) specify that spirometry should be used to diagnose these diseases. Given the burden of asthma and COPD, most people with these diseases will be diagnosed in the primary care setting. The present CTS position statement was developed to provide guidance on key factors affecting the quality of spirometry testing in the primary care setting. The present statement may also be used to inform and guide the accreditation process for spirometry in each province. Although many of the principles discussed are equally applicable to pulmonary function laboratories and interpretation of tests by respirologists, they are held to a higher standard and are outside the scope of the present statement. PMID:23457669

  2. [Basics of primary immunodeficiencies].

    PubMed

    Hernández-Martínez, Claudia; Espinosa-Rosales, Francisco; Espinosa-Padilla, Sara Elva; Hernández-Martínez, Ana Rosa; Blancas-Galicia, Lizbeth

    2016-01-01

    Primary immunodeficiencies (PID) are a heterogeneous group of inherited disorders, the etiology are the defects in the development or function of the immune system. The principal PID manifestations are the infections in early age, malignancy and diseases of immune dysregulation as autoimmunity and allergy. PIDs are genetics disorders and most of them are inherited as autosomal recessive, also this group of diseases is more prevalent in males and in childhood. The antibody immunodeficiency is the PID more common in adults. The more frequent disorders are the infections in the respiratory tract, abscesses, candidiasis, diarrhea, BCGosis etc. Initial approach included a complete blood count and quantification of immunoglobulins. The delay in diagnosis could be explained due to a perception that the recurrent infections are normal process or think that they are exclusively of childhood. The early diagnosis of PID by primary care physicians is important to opportune treatment and better prognosis.

  3. Primary pulmonary artery sarcoma.

    PubMed

    Jin, Tao; Zhang, Chong; Feng, Zhiying; Ni, Yiming

    2008-08-01

    Primary pulmonary artery sarcoma is an uncommon tumor. We report a case of a 73-year-old male patient with a two-week history of palpitations and shortness of breath, aggravated for two days and was believed to be pulmonary hypertension. Emergency heart ultrasound after admission presented a massive pulmonary embolism in the pulmonary artery. The patient's condition was successfully managed with urgent pulmonary artery embolectomy. The patient demonstrated improvement in hemodynamics after the operation. Histologic and immunohistochemical assays were performed and a diagnosis was made as primary pulmonary artery sarcoma arising from the left pulmonary artery. Resection of the tumor is recommended for the treatment of this rare malignant tumor. The corresponding chemotherapy, follow-up and prognosis are described as well in this case report.

  4. Primary hyperparathyroidism and nephrolithiasis.

    PubMed

    Vestergaard, Peter

    2015-05-01

    Calcifications in the kidneys may occur in the parenchyma (nephrocalcinosis), pelvis renis (nephrolithiasis) or ureters (ureterolithiasis). Several factors may protect against stone formation or promote precipitation of stones. Most stones contain calcium, and the hypercalciuria seen in primary hyperparathyroidism is a contributing factor to stone formation in the kidneys and urinary tract. In early case series, renal stone formation was frequent, whereas the proportion of patients with symptomatic renal stones has declined in recent years. However, a substantial proportion of patients presents with asymptomatic nephrocalcinosis or nephrolithiasis. Before diagnosis and treatment of primary hyperparathyroidism, renal stone events are more frequent than in the general population. However, even after surgical cure, an increased rate of renal stone events may be seen. This may to some extent be the result of stones or calcifications already present at the time of diagnosis or sequelae to prior stones such as infections or ureter strictures.

  5. [Primary orthostatic tremor].

    PubMed

    Bottin, P; Sadzot, B; Hotermans, C

    2005-02-01

    Primary orthostatic tremor is a particular tremor exclusively present when a subject is standing. Patients experience a severe disabling sense of unsteadiness. Walking, sitting and lying down are unaffected Neurological examination and cerebral imagery are normal most of the time. Electromyography in standing position confirms the diagnosis in showing a regular rapid tremor with a frequency of 12 to 18 Hz. Its physiopathology is partially unknown. A few symptomatic therapies can be proposed.

  6. [Primary care in France].

    PubMed

    Sánchez-Sagrado, T

    2016-01-01

    The poor planning of health care professionals in Spain has led to an exodus of doctors leaving the country. France is one of the chosen countries for Spanish doctors to develop their professional career. The French health care system belongs to the Bismarck model. In this model, health care system is financed jointly by workers and employers through payroll deduction. The right to health care is linked to the job, and provision of services is done by sickness-funds controlled by the Government. Primary care in France is quite different from Spanish primary care. General practitioners are independent workers who have the right to set up a practice anywhere in France. This lack of regulation has generated a great problem of "medical desertification" with problems of health care access and inequalities in health. French doctors do not want to work in rural areas or outside cities because "they are not value for money". Medical salary is linked to professional activity. The role of doctors is to give punctual care. Team work team does not exist, and coordination between primary and secondary care is lacking. Access to diagnostic tests, hospitals and specialists is unlimited. Duplicity of services, adverse events and inefficiencies are the norm. Patients can freely choose their doctor, and they have a co-payment for visits and hospital care settings. Two years training is required to become a general practitioner. After that, continuing medical education is compulsory, but it is not regulated. Although the French medical Health System was named by the WHO in 2000 as the best health care system in the world, is it not that good. While primary care in Spain has room for improvement, there is a long way for France to be like Spain.

  7. [Multiple primary pulmonary carcinomas].

    PubMed

    Guitart, A C; Gómez, G; Estrada, G; Rodríguez, C; León, C; Cornudella, R

    1991-02-01

    Three cases of multiple simultaneous primary lung carcinomas are presented, in which diagnosis was established by post-surgery pathological exam. In all three cases, chest X-ray showed pulmonary masses suggestive or clinical malignancy, and pre-surgery pathological diagnosis or squamous lung carcinoma. During thoracotomy or in the resected segment, a second lesion we confirmed which made resection necessary being this second lesion classified as lung adenocarcinoma.

  8. Simulation of LMFBR

    SciTech Connect

    Agrawal, A.K.

    1982-01-01

    The title of this session is taken to imply the system-wide thermohydraulic simulation of liquid metal fast breeder reactors (LMFBR). One is interested in predicting the temperatures, pressures, and the coolant flow rates throughout the entire plant including the reactor core, the primary and secondary sodium heat transport circuits, the steam generating system as well as other auxiliary circuits. Such a simulation is needed for (1) scoping studies (i.e., in the pre-design phase of a plant), (2) detailed design development, (3) the safety analysis (post-design development phase), and (4) the operator training and plant operation. This session emphasizes the simulation of LMFBRs for only two key categories of transients: operational disturbances or events and the post-shutdown heat removal.

  9. [Primary care in Sweden].

    PubMed

    Sánchez-Sagrado, T

    2016-09-01

    Sweden was one of the first European Union countries that saw the opportunity in the free movement of professionals. First offers for jobs were managed in 2000. Since then, a large number of professionals have taken the opportunity of a decent job and have moved from Spain to Sweden. The Swedish health care model belongs to the group of national health systems. The right to health care is linked to legal citizenship. Health is financed through regional taxes, but there is a compulsory co-payment regardless of the financial situation of the patient. The provision of health care is decentralised at a regional level, and there is a mixture of private and public medical centres. Primary care is similar to that in Spain. Health professionals work as a team with a division of tasks. Like in Spain, waiting lists and coordination between primary and specialised care are a great problem. Patients may register with any public or private primary care centre and hospital provider within their region. Access to diagnostic tests and specialists are restricted to those selected by specialists. Doctors are salaried and their job and salary depend on their experience, professional abilities and regional needs. Medicine is curative. General practitioners are the gateway to the system, but they do not act as gatekeeper. Hospitals offer a number of training post, and the access is through an interview. Continuing medical education is encouraged and financed by the health centre in order to increase its revenues.

  10. Primary gastric lymphoma

    PubMed Central

    Al-Akwaa, Ahmad M; Siddiqui, Neelam; Al-Mofleh, Ibrahim A

    2004-01-01

    AIM: The purpose of this review is to describe the various aspects of primary gastric lymphoma and the treatment options currently available. METHODS: After a systematic search of Pubmed, Medscape and MDconsult, we reviewed and retrieved literature regarding gastric lymphoma. RESULTS: Primary gastric lymphoma is rare however, the incidence of this malignancy is increasing. Chronic gastritis secondary to Helicobacter pylori (H pylori) infection has been considered a major predisposing factor for MALT lymphoma. Immune histochemical marker studies and molecular biology utilizing polymerase chain reaction have facilitated appropriate diagnosis and abolished the need for diagnostic surgical resection. Advances in imaging techniques including Magnetic Resonance Imaging (MRI) and Endoscopic Ultrasonography (EUS) have helped evaluation of tumor extension and invasion. The clinical course and prognosis of this disease is dependent on histopathological sub-type and stage at the time of diagnosis. Controversy remains regarding the best treatment for early stages of this disease. Chemotherapy, surgery and combination have been studied and shared almost comparable results with survival rate of 70%-90%. However, chemotherapy possesses the advantage of preserving gastric anatomy. Radiotherapy alone has been tried and showed good results. Stage IIIE, IVE disease treatment is solely by chemotherapy and surgical resection has been a remote consideration. CONCLUSION: We conclude that methods of diagnosis and staging of the primary gastric lymphoma have dramatically improved. The modalities of treatment are many and probably chemotherapy is superior because of high success rate, preservation of stomach and tolerable complications. PMID:14695759

  11. [Primary renal angiosarcoma].

    PubMed

    Costero-Barrios, Cesáreo B; Oros-Ovalle, Cuauhtémoc

    2004-01-01

    The twenty-fourth case of primary renal angiosarcoma is described, according to the available international literature, this present in a 71-year-old male, a mechanic by trade, without carcinogenic antecedents. Hematuria, pain in flank, and left-side tumoral mass of approximately 20 cm in diameter located in kidney by computerized axial tomography (CT) constituted manifestations. A left nefrectomy was performed. No metastasis was found. The tumor replaced 4/5 of the organ and weighed 1145 g. It showed angiomatous structure with atypical proliferation of endothelial cells in a sinusoldal trauma and anastomosatic vascular channels that invaded neighboring parenchymal and capsule. Tymorous cells were positive for CD31 and CD34 and negative for cytokeratins, S100 and HMB 45 proteins. The patient was subjected to treatment with chemotherapy and radiotherapy (lineal accelerator), but 12 months after surgery he presented retroperitonal tumoral relapse and hepatic metastasis. Diagnostic differentiation with benign vascular tumors is pointed out, as well as carcinomas and sarcomas that showed an outstanding angiomatous component, both primary and/or secondary. Primary renal angiosarcoma exposes the multiplicity of localizations that it is capable of with a tumor of this type, as well as renal parenquimatous capacity to be the seat of a great variety of neoplasias.

  12. Primary stabbing headache.

    PubMed

    Pareja, Juan A; Sjaastad, Ottar

    2010-01-01

    Primary stabbing headache is characterized by transient, cephalic ultrashort stabs of pain. It is a frequent complaint with a prevalence of 35.2%, a female preponderance, and a mean age of onset of 28 years (Vågå study). Attacks are generally characterized by moderate to severe, jabbing or stabbing pain, lasting from a fraction of a second to 3s. Attack frequency is generally low, with one or a few attacks per day. The paroxysms generally occur spontaneously, during daytime. Most patients exhibit a sporadic pattern, with an erratic, unpredictable alternation between symptomatic and non-symptomatic periods. Paroxysms are almost invariably unilateral. Temporal and fronto-ocular areas are most frequently affected. Attacks tend to move from one area to another, in either the same or the opposite hemicranium. Jabs may be accompanied by a shock-like feeling and even by head movement - "jolts" -or vocalization. On rare occasions, conjunctival hemorrhage and monocular vision loss have been described as associated features. Primary stabbing headache may concur, synchronously or independently, with other primary headaches. In contrast to what is the case in adults, in childhood it is not usually associated with other headaches. Treatment is rarely necessary. Indomethacin, 75-150 mg daily, may seem to be of some avail. Celecoxib, nifedipine, melatonin, and gabapentin have been reported to be effective in isolated cases and small series of patients. The drug studies need corroboration.

  13. Achieving Value in Primary Care: The Primary Care Value Model.

    PubMed

    Rollow, William; Cucchiara, Peter

    2016-03-01

    The patient-centered medical home (PCMH) model provides a compelling vision for primary care transformation, but studies of its impact have used insufficiently patient-centered metrics with inconsistent results. We propose a framework for defining patient-centered value and a new model for value-based primary care transformation: the primary care value model (PCVM). We advocate for use of patient-centered value when measuring the impact of primary care transformation, recognition, and performance-based payment; for financial support and research and development to better define primary care value-creating activities and their implementation; and for use of the model to support primary care organizations in transformation.

  14. Auditory perceptual simulation: Simulating speech rates or accents?

    PubMed

    Zhou, Peiyun; Christianson, Kiel

    2016-07-01

    When readers engage in Auditory Perceptual Simulation (APS) during silent reading, they mentally simulate characteristics of voices attributed to a particular speaker or a character depicted in the text. Previous research found that auditory perceptual simulation of a faster native English speaker during silent reading led to shorter reading times that auditory perceptual simulation of a slower non-native English speaker. Yet, it was uncertain whether this difference was triggered by the different speech rates of the speakers, or by the difficulty of simulating an unfamiliar accent. The current study investigates this question by comparing faster Indian-English speech and slower American-English speech in the auditory perceptual simulation paradigm. Analyses of reading times of individual words and the full sentence reveal that the auditory perceptual simulation effect again modulated reading rate, and auditory perceptual simulation of the faster Indian-English speech led to faster reading rates compared to auditory perceptual simulation of the slower American-English speech. The comparison between this experiment and the data from Zhou and Christianson (2016) demonstrate further that the "speakers'" speech rates, rather than the difficulty of simulating a non-native accent, is the primary mechanism underlying auditory perceptual simulation effects.

  15. Depth Perception in Visual Simulation.

    DTIC Science & Technology

    1980-08-01

    ANALYSIS The Visual Simulation Systems Analysis was conducted by means of on-site inspections of the ASPT simulator and a Primary Instrument Trainer at...acceptance testing of a CIG system at Luke AFB, which included aerial refueling. During the analysis of the ASPT system we witnessed a low level weapon...of being presented. Although the ASPT system at Williams AFB provides only a monochrome display, full color CIG systems are in operation at various

  16. Treatment of Primary Hyperoxaluria

    PubMed Central

    Dent, C. E.; Stamp, T. C. B.

    1970-01-01

    Nine patients with primary hyperoxaluria have been followed regularly for 1 to 11 years, and their treatment and progress are discussed in relation to the known natural history of the disease. 6 of them probably have the usual form of primary hyperoxaluria associated with increased glycollic acid excretion, while 3 who are sibs have the recently described variant associated with L-glyceric aciduria and normal glycollic acid excretion. All 9 patients have been on regimens designed to increase the urinary solubility of calcium oxalate, with or without the simultaneous lowering of urinary calcium and raising of urinary phosphate excretions. 8 patients have been treated for 1½-7½ years (average duration 4 years) with oral magnesium hydroxide, and 2 patients have been treated with sodium phosphate. One of the latter was changed after 3½ years to magnesium hydroxide and the other has been on sodium phosphate combined with a low calcium diet and cellulose phosphate continuously for 5½ years. 2, not at first diagnosed as hyperoxalurics, were first given sodium bicarbonate for their presumably secondary renal tubular acidosis. The over-all progress of the whole group is felt to have been better than could be expected from the known natural history of primary hyperoxaluria. They average 4¼ years on treatment during 5 years of our observation and all remain clinically well after an average of 9½ years since the onset of their first symptoms. Results warrant the recommendation that, until reliable means are available to decrease oxalate over-production, affected patients should be treated continuously with magnesium hydroxide. A more final opinion must await many more years of follow-up. The failure of several attempts to lower urinary oxalate excretion in these patients is also reported. ImagesFIG. 2FIG. 3FIG. 4FIG. 5 PMID:5491877

  17. Epigenetics and primary care.

    PubMed

    Wright, Robert; Saul, Robert A

    2013-12-01

    Epigenetics, the study of functionally relevant chemical modifications to DNA that do not involve a change in the DNA nucleotide sequence, is at the interface between research and clinical medicine. Research on epigenetic marks, which regulate gene expression independently of the underlying genetic code, has dramatically changed our understanding of the interplay between genes and the environment. This interplay alters human biology and developmental trajectories, and can lead to programmed human disease years after the environmental exposure. In addition, epigenetic marks are potentially heritable. In this article, we discuss the underlying concepts of epigenetics and address its current and potential applicability for primary care providers.

  18. Primary leptomeningeal melanoma.

    PubMed

    Xie, Zhao-Yu; Hsieh, Kevin Li-Chun; Tsang, Yuk-Ming; Cheung, Wing-Keung; Hsieh, Chen-Hsi

    2014-06-01

    Primary melanoma of the central nervous system is a rare melanocytic tumor typically located in the leptomeninges. We report a 57-year-old woman with an intracranial leptomeningeal melanoma who presented with myoclonic seizures. Brain CT scan and MRI revealed a hemorrhagic intracranial tumor. The tumor was completely removed and leptomeningeal melanoma was proven pathologically. Follow-up imaging studies up to 19 months showed no recurrence of the disease. Here we present radiological, gross, and pathological images of leptomeningeal melanoma, discuss its characteristics, and review the relevant literature.

  19. Primary amoebic (Naegleria) meningoencephalitis.

    PubMed

    Lam, A H; de Silva, M; Procopis, P; Kan, A

    1982-06-01

    The computed tomographic (CT) appearance of a case of primary meningoencephalitis due to the free living amoeba Naegleria fowleri is presented. The cisterns around and above the midbrain and the subarachnoid spaces were obliterated on precontrast CT. Marked enhancement in these regions was seen after intravenous contrast medium administration. The sulci and adjacent grey matter were also strongly enhanced. The ventricular size was normal. Pathological findings were those of arachnoiditis and invasion of the leptomeninges and brain substance by amoebae, especially at the base of the brain and cerebellum.

  20. Primary Ciliary Dyskinesia.

    PubMed

    Knowles, Michael R; Zariwala, Maimoona; Leigh, Margaret

    2016-09-01

    Primary ciliary dyskinesia (PCD) is a recessive genetically heterogeneous disorder of motile cilia with chronic otosinopulmonary disease and organ laterality defects in ∼50% of cases. The prevalence of PCD is difficult to determine. Recent diagnostic advances through measurement of nasal nitric oxide and genetic testing has allowed rigorous diagnoses and determination of a robust clinical phenotype, which includes neonatal respiratory distress, daily nasal congestion, and wet cough starting early in life, along with organ laterality defects. There is early onset of lung disease in PCD with abnormal airflow mechanics and radiographic abnormalities detected in infancy and early childhood.

  1. [Primary pancreatic plasmacytoma].

    PubMed

    Sánchez Acevedo, Z; Pomares Rey, B; Alpera Tenza, M R; Andrada Becerra, E

    2014-01-01

    Extramedullary plasmacytomas are uncommon malignant plasma cell tumors that present outside the bone marrow; 80% of extramedullary plasmacytomas are located in the upper respiratory tract, and gastrointestinal plasmacytomas are rare. We present the case of an asymptomatic 65-year-old man in whom a pancreatic mass was found incidentally. The lesion was determined to be a pancreatic plasmacytoma after fine-needle aspiration cytology and surgical resection. No clinical, laboratory, or imaging findings indicative of multiple myeloma or association with other plasmacytomas were found, so the tumor was considered to be a primary pancreatic plasmacytoma.

  2. [Primary esophageal lymphoma].

    PubMed

    Ximenes, Manoel; Piauilino, Marcos Amorim; Oliveira, Humberto Alves; Vaz Neto, Jorge Pinto

    2012-01-01

    We describe the case of a 54 year old woman seen with an esophageal mass diagnosed as a primary esophageal lymphoma. The main symptom was dysphagia of seven months duration. The treatment consisted in resection of the tumor, and reconstruction of the defect with a reversed pleural flap, followed by a chemotherapy regimen that consisted of five drugs, cyclophosphamid, prednisone, doxorubicin, rituximab and vincristine (R-CHOP). The patient developed an esophageal pleural fistula treated with pleural drainage and irrigation that closed in 45 days. Two and one half years later she is doing well and disease free.

  3. JASMINE Simulator

    NASA Astrophysics Data System (ADS)

    Yamada, Y.; Gouda, N.; Yano, T.; Kobayashi, Y.; Suganuma, M.; Tsujimoto, T.; Sako, N.; Hatsutori, Y.; Tanaka, T.

    2006-08-01

    We explain simulation tools in JASMINE project (JASMINE simulator). The JASMINE project stands at the stage where its basic design will be determined in a few years. Then it is very important to simulate the data stream generated by astrometric fields at JASMINE in order to support investigations of error budgets, sampling strategy, data compression, data analysis, scientific performances, etc. Of course, component simulations are needed, but total simulations which include all components from observation target to satellite system are also very important. We find that new software technologies, such as Object Oriented (OO) methodologies are ideal tools for the simulation system of JASMINE (the JASMINE simulator). The simulation system should include all objects in JASMINE such as observation techniques, models of instruments and bus design, orbit, data transfer, data analysis etc. in order to resolve all issues which can be expected beforehand and make it easy to cope with some unexpected problems which might occur during the mission of JASMINE. So, the JASMINE Simulator is designed as handling events such as photons from astronomical objects, control signals for devices, disturbances for satellite attitude, by instruments such as mirrors and detectors, successively. The simulator is also applied to the technical demonstration "Nano-JASMINE". The accuracy of ordinary sensor is not enough for initial phase attitude control. Mission instruments may be a good sensor for this purpose. The problem of attitude control in initial phase is a good example of this software because the problem is closely related to both mission instruments and satellite bus systems.

  4. Primary testicular lymphoma.

    PubMed Central

    Vural, Filiz; Cagirgan, Seckin; Saydam, Guray; Hekimgil, Mine; Soyer, Nur Akad; Tombuloglu, Murat

    2007-01-01

    We evaluated clinical features, management and survival of 12 patients with primary testicular non-Hodgkin's lymphoma presented to our hematology unit between January 1992 and July 2006, retrospectively. The median age of patients was 47 years at presentation (range 29-78 years) and > 80% of them were < 50 years old. In the majority of cases, orchidectomy was performed as diagnostic and first-line therapeutic procedures. Dominant histological subtype was diffuse large B-cell non-Hodgkin's lymphoma. Seven patients out of 12 (58%) were Ann Arbor stages I and II, and the remaining five patients (42%) were stages III and IV. All the patients received doxorubicin-based chemotherapy and achieved complete remission. The addition of rituximab and central nervous system prophylaxis with intrathecal combined chemotherapy containing methotrexate, cytarabine and dexametasone were applied to three patients who were recently admitted. The rate of relapse was 8% and progression-free survival (PFS) at 10 years was 88%. Median duration of response was 84 months (range 14-173 months), median 97.5 months of follow-up. All patients are alive and in case remission. Because of the spreading nature and relapse probability at different sites, including central nervous system and contralateral testis, systemic treatment with doxorubicin-based chemotherapy with or without prophylaxis for contralateral testis and the central nervous system seems to improve the outcome of primary testicular lymphoma. PMID:18020104

  5. Update on primary hypobetalipoproteinemia.

    PubMed

    Hooper, Amanda J; Burnett, John R

    2014-07-01

    "Primary hypobetalipoproteinemia" refers to an eclectic group of inherited lipoprotein disorders characterized by low concentrations of or absence of low-density lipoprotein cholesterol and apolipoprotein B in plasma. Abetalipoproteinemia and homozygous familial hypobetalipoproteinemia, although caused by mutations in different genes, are clinically indistinguishable. A framework for the clinical follow-up and management of these two disorders has been proposed recently, focusing on monitoring of growth in children and preventing complications by providing specialized dietary advice and fat-soluble vitamin therapeutic regimens. Other recent publications on familial combined hypolipidemia suggest that although a reduction of angiopoietin-like 3 activity may improve insulin sensitivity, complete deficiency also reduces serum cholesterol efflux capacity and increases the risk of early vascular atherosclerotic changes, despite low low-density lipoprotein cholesterol levels. Specialist laboratories offer exon-by-exon sequence analysis for the molecular diagnosis of primary hypobetalipoproteinemia. In the future, massively parallel sequencing of panels of genes involved in dyslipidemia may play a greater role in the diagnosis of these conditions.

  6. Primary pulmonary choriocarcinoma

    PubMed Central

    Snoj, Ziga; Kocijancic, Igor

    2017-01-01

    Abstract Background The aim of the study was to establish whether there are different clinical entities of primary pulmonary choriocarcinoma (PPC) that deserve different diagnostic approach and the most optimal treatment. Patients and methods A systematic review with PubMed search was conducted to identify studies that reported cases of PPC. The eligibility criteria were histological diagnosis of pulmonary choriocarcinoma and thorough examination of the reproductive organs to exclude potential primary choriocarcinoma in the gonads. Furthermore, to illustrate the review we additionally present a patient referred at our institution. Results 55 cases (17 men) were included in the review with a median age of 34 years. Women with the history of gestational event showed better survival outcome than women without the history of gestational event. Patients treated with combined modality treatment (surgery and chemotherapy) survived longer than the patients without combined modality treatment. Furthermore, multivariate analysis of prognostic factors showed that the combined modality treatment had independent prognostic significance. Size of the tumour showed significant prognostic influence in univariate and multivariate analysis. Conclusions PPC is an extreme rarity with variable clinical characteristics and outcome. It is important to capture and treat patients in the early stages of the disease. Women with the history of gestational event may show better survival, therefore genetic examination could help us to predict patient’s prognosis. Surgery followed by adjuvant chemotherapy appears to represent the best treatment for PPC. PMID:28265226

  7. [Normocalcemic primary hyperparathyroidism].

    PubMed

    Spivacow, Francisco R; Sapag Durán, Ana; Zanchetta, María B

    2014-01-01

    This report shows our conclusions on the clinical, biochemical and densitometry characteristics of 35 normocalcemic primary hyperparathyroidism (PHPT) patients. This condition is defined by a high level of intact parathyroid hormone (iPTHI) with persistently normal serum and ionized calcium in the absence of secondary hyperparathyroidism. Our selection consisted of 30 women (90%) and 5 men (10%). The control group of 55 hypercalcemic patients with primary hyperparathyroidism included 51 women (93%) and 4 men (7%). The average age at diagnosis of normocalcemic PHPT was 61.4 ± 11.7 years and 56.4 ± 11.3 years in hypercalcemic PHPT. Besides the expected differences in serum calcium, ionized calcium, phosphorus and 24 h urinary calcium, we found no significant changes in other biochemical variables, and no differences in densitometry evaluations such as the presence of osteopenia or osteoporosis and the number of fractures in the two types of PHPT. But there was a significant difference in the presence of renal lithiasis between normocalcemic PHPT (11.4%) and clasic PHPT (49.1%) p < 0.0005, to some extent associated to the presence of hypercalciuria in classic PHPT. Two of the 35 patients with normocalcemic PHPT became classic hypercalcemic PHPT over a 4 year follow-up period. Our findings support the hypothesis that the normocalcemic PHPT could be an early stage of the classic PHPT, both having similar clinical effects to metabolic renal and bone levels.

  8. Primary colonic lymphoma.

    PubMed

    Gonzalez, Quintín H; Heslin, Martin J; Dávila-Cervantes, Andrea; Alvarez-Tostado, Javier; de los Monteros, Antonio Espinosa; Shore, Gregg; Vickers, Selwyn M

    2008-03-01

    Surgical resection of primary colonic lymphoma can be an important therapeutic tool. We performed a nonrandomized retrospective descriptive study at the University hospital tertiary care center. From January 1990 to June 2002, a total of 15 patients with primary colonic lymphoma were identified from the tumor registry at University of Alabama at Birmingham and retrospectively reviewed under Institutional Review Board approved protocol. Demographic data, clinical features, treatment method (surgery and/or chemotherapy), recurrence rate, and survival were analyzed. The results are presented as mean +/- standard deviation or median and range. Differences in survival were evaluated by the log-rank test and the interval of disease-free survival was calculated using the Kaplan-Meier method. A P value of <0.05 was considered statistically significant. Main outcome measures included surgical results, morbidity, mortality, and recurrence rate. Mean age was 51.5 years (standard deviation 16.4), 33 per cent were male and 67 per cent were female. Presenting symptoms were diarrhea (53.5%), lower gastrointestinal bleeding (13.3%), and nausea and vomiting (46.7%) secondary to low-grade obstruction. Concomitant colorectal disease was present in one patient with ulcerative colitis. Preoperative diagnosis of lymphoma was made in 13 patients (87%) with colonoscopy and biopsy. CT scan was performed in all patients; and none had radiographic evidence of systemic extension. Only one patient had a history of lymphoproliferative disease and exposure to radiation. The most common disease location was the cecum (60%), followed by the right colon (27%), and the sigmoid colon (13%). The mean lactic dehydrogenase (LDH) value was 214.9 u/L (range 129-309). Thirty-three per cent of the patients had an LDH value that was above the upper normal limit. LDH returned to normal after treatment in all patients. Operations performed consisted of right hemicolectomy (13), total proctocolectomy with ileal

  9. TSH Regulation Dynamics in Central and Extreme Primary Hypothyroidism

    PubMed Central

    Santini, Ferruccio; Marsili, Alessandro; Pinchera, Aldo; DiStefano, Joseph J.

    2010-01-01

    Background Thyrotropin (TSH) changes in extreme primary hypothyroidism include increased secretion, slowed degradation, and diminished or absent TSH circadian rhythms. Diminished rhythms are also observed in central hypothyroid patients and have been speculated to be a cause of central hypothyroidism. We examined whether TSH secretion saturation, previously suggested in extreme primary hypothyroidism, might explain diminished circadian rhythms in both disorders. Methods We augmented and extended the range of our published feedback control system model to reflect nonlinear changes in extreme primary hypothyroidism, including putative TSH secretion saturation, and quantified and validated it using multiple clinical datasets ranging from euthyroid to extreme hypothyroid (postthyroidectomy). We simulated central hypothyroidism by reducing overall TSH secretion and also simulated normal TSH secretion without circadian oscillation, maintaining plasma TSH at constant normal levels. We also utilized the validated model to explore thyroid hormone withdrawal protocols used to prepare remnant ablation in thyroid cancer patients postthyroidectomy. Results Both central and extreme primary hypothyroidism simulations yielded low thyroid hormone levels and reduced circadian rhythms, with simulated daytime TSH levels low-to-normal for central hypothyroidism and increased in primary hypothyroidism. Simulated plasma TSH showed a rapid rise immediately following triiodothyronine (T3) withdrawal postthyroidectomy, compared with a slower rise after thyroxine withdrawal or postthyroidectomy without replacement. Conclusions Diminished circadian rhythms in central and extreme primary hypothyroidism can both be explained by pituitary TSH secretion reaching maximum capacity. In simulated remnant ablation protocols using the extended model, TSH shows a more rapid rise after T3 withdrawal than after thyroxine withdrawal postthyroidectomy, supporting the use of replacement with T3 prior to 131I

  10. Simulation Accelerator

    NASA Technical Reports Server (NTRS)

    1998-01-01

    Under a NASA SBIR (Small Business Innovative Research) contract, (NAS5-30905), EAI Simulation Associates, Inc., developed a new digital simulation computer, Starlight(tm). With an architecture based on the analog model of computation, Starlight(tm) outperforms all other computers on a wide range of continuous system simulation. This system is used in a variety of applications, including aerospace, automotive, electric power and chemical reactors.

  11. Primary and secondary stabbing headache.

    PubMed

    Robbins, Matthew S; Evans, Randolph W

    2015-04-01

    Eight out of the 33 cases of primary stabbing headache seen in a general neurology clinic (40% have headache as their chief complaint) in the last 3.5 years are presented. The epidemiology, association with other primary headache disorders, secondary associations, testing, and treatment of primary stabbing headache are reviewed.

  12. Primary malignant melanoma of prostate.

    PubMed

    Doublali, M; Chouaib, A; Khallouk, A; Tazi, M F; El Fassi, M J; Farih, My H; Elfatmi, H; Bendahou, M; Benlemlih, A; Lamarti, O

    2010-05-01

    Primary genitourinary melanoma accounts for less than one per cent of all cases of melanoma. Most cases attributed to the prostate actually originate from the prostatic urethra. Due to its infrequency, primary malignant melanoma of the genitourinary tract presents a difficult diagnostic and management challenge. We report a case of primary malignant melanoma of the prostate found during transurethral resection of the prostate.

  13. Professional Issues for Primary Teachers

    ERIC Educational Resources Information Center

    Browne, Ann, Ed.; Haylock, Derek, Ed.

    2004-01-01

    This book is intended to be a contribution to raising the awareness of primary teachers and trainee teachers as to what is involved in all the different professional dimensions of their work in schools. The book deals with the key professional issues in primary teaching that are addressed in primary teacher training courses. The book aims to…

  14. Primary care: the next renaissance.

    PubMed

    Showstack, Jonathan; Lurie, Nicole; Larson, Eric B; Rothman, Arlyss Anderson; Hassmiller, Susan

    2003-02-04

    Three decades ago, a renaissance helped create the foundations of primary care as we know it today. In recent years, however, new challenges have confronted primary care. We believe that the current challenges can be overcome and may, in fact, present an opportunity for a new renaissance of primary care to address the needs of our population. In this paper, we suggest seven core principles and a set of actions that will support a renaissance in, and a positive future for, primary care. The seven principles are 1) Health care must be organized to serve the needs of patients; 2) the goal of primary care systems should be the delivery of the highest-quality care as documented by measurable outcomes; 3) information and information systems are the backbone of the primary care process; 4) current health care systems must be reconstructed; 5) the health care financing system must support excellent primary care practice; 6) primary care education must be revitalized, with an emphasis on new delivery models and training in sites that deliver excellent primary care; and 7) the value of primary care practice must be continually improved, documented, and communicated. At the start of the 21st century, a vital, patient-centered primary care system has much to offer a rapidly changing population with increasingly diverse needs and expectations. If we keep the needs of persons and patients clearly in sight and design systems to meet those needs, primary care will thrive and our patients will be well served.

  15. Primary Teacher Education in Malaysia

    ERIC Educational Resources Information Center

    Ching, Chin Phoi; Yee, Chin Peng

    2012-01-01

    In Malaysia the training of primary school teachers is solely carried out by teacher training institutes which offer the Bachelor of Teaching with Honors (Primary education) program and was first launched in 2007. This program prepares primary school teachers specializing in various subjects or major and is carried out in 27 teacher training…

  16. MAPE - Micros and Primary Education.

    ERIC Educational Resources Information Center

    Jones, R.

    1982-01-01

    The nature of the organization, Micros and Primary Education (MAPE) is detailed, and its history and development are noted. The primary purpose of MAPE is to promote and develop awareness and effective use of microelectronics as an integral part of the philosophy and practice of Primary Education. (MP)

  17. Primary Aluminum Plants Worldwide - 1998

    USGS Publications Warehouse

    1999-01-01

    The 1990 U.S. Bureau of Mines publication, Primary Aluminum Plants Worldwide, has been updated and is now available. The 1998 USGS edition of Primary Aluminum Plants Worldwide is published in two parts. Part I—Detail contains information on individual primary smelter capacity, location, ownership, sources of energy, and other miscellaneous information. Part II—Summary summarizes the capacity data by country

  18. Morphology Evolution of Primary Particles in Lspsf Rheocasting Process

    NASA Astrophysics Data System (ADS)

    Guo, Hong-Min; Yang, Xiang-Jie

    Experimental and microstructure simulation approaches were taken to investigate the morphological evolutions of primary particles in an Al-20wt% pct Cu alloy under LSPSF (low superheat pouring with a shear field) rheocasting conditions. The results indicate that crystals are globular and present in non-entrapped eutectic, after 3s of solidification. The morphology of these crystals during the subsequent free growth is determined by both the number of free crystals and the cooling intensity of melt. Analyzed results from microstructure simulation and two stability models suggest that the primary globular particles formed in the earlier stage of solidification can attain growth stability up to a larger size scale.

  19. [Antibiotics in primary care].

    PubMed

    Steciwko, Andrzej; Lubieniecka, Małgorzata; Muszyńska, Agnieszka

    2011-05-01

    Discovered in the forties of the twentieth century antimicrobial agents have changed the world. Currently, due to their overuse, we are threatened by the increasing resistance of bacteria to antibiotics, and soon we may face a threat of inability to fight these pathogens. For that reason, the world, European and national organizations introduce antibiotics protection programs. In Poland since 2004, the National Program of Protection of Antibiotics is being held. The concept of rational antibiotic therapy is associated not only with the appropriate choice of therapy or antimicrobial dosage but also with a reduction in costs associated with a refund of medicines. Antibiotics are prescribed mostly by primary care physicians (GP), and about one fifth of visits to family doctor's office ends with prescribing antimicrobial drug. These trends are probably related to both the difficulty in applying the differential diagnosis of viral and bacterial infection in a primary care doctor's office, as well as patient's conviction about the effectiveness of antibiotic therapy in viral infections. However, although patients often want to influence the therapeutic decisions and ask their doctor for prescribing antimicrobial drug, the right conversation with a doctor alone is the critical component in satisfaction with medical care. Many countries have established standards to clarify the indications for use of antibiotics and thereby reduce their consumption. The next step is to monitor the prescribing and use of these drugs and to assess the rise of drug resistance in the area. In Poland, the recommendations regarding outpatient respiratory tract infections treatment were published and usage of antimicrobial agents monitoring has begun. However, lack of publications covering a broad analysis of antibiotic therapy and drug resistance on Polish territory is still a problem. Modem medicine has yet another tool in the fight against bacteria--they are bacteriophages. Phage therapy is

  20. Primary hyperparathyroidism in pregnancy.

    PubMed

    Kamenický, Peter; Lecoq, Anne-Lise; Chanson, Philippe

    2016-06-01

    Primary hyperparathyroidism (PHPT) is one of the most common endocrine disorders in the general population but is rarely diagnosed during pregnancy. Symptoms of gestational PHPT may be unrecognized, or masked by physiological changes in calcium homeostasis associated with pregnancy. Gestational PHPT may have severe consequences for both mother and fetus. However, nowadays, gestational PHPT is usually diagnosed in earlier stages and milder forms, with low complication rates. Treatment should be individually tailored according to gestational age, the severity of hypercalcemia, and the risk-benefit balance. The conservative approach is preferred in mild forms, whereas surgery, usually performed during the second trimester, is reserved for symptomatic hypercalcemic PHPT. Given the young age of the patients, genetic causes should be considered.