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Sample records for simulated pwr primary

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

    SciTech Connect

    Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2009-12-01

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

  2. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-05-01

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

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

    SciTech Connect

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

    2013-08-15

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

  5. 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.

  6. 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.

  7. 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

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

    SciTech Connect

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

    1997-04-01

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

  9. 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. Calculational limitations in PWR system simulation

    SciTech Connect

    Abramson, P.B.; Kennedy, M.F.; Speis, T.P.

    1982-01-01

    Engineering transient analysis codes, which are in general more accurate than the present generation of simulator software, can be expected to yield reasonably accurate results (+-20% or so on system pressure) if carefully utilized and if the two-phase and transient flow conditions are not severe. As the severity of the transient increases, the confidence that one may have in the results decreases. None of the existing engineering analysis codes is well assessed or verified for transient analysis, but all give qualitatively the same results lending credence to their results. Recent comparisons to transients in LOFT and SEMISCALE are encouraging as are various comparisons to actual plant data.

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

    SciTech Connect

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

    1996-06-01

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

  12. Primary: A Simulation Game.

    ERIC Educational Resources Information Center

    Shelly, Ann

    A game designed to introduce elementary and middle school students to the primary election process and its operational elements through simulation and role playing activities is presented in this document. A hypothetical town of Notae which has been controlled by one political party for years without resulting change is described. Issues in the…

  13. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR

    SciTech Connect

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed.

  14. Primary-side deposits on PWR steam-generator tubes

    SciTech Connect

    Bergmann, C.A.; Roesmer, J.; Perone, D.W.

    1983-03-01

    The evaluation of analyses of material removed from primary side steam generator tubing samples taken from nuclear plants that had operated for up to 7 effective full power years are presented in this report. The types of analyses incuded radiochemical, chemical, scanning electron microscope (SEM), and energy dispersive x-ray (EDAX) techniques to characterize the surfaces and composition of the tubing material. An evaluation of the data obtained and a comparison with in-core crud data and with values calculated by a mathematical activity transport model (CORA) are also given in the report.

  15. Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment

    NASA Astrophysics Data System (ADS)

    Zhang, Litao; Wang, Jianqiu

    2014-03-01

    Stress corrosion crack growth tests of a cold worked nuclear grade 316L stainless steel were conducted in simulated pressurized water reactor (PWR) primary water environment containing various dissolved oxygen (DO) contents but no dissolved hydrogen. The crack growth rate (CGR) increased with increasing DO content in the simulated PWR primary water. The fracture surface exhibited typical intergranular stress corrosion cracking (IGSCC) characteristics.

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

    NASA Astrophysics Data System (ADS)

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

    2007-05-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2011-12-01

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

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

    SciTech Connect

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

    2013-07-01

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

  19. 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.

  20. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    SciTech Connect

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

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  1. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 1. Summary, Load Combination Program. Project I final report

    SciTech Connect

    Lu, S.; Streit, R.D.; Chou, C.K.

    1981-06-01

    This report summarizes work performed to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading and to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR, is the demonstration plant used in this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated by a deterministic fracture mechanics model with stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without earthquake, is very small (on the order of 10/sup -12/). The probability of a leak was found to be several orders of magnitude greater than that of a large LOCA, complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported.

  2. 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.

  3. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2006-04-01

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

  5. 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.

  6. 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.

  7. Proceedings: 1983 Workshop on Primary-Side Stress Corrosion Cracking of PWR Steam Generator Tubing

    SciTech Connect

    1987-11-01

    Utility and vendor representatives from around the world met to share information on stress corrosion cracking of steam generator tubing from the primary side. In 32 presentations, speakers discussed in-plant experience with the phenomenon and related laboratory data. The workshop was the first to present results of remedial stress relief programs.

  8. Proceedings: 1985 Workshop on Primary-Side Stress Corrosion Cracking of PWR Steam Generator Tubing

    SciTech Connect

    1987-06-01

    To date, more than 30 PWRs have reported stress corrosion cracking of steam generator tubing from the primary water side. In 32 presentations, this report offers in-plant and laboratory data on the contributing factors, as well as discussing some promising remedial measures.

  9. Multidimensional effects in the thermal response of fuel rod simulators. [PWR

    SciTech Connect

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

    1980-01-01

    One of the primary objectives of the Oak Ridge National Laboratory Pressurized-Water Reactor Blowdown Heat Transfer Separate-Effects Program is the determination of the transient surface temperature and surface heat flux of fuel pin simulators (FPSs) from internal thermocouple signals obtained during a loss-of-coolant experiment (LOCE) in the Thermal-Hydraulics Test Facility. This analysis requires the solution of the classical inverse heat conduction problem. The assumptions that allow the governing differential equation to be reduced to one dimension can introduce significant errors in the computed surface heat flux and surface temperature. The degree to which these computed variables are perturbed is addressed and quantified.

  10. Loads on steam generator tubes during simulated loss-of-coolant accident conditions. Final report. [PWR

    SciTech Connect

    Guerrero, H.N.; Hiestand, J.W.; Rossano, F.V.; Shah, P.K.; Thakkar, J.G.

    1982-11-01

    This report presents the work performed to verify the CEFLASH digital computer code modeling of the hydro-dynamic loads in a steam generator tube during a loss-of-coolant accident (LOCA). The test loop simulated the primary side thermal-hydraulic conditions in an operational nuclear steam generator. The loop consisted of 5 full size double 90/sup 0/ bend tubes and steam generator plena, a pressurizer, a reactor resistance simulator, a heater, a pump, and associated pipes and valves to complete the system. The tubes used were of typical length and the same outside diameter as those used in C-E steam generators. Prototypical supports were provided for the bundle of 5 tubes. Cold leg guillotine breaks were simulated using quick opening valve and rupture disks. Break opening times ranged from less than 1 msec to as much as 67 milliseconds. The loop instrumentation was designed to measure the transient pressure history at various locations and monitor the structural response of the tube to the LOCA hydrodynamic loading. A series of blowdown tests was performed for different operating and boundary conditions. Analytically predicted transient pressure histories and the differential pressure history across the tube span were compared with the experimental data.

  11. Multidimensional effects in the thermal response of fuel pin simulators. [PWR

    SciTech Connect

    Ott, L.J.

    1980-01-01

    One of the primary objectives of the Oak Ridge National Laboratory Pressurized-Water Reactor Blowdown Heat Transfer Separate-Effects Program is the determination of the transient surface temperature and surface heat flux of fuel pin simulators (FPSs) from internal thermocouple signals obtained during a loss-of-coolant experiment (LOCE) in the Thermal-Hydraulics Test Facility. This analysis requires the solution of the classical inverse heat conduction problem. The assumptions that allow the governing differential equation to be reduced to one dimension can introduce significant errors in the computed surface heat flux and surface temperature. The degree to which these computed variables are perturbed is addressed and quantified, and experimental evidence is presented which indicates significant differences in surface behavior circumferentially at one axial plane of an FPS during a LOCE.

  12. Effect of temperature and dissolved hydrogen on oxide films formed on Ni and Alloy 182 in simulated PWR water

    NASA Astrophysics Data System (ADS)

    Mendonça, R.; Bosch, R.-W.; Van Renterghem, W.; Vankeerberghen, M.; de Araújo Figueiredo, C.

    2016-08-01

    Alloy 182 is a nickel-based weld metal, which is susceptible to stress corrosion cracking in PWR primary water. It shows a peak in SCC susceptibility at a certain temperature and hydrogen concentration. This peak is related to the electrochemical condition where the Ni to NiO transition takes place. One hypothesis is that the oxide layer at this condition is not properly developed and so the material is not optimally protected against SCC. Therefore the oxide layer formed on Alloy 182 is investigated as a function of the dissolved hydrogen concentration and temperature around this Ni/NiO transition. Exposure tests were performed with Alloy 182 and Ni coupons in a PWR environment at temperatures between 300 °C and 345 °C and dissolved hydrogen concentration between 5 and 35 cc (STP)H2/kg. Post-test analysis of the formed oxide layers were carried out by SEM, EDS and XPS. The exposure tests with Ni coupons showed that the Ni/NiO transition curve is at a higher temperature than the curve based on thermodynamic calculations. The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures, but that the morphology changed from spinel crystals to needle like oxides when the Ni/NiO transition curve was approached. Oxide layers were present below the Ni/NiO transition curve i.e. when the Ni coupon was still free of oxides. In addition an evolved slip dissolution model was proposed that could explain the observed experimental results and the peak in SCC susceptibility for Ni-based alloys around the Ni/NiO transition.

  13. 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.

  14. 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)

  15. 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

  16. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    NASA Astrophysics Data System (ADS)

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

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

    SciTech Connect

    Chou, C.K.

    1981-06-01

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

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

    SciTech Connect

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

    1982-09-01

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

  19. 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.

  20. 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

  1. 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.

  2. Limits on the experimental simulation of nuclear fuel rod response. [PWR

    SciTech Connect

    Hagar, R.C.

    1980-01-01

    The steady-state and transient effects of intrinxic geometric and material property differences between typical nuclear fuel pins and electric fuel pin simulators (FPSs) are identified. The effectiveness of varying the transient power supplied to the FPS in reducing the differences between the transient responses of nuclear fuel pins and FPSs is investigated. This effectiveness is shown to be limited by the heat capacity of the FPS, the allowed range of the power program, and different FPS power requirements at different positions on a full-length FPS.

  3. CRACK GROWTH RESPONSE OF ALLOY 690 IN SIMULATED PWR PRIMARY WATER

    SciTech Connect

    Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2009-12-01

    The stress corrosion crack growth response of three extruded alloy 690 CRDM tube heats was investigated in several thermomechanical conditions. Extremely low propagation rates (< 1 x 10{sup -9} mm/s) were observed under constant stress intensity factor (K) loading at 325-350 C in the as-received, thermally treated (TT) materials despite using a variety of transitioning techniques. Post-test observation of the crack-growth surfaces revealed only isolated intergranular (IG) cracking. One-dimensional cold rolling to 17% reduction and testing in the S-L orientation did not promote enhanced stress corrosion rates. However, somewhat higher propagation rates were observed in a 30% cold-rolled alloy 690TT specimen tested in the T-L orientation. Cracking of the cold-rolled material was promoted on grain boundaries oriented parallel to the rolling plane with the % IG increasing with the amount of cold rolling.

  4. 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.

  5. 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.

  6. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 6: failure mode analysis. Final report

    SciTech Connect

    Streit, R.D.

    1981-09-01

    Material properties and failure criteria were evaluated to assess the requirements for double-ended guillotine break in the primary coolant loop of the Zion Unit 1 pressurized water reactor. The properties of the 316 stainless steel piping materials were obtained from the literature. Statistical distributions of both the tensile and fracture properties at room and operating temperatures were developed. Yield and ultimate strength tensile properties were combined to estimate the material flow strength. The flow strength and fracture properties were used in the various failure models analyzed. Linear-elastic, elastic-plastic, and fully plastic fracture models were compared, and the governing fracture criterion was determined. For the particular case studied, the fully plastic flow requirement was found to be the controlling fracture criterion leading to a double-ended guillotine pipe break.

  7. 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.

  8. 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.

  9. 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.

  10. 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)

    Javed Iqbal, M.; 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.

  11. 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…

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

    NASA Astrophysics Data System (ADS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

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

  13. 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

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

    SciTech Connect

    Wahi, K.K.

    1980-08-01

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

  15. 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 Poisson's 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.

  16. Primary simulation and experimental results of a coaxial plasma accelerator

    NASA Astrophysics Data System (ADS)

    Chen, Z.; Huang, J.; Han, J.; Zhang, Z.; Quan, R.; Wang, L.; Yang, X.; Feng, C.

    A coaxial plasma accelerator with a compressing coil is developed to simulate the impacting and erosion effect of space debris on exposed materials of spacecrafts During its adjustment operation some measurements are conducted including discharging current by Rogowski coil average plasma speed in the coaxial gun by magnetic coils and ejected particle speed by piezoelectric sensor etc In concert with the experiment a primary physical model is constructed in which only the coaxial gun is taken into account with the compressor coil not considered for its unimportant contribution to the plasma ejection speed The calculation results by the model agree well with the diagnostic results considering some assumptions for simplification Based on the simulation result some important suggestions for optimum design and adjustment of the accelerator are obtained for its later operation

  17. 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.

  18. 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

  19. Modeling the electrochemistry of the primary circuits of light water reactors

    SciTech Connect

    Bertuch, A.; Macdonald, D.D.; Pang, J.; Kriksunov, L.; Arioka, K.

    1994-12-31

    To model the corrosion behaviors of the heat transport circuits of light water reactors, a mixed potential model (NTM) has been developed and applied to both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Using the data generated by the GE/UKEA-Harwell radiolysis model, electrochemical potentials (ECPs) have been calculated for the heat transport circuits of eight BWRs operating under hydrogen water chemistry (HWC). By modeling the corrosion behaviors of these reactors, the effectiveness of HWC at limiting IGSCC and IASCC can be determined. For simulating PWR primary circuits, a chemical-radiolysis model (developed by the authors) was used to generate input parameters for the MPM. Corrosion potentials of Type 304 and 316 SSs in PWR primary environments were calculated using the NTM and were found to be in good agreement with the corrosion potentials measured in the laboratory for simulated PWR primary environments.

  20. 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.

  1. Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 6. Failure mode analysis load combination program. Project I, final report

    SciTech Connect

    Streit, R.D.

    1981-06-01

    Material properties and failure criteria were evaluated to assess the requirements for double-ended guillotine break in the primary coolant loop of the Zion Unit 1 pressurized water reactor. The properties of the 316 stainless steel piping materials were obtained from the literature. Statistical distributions of both the tensile and fracture properties at room and operating temperatures were developed. Yield and ultimate strength tensile properties were combined to estimate the material flow strength. The flow strength and fracture properties were used in the various failure models analyzed. Linear-elastic, elastic-plastic, and fully plastic fracture models were compared, and the governing fracture criterion was determined. For the particular case studied, the fully plastic requirement was found to be the controlling fracture criterion leading to a double-ended guillotine pipe break.

  2. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9. PRAISE computer code user's manual. Load Combination Program Project I final report

    SciTech Connect

    Lim, E.Y.

    1981-06-01

    The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor. PRAISE has a very modular structure and can be tailored to a variety of crack growth and piping reliability problems. Although PRAISE was developed on a CDC-7600 computer, it was, however, coded in standard FORTRAN IV and is readily transportable to other machines.

  3. Duodenal plasmacytoma. A rare primary extramedullary localization simulating a carcinoma.

    PubMed

    Pentimone, F; Camici, M; Cini, G; Levorato, D

    1979-01-01

    A patient with duodenal plasmacytoma is described. The extramedullary localization preceded the complete humoural picture of multiple myeloma by 15 months. The previously reported cases of myeloma of primary gastrointestinal localization are reviewed. PMID:108892

  4. Computer simulation of thermal modeling of primary lithium cells

    NASA Technical Reports Server (NTRS)

    Young, I. Cho; Frank, Harvey; Halpert, Gersid

    1987-01-01

    The objective was to gain a better understanding of the safety problem of primary Li-SOCl2 and Li-SO2 cells by carrying out detailed thermal modeling work. In particular, the transient heat generation rates during moderate and extermely high discharge rate tests of Li-SOCl2 cells were predicted and compared with those from the electrochemical heating. The difference between the two may be attributed to the lithium corrosion and other chemical reactions. The present program was also tested for charging of Li-SO2. In addition, the present methodology should be applicable to other primary cylindrical cells as well as rechargeable battery analyses with minor modifications.

  5. 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.

  6. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGESBeta

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

    2016-07-17

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

  7. 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)

  8. Hypokalemic paralysis due to primary hyperaldosteronism simulating Gitelman's syndrome.

    PubMed

    Kasifoglu, Timucin; Akalin, Aysen; Cansu, Dondu Uskudar; Korkmaz, Cengiz

    2009-03-01

    Some diseases, such as Gitelman's syndrome, Bartter's syndrome, and primary hyperaldosteronism (Conn's syndrome), may bear some similar clinical and laboratory findings. Their treatment modalities being different from one another, the need for a scrupulous diagnostic evaluation arises as far as clinical practice is concerned. In this report, we present a patient with Conn's syndrome who was initially considered to have Gitelman's syndrome due to displaying a few overlapping features of both diseases. We also give an account of the hardships encountered during the diagnostic evaluation. PMID:19237821

  9. Advanced material modelling in numerical simulation of primary acetabular press-fit cup stability.

    PubMed

    Souffrant, R; Zietz, C; Fritsche, A; Kluess, D; Mittelmeier, W; Bader, R

    2012-01-01

    Primary stability of artificial acetabular cups, used for total hip arthroplasty, is required for the subsequent osteointegration and good long-term clinical results of the implant. Although closed-cell polymer foams represent an adequate bone substitute in experimental studies investigating primary stability, correct numerical modelling of this material depends on the parameter selection. Material parameters necessary for crushable foam plasticity behaviour were originated from numerical simulations matched with experimental tests of the polymethacrylimide raw material. Experimental primary stability tests of acetabular press-fit cups consisting of static shell assembly with consecutively pull-out and lever-out testing were subsequently simulated using finite element analysis. Identified and optimised parameters allowed the accurate numerical reproduction of the raw material tests. Correlation between experimental tests and the numerical simulation of primary implant stability depended on the value of interference fit. However, the validated material model provides the opportunity for subsequent parametric numerical studies. PMID:22817471

  10. 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).

  11. PHOSPHOROUS AND NITROGEN LIMITATION OF PRIMARY PRODUCTION IN A SIMULATED ESTUARINE GRADIENT

    EPA Science Inventory

    The transition between phosphorus limitation of primary production in freshwater and nitrogen limitation in seawater was examined along an estuarine gradient simulated in 4 large 13 M3 enclosures connected in a series and containing pelagic and benthic subsystems. ominal saliniti...

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

    SciTech Connect

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

    1982-05-01

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

  13. 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.

  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. 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.

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

    SciTech Connect

    Russell, M.L.

    1981-01-01

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

  17. Scatter correction using a primary modulator for dual energy digital radiography: A Monte Carlo simulation study

    NASA Astrophysics Data System (ADS)

    Jo, Byung-Du; Lee, Young-Jin; Kim, Dae-Hong; Kim, Hee-Joung

    2014-08-01

    In conventional digital radiography (DR) using a dual energy subtraction technique, a significant fraction of the detected photons are scattered within the body, making up the scatter component. Scattered radiation can significantly deteriorate image quality in diagnostic X-ray imaging systems. Various methods of scatter correction, including both measurement- and non-measurement-based methods, have been proposed in the past. Both methods can reduce scatter artifacts in images. However, non-measurement-based methods require a homogeneous object and have insufficient scatter component correction. Therefore, we employed a measurement-based method to correct for the scatter component of inhomogeneous objects from dual energy DR (DEDR) images. We performed a simulation study using a Monte Carlo simulation with a primary modulator, which is a measurement-based method for the DEDR system. The primary modulator, which has a checkerboard pattern, was used to modulate the primary radiation. Cylindrical phantoms of variable size were used to quantify the imaging performance. For scatter estimates, we used discrete Fourier transform filtering, e.g., a Gaussian low-high pass filter with a cut-off frequency. The primary modulation method was evaluated using a cylindrical phantom in the DEDR system. The scatter components were accurately removed using a primary modulator. When the results acquired with scatter correction and without scatter correction were compared, the average contrast-to-noise ratio (CNR) with the correction was 1.35 times higher than that obtained without the correction, and the average root mean square error (RMSE) with the correction was 38.00% better than that without the correction. In the subtraction study, the average CNR with the correction was 2.04 (aluminum subtraction) and 1.38 (polymethyl methacrylate (PMMA) subtraction) times higher than that obtained without the correction. The analysis demonstrated the accuracy of the scatter correction and the

  18. 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.

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

    SciTech Connect

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

    2014-01-01

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

  20. 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.

  1. PWR Cross Section Libraries for ORIGEN-ARP

    SciTech Connect

    McGraw, Carolyn; Ilas, Germina

    2012-01-01

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

  2. 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.

  3. 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.

  4. 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.

  5. Nanoseconds molecular dynamics simulation of primary mechanical energy transfer steps in F1-ATP synthase.

    PubMed

    Böckmann, Rainer A; Grubmüller, Helmut

    2002-03-01

    The mitochondrial membrane protein FoF1-ATP synthase synthesizes adenosine triphosphate (ATP), the universal currency of energy in the cell. This process involves mechanochemical energy transfer from a rotating asymmetric gamma-'stalk' to the three active sites of the F1 unit, which drives the bound ATP out of the binding pocket. Here, the primary structural changes associated with this energy transfer in F1-ATP synthase were studied with multi-nanosecond molecular dynamics simulations. By forced rotation of the gamma-stalk that mimics the effect of proton motive Fo-rotation during ATP synthesis, a time-resolved atomic model for the structural changes in the F1 part in terms of propagating conformational motions is obtained. For these, different time scales are found, which allows the separation of nanosecond from microsecond conformational motions. In the simulations, rotation of the gamma-stalk lowers the ATP affinity of the betaTP binding pocket and triggers fast, spontaneous closure of the empty betaE subunit. The simulations explain several mutation studies and the reduced hydrolysis rate of gamma-depleted F1-ATPase. PMID:11836535

  6. 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.

  7. 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%. PMID:17129660

  8. 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.

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

    SciTech Connect

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

    1982-01-01

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

  10. Future Arctic Primary Productivity from CMIP5 Simulations: Uncertain Outcome, but Consistent Mechanisms

    NASA Astrophysics Data System (ADS)

    Vancoppenolle, Martin; Bopp, Laurent; Madec, Gurvan; Dunne, John; Ilyina, Tatiana; Halloran, Paul; Steiner, Nadja

    2013-04-01

    Net primary production (PP) in the Arctic should increase over this century, due to sea ice retreat, inducing an increase in available light, but could decrease if nitrate renewal is insufficient. Here, simulations performed with 11 Earth System Models from the CMIP5 exercise, covering 1900-2100, are analyzed using Arctic PP, surface nitrate and sea ice concentrations. Whereas the mean model well simulates Arctic-integrated PP at 511 TgC/yr for 1998-2005 and projects a 58 TgC/yr increase by 2080-2099, models neither agree on what limits PP today, nor on the sign of future PP change. However, the same mechanisms operate in all models. First, both sea ice and nitrate decrease over the 21st century. Depending on the model, the strengthening nitrate stress is sufficient to overcome the effect of light increase. The inter-model spread stems from present nitrate stocks, poorly constrained by observations and characterized by an inter-model uncertainty of >50% of the mean. Second, virtually all models agree in the open ocean zones on more spatially-integrated PP and less PP per unit area. Where models disagree is the sea ice zone, where a subtle balance between light and nutrient limitations determines the change in productivity. Hence, it is argued that reducing uncertainty on present Arctic nitrate would render Arctic PP projections much more consistent. That is definitely required to understand the impact of climate change on the Arctic food webs and carbon cycle.

  11. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 8. Pipe fracture indirectly induced by an earthquake. Load Combination Program, Project I final report

    SciTech Connect

    Streit, R.D.

    1981-06-01

    This volume considers the probability that a double-ended guillotine break in the primary coolant loop of a pressurized water reactor occurs simultaneously with (and is indirectly caused by) a seismic event. The pipe break is a consequence of a seismically initiated failure in a system other than the primary piping itself. Events studied that can lead to an indirectly induced pipe break include structural and mechanical failures, missile impact, pressure transients, jet impingement, fire, and explosion. Structural failures of the supports for the reactor pressure vessel, reactor coolant pump, and steam generator have the highest probability of causing a double-ended pipe break. Furthermore, we found that structural failure of the containment dome and failure of the reactor coolant pump flywheel have the highest potential for a missile-caused pipe break. Since structural failure proved to be a major factor, we developed a model to estimate the probability of structural failure; this model is based on the engineering factors of safety and seismic hazard. preliminary results indicate that the probability of a double-ended pipe break indirectly caused by a seismic event during the plant life is on the order of 10/sup -9/.

  12. Morphologies of Primary Silicon in Hypereutectic Al-Si Alloys: Phase-Field Simulation Supported by Key Experiments

    NASA Astrophysics Data System (ADS)

    Wang, Kai; Wei, Ming; Zhang, Lijun; Du, Yong

    2016-04-01

    We realized a three-dimensional visualization of the morphology evolution and the growth behavior of the octahedral primary silicon in hypereutectic Al-20wtpctSi alloy during solidification in a real length scale by utilizing the phase-field simulation coupled with CALPHAD databases, and supported by key experiments. Moreover, through two-dimensional cut of the octahedral primary silicon at random angles, different morphologies observed in experiments, including triangle, square, trapezoid, rhombic, pentagon, and hexagon, were well reproduced.

  13. 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.

  14. A reduced chemical kinetic model for IC engine combustion simulations with primary reference fuels

    SciTech Connect

    Ra, Youngchul; Reitz, Rolf D.

    2008-12-15

    A reduced chemical kinetic mechanism for the oxidation of primary reference fuel (PRF) has been developed and applied to model internal combustion engines. Starting from an existing reduced reaction mechanism for n-heptane oxidation, a new reduced n-heptane mechanism was generated by including an additional five species and their relevant reactions, by updating the reaction rate constants of several reactions pertaining to oxidation of carbon monoxide and hydrogen, and by optimizing reaction rate constants of selected reactions. Using a similar approach, a reduced mechanism for iso-octane oxidation was built and combined with the n-heptane mechanism to form a PRF mechanism. The final version of the PRF mechanism consists of 41 species and 130 reactions. Validation of the present PRF mechanism was performed with measurements from shock tube tests, and HCCI and direct injection engine experiments available in the literature. The results show that the present PRF mechanism gives reliable performance for combustion predictions, as well as computational efficiency improvements for multidimensional CFD simulations. (author)

  15. 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

  16. Measurement methods for surface oxides on SUS 316L in simulated light water reactor coolant environments using synchrotron XRD and XRF

    NASA Astrophysics Data System (ADS)

    Watanabe, Masashi; Yonezawa, Toshio; Shobu, Takahisa; Shoji, Tetsuo

    2013-03-01

    Synchrotron X-ray diffraction (XRD) and X-ray fluorescent (XRF) measurement techniques have been used for non-destructive characterization of surface oxide films on Type 316L austenitic stainless steels that were exposed to simulated primary water environments of pressurized water reactors (PWR) and boiling water reactors (BWR). The layer structures of the surface spinel oxides were revealed ex situ after oxidation by measurements made as a function of depth. The layer structure of spinel oxides formed in simulated PWR primary water should normally be different from that formed in simulated BWR water. After oxidation in the simulated BWR environment, the spinel oxide was observed to contain NiFe2O4 at shallow depths, and FeCr2O4 and Fe3O4 at deeper depths. By contrast, after oxidation in the simulated PWR primary water environment, a Fe3O4 type spinel was observed near the surface and FeCr2O4 type spinel near the interface with the metal substrate. Furthermore, by in situ measurements during oxidation in the simulated BWR environment, it was also demonstrated that the ratio between spinel and hematite Fe2O3 can be changed depending on the water condition such as BWR normal water chemistry or BWR hydrogen water chemistry.

  17. 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

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

    NASA Astrophysics Data System (ADS)

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

    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

  19. Ultrasonic Backscattering in Polycrystalline Materials of Pwr Components

    NASA Astrophysics Data System (ADS)

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

    2011-06-01

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

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

    SciTech Connect

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

    2009-01-01

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

  1. 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.

  2. Improvements in the simulation of a main steam line break with steam generator tube rupture

    NASA Astrophysics Data System (ADS)

    Gallardo, Sergio; Querol, Andrea; Verdú, Gumersindo

    2014-06-01

    The result of simultaneous Main Steam Line Break (MSLB) and a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) is a depressurization in the secondary and primary system because both systems are connected through the SGTR. The OECD/NEA ROSA-2 Test 5 performed in the Large Scale Test Facility (LSTF) reproduces these simultaneous breaks in a Pressurized Water Reactor (PWR). A simulation of this Test 5 was made with the thermal-hydraulic code TRACE5. Some discrepancies found, such as an underestimation of SG-A secondary pressure during the depressurization and overestimation of the primary pressure drop after the first Power Operated Relief Valve (PORV) opening can be improved increasing the nodalization of the Upper Head in the pressure vessel and meeting the actual fluid conditions of Upper Head during the transient.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. 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.

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

    DOE PAGESBeta

    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 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

  9. 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.

  10. 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.

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

    SciTech Connect

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

    1993-05-01

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

  12. 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.

  13. 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

  14. PSTAR: Primary and secondary terms analysis and renormalization: A unified approach to building energy simulations and short-term monitoring

    SciTech Connect

    Subbarao, K.

    1988-09-01

    This report presents a unified method of hourly simulation of a building and analysis of performance data. The method is called Primary and Secondary Terms Analysis and Renormalization (PSTAR). In the PSTAR method, renormalized parameters are introduced for the primary terms such that the renormalized energy balance equation is best satisfied in the least squares sense, hence, the name PSTAR. PSTAR allows extraction of building characteristics from short-term tests on a small number of data channels. These can be used for long-term performance prediction (''ratings''), diagnostics, and control of heating, ventilating, and air conditioning systems (HVAC), comparison of design versus actual performance, etc. By combining realistic building models, simple test procedures, and analysis involving linear equations, PSTAR provides a powerful tool for analyzing building energy as well as testing and monitoring. It forms the basis for the Short-Term Energy Monitoring (STEM) project at SERI.

  15. [Primary hepatic lymphoma of MALT-type: a tumor that can simulate a liver metastasis].

    PubMed

    Chatelain, D; Maes, C; Yzet, T; Brevet, M; Bounicaud, D; Plachot, J-P; Verhaeghe, P

    2006-02-01

    Primary hepatic lymphomas are rare tumors. We report a case of a 72 year-old woman with a past history of colonic adenocarcinoma who presented primary hepatic lymphoma of MALT-type. The patient had been operated on 3 years before for colonic adenocarcinoma, pT3N0, revealed by a bowel obstructive syndrome. She had been treated by chemotherapy for 6 months. During the follow-up, the computed tomography-scan (CT-scan) revealed the presence of a not well-demarcated mass in segment III of the liver, measuring 4 cm in diameter. The tumor was hypodense and was not enhanced on dynamic study. The mass was already present on the initial CT-scan. Left lobectomy was performed with the diagnosis of liver metastasis of the colonic adenocarcinoma. Surgical specimen showed a tumor composed of a dense infiltrate of small lymphocytes positive for B-cell markers on immunohistochemistry. The tumor contained reactive lymphoid follicles and there were numerous lympho-epithelial biliary lesions. The patient is alive and free of disease 2 years after the diagnosis. Primary hepatic lymphoma of MALT-type is a low-grade B cell lymphoma. Twenty-five cases had been reported in the literature so far. The patients were 16 females and 9 males, mean age 63.5 years. The pathogenesis is still unclear but half of the patients had a past history of chronic inflammatory liver disease (hepatitis B or C virus infection, ascaris infection, primary biliary cirrhosis) or malignant neoplasm. This tumor has a good prognosis; it is usually limited to the liver and surgical resection cures the patient in most cases. PMID:16246295

  16. Lab on a chip-based hepatic sinusoidal system simulator for optimal primary hepatocyte culture.

    PubMed

    Choi, Yoon Young; Kim, Jaehyung; Lee, Sang-Hoon; Kim, Dong-Sik

    2016-08-01

    Primary hepatocyte cultures have been used in studies on liver disease, physiology, and pharmacology. While they are an important tool for in vitro liver studies, maintaining liver-specific characteristics of hepatocytes in vitro is difficult, as these cells rapidly lose their unique characteristics and functions. Portal flow is an important condition to preserve primary hepatocyte functions and liver regeneration in vivo. We have developed a microfluidic chip that does not require bulky peripheral devices or an external power source to investigate the relationship between hepatocyte functional maintenance and flow rates. In our culture system, two types of microfluidic devices were used as scaffolds: a monolayer- and a concave chamber-based device. Under flow conditions, our chips improved albumin and urea secretion rates after 13 days compared to that of the static chips. Reverse transcription polymerase chain reaction demonstrated that hepatocyte-specific gene expression was significantly higher at 13 days under flow conditions than when using static chips. For both two-dimensional and three-dimensional culture on the chips, flow resulted in the best performance of the hepatocyte culture in vitro. We demonstrated that flow improves the viability and efficiency of long-term culture of primary hepatocytes and plays a key role in hepatocyte function. These results suggest that this flow system has the potential for long-term hepatocyte cultures as well as a technique for three-dimensional culture. PMID:27334878

  17. Improving the oxidation resistance of 316L stainless steel in simulated pressurized water reactor primary water by electropolishing treatment

    NASA Astrophysics Data System (ADS)

    Han, Guangdong; Lu, Zhanpeng; Ru, Xiangkun; Chen, Junjie; Xiao, Qian; Tian, Yongwu

    2015-12-01

    The oxidation behavior of 316L stainless steel specimens after emery paper grounding, mechanical polishing, and electropolishing were investigated in simulated pressurized water reactor primary water at 310 °C for 120 and 500 h. Electropolishing afforded improved oxidation resistance especially during the early immersion stages. Duplex oxide films comprising a coarse Fe-rich outer layer and a fine Cr-rich inner layer formed on all specimens after 500 h of immersion. Only a compact layer was observed on the electropolished specimen after 120 h of immersion. The enrichment of chromium in the electropolished layer contributed to the passivity and protectiveness of the specimen.

  18. 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…

  19. Sensitivity of global terrestrial gross primary production to hydrologic states simulated by the Community Land Model using two runoff parameterizations

    NASA Astrophysics Data System (ADS)

    Lei, Huimin; Huang, Maoyi; Leung, L. Ruby; Yang, Dawen; Shi, Xiaoying; Mao, Jiafu; Hayes, Daniel J.; Schwalm, Christopher R.; Wei, Yaxing; Liu, Shishi

    2014-09-01

    Soil moisture plays an important role in the coupled water, energy, and carbon cycles. In addition to surface processes such as evapotranspiration, the boundary fluxes that influence soil moisture are closely related to surface or subsurface runoff. To elucidate how uncertainties in representing surface and subsurface hydrology may influence simulations of the carbon cycle, numerical experiments were performed using version 4 of the Community Land Model with two widely adopted runoff generation parameterizations from the TOPMODEL and Variable Infiltration Capacity (VIC) model under the same protocol. The results showed that differences in the runoff generation schemes caused a relative difference of 36% and 34% in global mean total runoff and soil moisture, respectively, with substantial differences in their spatial distribution and seasonal variability. Changes in the simulated gross primary production (GPP) were found to correlate well with changes in soil moisture through its effects on leaf photosynthesis (An) and leaf area index (LAI), which are the two dominant components determining GPP. Soil temperature, which is influenced by soil moisture, also affects LAI and GPP for the seasonal-deciduous and stress-deciduous plant functional types that dominate in cold regions. Consequently, the simulated global mean GPP differs by 20.4% as a result of differences in soil moisture and soil temperature simulated between the two models. 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.

  20. 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.

  1. Photochemical Reaction Dynamics of the Primary Event of Vision Studied by Means of a Hybrid Molecular Simulation

    PubMed Central

    Hayashi, Shigehiko; Tajkhorshid, Emad; Schulten, Klaus

    2009-01-01

    Abstract The photoisomerization reaction dynamics of a retinal chromophore in the visual receptor rhodopsin was investigated by means of hybrid quantum mechanical/molecular mechanical (QM/MM) molecular dynamics (MD) simulations. The photoisomerization reaction of retinal constitutes the primary step of vision and is known as one of the fastest reactions in nature. To elucidate the molecular mechanism of the high efficiency of the reaction, we carried out hybrid ab initio QM/MM MD simulations of the complete reaction process from the vertically excited state to the photoproduct via electronic transition in the entire chromophore-protein complex. An ensemble of reaction trajectories reveal that the excited-state dynamics is dynamically homogeneous and synchronous even in the presence of thermal fluctuation of the protein, giving rise to the very fast formation of the photoproduct. The synchronous nature of the reaction dynamics in rhodopsin is found to originate from weak perturbation of the protein surroundings and from dynamic regulation of volume-conserving motions of the chromophore. The simulations also provide a detailed view of time-dependent modulations of hydrogen-out-of-plane vibrations during the reaction process, and identify molecular motions underlying the experimentally observed dynamic spectral modulations. PMID:19167292

  2. 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.

  3. 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)

  4. 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

  5. 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

  6. 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.

  7. 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.

  8. [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. PMID:23667988

  9. Zebra: An advanced PWR lattice code

    SciTech Connect

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

    2012-07-01

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

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

    SciTech Connect

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

    2012-07-01

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

  11. Sensitivity of boreal forest regional water flux and net primary production simulations to sub-grid-scale land cover complexity

    NASA Astrophysics Data System (ADS)

    Kimball, J. S.; Running, S. W.; Saatchi, S. S.

    1999-11-01

    We use a general ecosystem process model (BIOME-BGC) coupled with remote sensing information to evaluate the sensitivity of boreal forest regional evapotranspiration (ET) and net primary production (NPP) to land cover spatial scale. Simulations were conducted over a 3 year period (1994-1996) at spatial scales ranging from 30 to 50 km within the BOREAS southern modeling subarea. Simulated fluxes were spatially complex, ranging from 0.1 to 3.9 Mg C ha-1 yr-1 and from 18 to 29 cm yr-1. Biomass and leaf area index heterogeneity predominantly controlled this complexity, while biophysical differences between deciduous and coniferous vegetation were of secondary importance. Spatial aggregation of land cover characteristics resulted in mean monthly NPP estimation bias from 25 to 48% (0.11-0.20 g C m-2 d-1) and annual estimation errors from 2 to 14% (0.04-0.31 Mg C ha-1 yr-1). Error was reduced at longer time intervals because coarse scale overestimation errors during spring were partially offset by underestimation of fine scale results during summer and winter. ET was relatively insensitive to land cover spatial scale with an average bias of less than 5% (0.04 kg m-2 d-1). Factors responsible for differences in scaling behavior between ET and NPP included compensating errors for ET calculations and boreal forest spatial and temporal NPP complexity. Careful consideration of landscape spatial and temporal heterogeneity is necessary to identify and mitigate potential error sources when using plot scale information to understand regional scale patterns. Remote sensing data integrated within an ecological process model framework provides an efficient mechanism to evaluate scaling behavior, interpret patterns in coarse resolution data, and identify appropriate scales of operation for various processes.

  12. 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)

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

    SciTech Connect

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

    1980-01-01

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

  14. 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)

  15. 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.

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

    DOE PAGESBeta

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; Wang, Jy-An John

    2016-06-10

    The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

  17. Timing analysis of PWR fuel pin failures

    SciTech Connect

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

    1992-09-01

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

  18. 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)

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

    SciTech Connect

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

    1980-01-01

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

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

    SciTech Connect

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

    1989-09-01

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

  1. A comparison of rating scale, secondary-task, physiological, and primary-task workload estimation techniques in a simulated flight task emphasizing communications load

    NASA Technical Reports Server (NTRS)

    Casali, J. G.; Wierwille, W. W.

    1983-01-01

    Sixteen potential metrics of pilot mental workload were investigated regarding their sensitivity to communication load and their intrusion on primary-task performance. A moving-base flight simulator was used to present three cross-country flights. The flights varied only in the difficulty of the communications requirements. Rating scale measures were obtained immediately postflight; all others were taken over a 7-min segment of the flight task. The results indicated that both the Modified Cooper-Harper Scale and the workload Multi-descriptor Scale were sensitive to changes in communications load. The secondary-task measure of time estimation and the physiological measure of pupil diameter were also sensitive. As expected, those primary-task measures that were direct measures of communicative performance were also sensitive to load, whereas aicraft control primary-task measures were not, attesting to the task specificity of such measures. Finally, the intrusion analysis revealed no differential interference between workload measures.

  2. Primary production response to seasonal-scale extremes in the Bering Sea simulated by the Community Earth System Model, version 1

    NASA Astrophysics Data System (ADS)

    Walston, J. M.; Gibson, G. A.; Walsh, J. E.

    2015-08-01

    The biological response to long-term trends and the co-occurrence of seasonal extremes of the physical environment and primary production in the eastern Bering Sea, as simulated by the Community Earth System Model (CESM1), are presented. This analysis covers the late-twentieth century (1950-2005) and focuses on critical drivers of the eastern Bering Sea ecosystem, including air temperature, sea ice area, wind mixing, and mixed layer depth. Primary production showed strong linear relationships to both air temperature and sea ice area during winter and spring. The only season that had a positive linear correspondence between wind mixing and primary production was summer. Over the fifty-five year period the CESM1 simulates a trend toward warmer air temperatures and a subsequent reduction in sea ice for every season; however, no trends were seen in seasonally averaged wind mixing or mixed layer depth. Corresponding to the air temperature increase was an increase in occurrence of positive seasonal extremes in primary production, as well as a reduction in negative production extremes. There were some instances of seasonal production extremes coinciding with seasonal extremes in the physical environment; however, neither these co-occurrences, nor the direction of the biological response to the physics, were robust throughout the study period.

  3. VISA: a computer code for predicting the probability of reactor pressure-vessel failure. [PWR

    SciTech Connect

    Stevens, D.L.; Simonen, F.A.; Strosnider, J. Jr.; Klecker, R.W.; Engel, D.W.; Johnson, K.I.

    1983-09-01

    The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NRC staff evaluation of pressurized thermal shock. VISA uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics are used to model crack initiation and propagation. parameters for initial crack size, copper content, initial RT/sub NDT/, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents the version of VISA used in the NRC staff report (Policy Issue from J.W. Dircks to NRC Commissioners, Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982, SECY-82-465) and includes a user's guide for the code.

  4. 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)

  5. Evaluation of the Empirical Piecewise Regression Model in Simulating GPP in the Northern Great Plains

    NASA Astrophysics Data System (ADS)

    Zhang, L.; Wylie, B. K.; Fosnight, E. A.

    2005-12-01

    For better understanding the carbon fluxes in the grassland ecosystems, an empirical piecewise regression (PWR) model was developed to estimate gross primary production (GPP) for the grassland ecosystems in the Northern Great Plains and Northern Kazakhstan. The PWR model spatially scales up the localized flux tower measurements across an ecoregion at 1-km resolution. In this study, we compared the PWR GPP and the MODIS GPP with five grassland flux tower measurements. Then we employed cross-validation to evaluate the PWR GPP values. We also compared PWR GPP and MODIS GPP for grasslands for the entire study area. Factors that may explain the spatial pattern of the GPP differences between the two models were explored using decision tree technique. The results indicated that the PWR modeling approach was robust with a good agreement (agreement coefficient d=0.71-0.97) between PWR model and tower measurements. Cross-validation showed a relatively low agreement (d=0.71-0.78) at two influential flux tower sites. We also observed that the PWR GPP was lower than or similar to the MODIS GPP in the east and higher in the west and south. Further analysis suggested that percentage of C4 grasses, soil water holding capacity, percentage of clay, and percentage of crop mixed in the grassland contributed to the GPP difference of the PWR and MODIS models.

  6. Advanced TEM characterization of stress corrosion cracking of Alloy 600 in pressurized water reactor primary water environment

    NASA Astrophysics Data System (ADS)

    Sennour, M.; Laghoutaris, P.; Guerre, C.; Molins, R.

    2009-09-01

    Advanced transmission electron microscopy techniques were carried out in order to investigate stress corrosion cracking in Alloy 600 U-bend samples exposed in simulated PWR primary water at 330 °C. Using high-resolution imaging and fine-probe chemical analysis methods, ultrafine size oxides present inside cracks and intergranular attacks were nanoscale characterized. Results revealed predominance of Cr 2O 3 oxide and Ni-rich metal zones at the majority of encountered crack tip areas and at leading edge of intergranular attacks. However, NiO-structure oxide was predominant far from crack tip zones and within cracks propagating along twin boundaries and inside grains. These observations permit to suggest a mechanism for intergranular stress corrosion cracking of Alloy 600 in PWR primary water. Indeed, the results suggest that stress corrosion cracking is depending on chromium oxide growth in the grain boundary. Oxide growth seems to be dependent on oxygen diffusion in porous oxide and chromium diffusion in strained alloy and in grain boundary beyond crack tip. Strain could promote transport kinetic and oxide formation by increasing defaults rate like dislocations.

  7. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect

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

    2004-05-15

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

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

    SciTech Connect

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

    2013-07-01

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

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

    SciTech Connect

    Nagata, Akito; Sekimoto, Hiroshi

    2007-07-01

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

  10. 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.

  11. Simulation of the tensile properties of silica aerogels: the effects of cluster structure and primary particle size.

    PubMed

    Liu, Qiang; Lu, Zixing; Zhu, Man; Yuan, Zeshuai; Yang, Zhenyu; Hu, Zijun; Li, Junning

    2014-09-01

    A new two-level model is proposed to investigate the relationship between the mechanical properties and microstructure of silica aerogels. This two-level model consists of the particle-particle interaction model and the cluster structure model. The particle-particle interaction model is proposed to describe interactions between primary particles, in which the polymerization reaction between primary particles is considered. The cluster structure model represents the geometrical structure of silica aerogels, and it is established using a modified diffusion-limited colloid aggregation (DLCA) algorithm. This two-level model is used to investigate the tensile behavior of silica aerogels based on the discrete element method (DEM). The numerical results show that the primary particle size has significant effects on the elastic modulus and tensile strength of silica aerogels. Moreover, the power-law relationships between tensile properties and aerogel density are dependent on the variation of the primary particle radius with density. The present results can explain the difference among different experimental exponents to a certain extent. In comparison with experimental data within a wide density range, this two-level model provides good quantitative estimations of the elastic modulus and tensile strength of silica aerogels after the size effects of the primary particle are considered. This paper provides a fundamental understanding of the relationship between the mechanical properties and microstructure of silica aerogels. The two-level model can be extended to study the mechanical properties of other aerogels and aerogel composites. PMID:25022232

  12. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

    SciTech Connect

    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter; Bertholdt, Horst-Otto; Adams, Andreas; Impertro, Michael; Roesch, Josef

    2013-07-01

    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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. PMID:26492539

  18. Design and simulation of the active support system for a 1.2m meniscus primary mirror

    NASA Astrophysics Data System (ADS)

    Liu, Haitao; Fan, Bin; Zeng, Zhige; Li, Xiaojin; Wang, Hongqiao; Liu, Rong

    2015-07-01

    Thin meniscus primary mirrors with active support have been used successfully in many large telescopes, and also draw attention of many optical fabricators. Because the active support system can correct the low order figure errors, such as astigmatism, coma, trefoil 3rd astigmatism, the optical fabricators can just focus on to remove high order figure errors. This will shorten the fabrication time. In this paper, we present an active support system for a 1.2m meniscus parabolic primary mirror. It contains 37 axial push-pull force supports, 3 axial fixed points, and 4 lateral restraints. Some basic performance of the active support system is analyzed and the figure error correction capability is also studied based on Zernike modes.

  19. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect

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

    1997-12-01

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

  20. Transient cooldown in a model cold leg and downcomer. [PWR

    SciTech Connect

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

    1983-05-01

    This report describes an experimental program of fluid mixing experiments performed at atmospheric pressure in a 1/5-scale, transparent model of a cold leg, downcomer, lower plenum, pump simulator and loop seal typical of Westinghouse and Combustion Engineering Pressurized Water Reactor (PWRs). The tests were transient cooldown tests in that they simulated an extreme condition of Small Break Loss of Coolant Accident (SBLOCA) during which cold High Pressure Injection (HPI) fluid is injected into stagnant, hot, primary fluid with complete loss of natural circulation in the loop. Cooldown in this new test series is much slower than in previous tests that did not model the pump simulator and loop seal volumes. For the stagnant loop condition, the dominant buoyancy force diverts cool HPI water to the additional volumes.

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

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

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

  2. 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.

  3. 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.

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

    SciTech Connect

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

    2012-07-01

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

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

    SciTech Connect

    Bates, J.M.

    1986-01-01

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

  6. 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.

  7. 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.

  8. 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. PMID:27243386

  9. 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)

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

    SciTech Connect

    Maucec, Marko; Glumac, Bogdan

    2005-01-15

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

  11. 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).

  12. Edaphic and climatic effects on forest stand development, net primary production, and net ecosystem productivity simulated for Coastal Plain loblolly pine in Virginia

    NASA Astrophysics Data System (ADS)

    Sampson, D. A.; Wynne, R. H.; Seiler, J. R.

    2008-03-01

    We used SECRETS-3PG to simulate net primary production (NPP) and net ecosystem productivity (NEP) of loblolly pine (Pinus taeda L.) growing on the Virginia Coastal Plain, focusing on the effects of soils and climate, and stand age over a 30-year rotation. Soil type was influential, with heavier soils having greater NEP earlier in the rotation than lighter, sandier soils, although these differences disappeared by the rotation end. Climate had only a small effect. Stand age had the largest effect, with simulated annual NEP strongly negative during the first 5 to 8 years of development but peaking at +600 g C m-2 a-1 by age 13. Modest declines in NEP after 13 years were associated with declines in LAI as stands aged. The 30-year mean annual NEP was positive over most of the study area but in a few cases was indistinguishable from zero for northwestern portions of the study. Simulated annual NPP rose from zero to over 2300 g biomass m-2 a-1 by age 12, after which it declined to ˜1700 g biomass m-2 a-1 by rotation end. These results suggest that loblolly pine plantations on the Coastal Plain of Virginia may become net annual C sinks 5 to 9 years after planting but that when averaged over a whole rotation the net carbon accumulation during the baseline rotation simulated here is indistinguishable from zero. Our results also suggest, however, that this finding is sensitive to the length of the rotation, soil type (and thus fertility), and climate, implying that changes in management practices could significantly influence the carbon balance in managed loblolly pine plantations.

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

    SciTech Connect

    Pottorf, J.; Bajorek, S.M.

    2002-07-01

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

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

    Energy Science and Technology Software Center (ESTSC)

    2013-11-01

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

  15. 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

  16. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect

    J.M. Scaglione

    2004-12-17

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

  17. Electropolishing process development for PWR steam generator channel heads

    SciTech Connect

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

    1991-04-01

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

  18. 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. PMID:16395791

  19. Receptionist rECognition and rEferral of Patients with Stroke (RECEPTS): unannounced simulated patient telephone call study in primary care

    PubMed Central

    Mellor, Ruth M; Sheppard, James P; Bates, Elizabeth; Bouliotis, George; Jones, Janet; Singh, Satinder; Skelton, John; Wiskin, Connie; McManus, Richard J

    2015-01-01

    Background Stroke is a leading cause of morbidity and mortality. Timely recognition and referral are essential for treatment. Aim To examine the ability of receptionists in general practices to recognise symptoms of stroke and direct patients to emergency care. Design and setting Unannounced simulated patient telephone calls and prospective cross-sectional survey study in general practices in the Birmingham and Solihull area. Method A total of 52 general practices participated in a total of 520 simulated telephone calls, with 183 receptionists completing questionnaires. Logistic regression analyses were used to examine likelihood of referral for immediate care by ease of vignette recognition and number of common stroke symptoms present. Results General practice receptionists correctly referred 69% of simulated calls for immediate care. Calls classed as ‘difficult’ to recognise were less likely to be immediately referred. Compared with ‘easy’ calls: ‘difficult’ calls odds ratio (OR) 0.15, 95% confidence interval (CI) = 0.08 to 0.26; ‘moderate’ calls OR 0.55, 95% CI = 0.32 to 0.92. Similarly, calls including one or two ‘FAST’ symptoms were less likely to be referred immediately (compared with three FAST symptoms: one symptom OR 0.30, 95% CI = 0.13 to 0.72; two symptoms OR 0.35, 95% CI = 0.15 to 0.83). Conclusion General practice receptionists refer patients with stroke for immediate care when they present with several symptoms; however, they are less likely to refer patients presenting with only one symptom or less common symptoms of stroke. Optimum management of acute stroke in primary care requires interventions that improve receptionists’ knowledge of lesser-known stroke symptoms. PMID:26120134

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

    SciTech Connect

    J.S. Tang

    2001-05-03

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

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

    SciTech Connect

    King, D.B.

    1989-07-01

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

  2. 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.

  3. 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.

  4. 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.

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

    NASA Astrophysics Data System (ADS)

    Charles, M.; Lemaignan, C.

    1992-06-01

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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

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

    SciTech Connect

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

    2001-06-17

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

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

    SciTech Connect

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

    1982-12-01

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

  12. 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.

  13. 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.

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

    SciTech Connect

    M.Krug, R.Shogan

    2004-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2008-04-01

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

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

    SciTech Connect

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

    1980-01-01

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

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

    SciTech Connect

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

    1982-10-01

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

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

    SciTech Connect

    Hartini, Entin Andiwijayakusuma, Dinan

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  19. 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.

  20. 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

  1. 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

  2. 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)

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

    SciTech Connect

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

    1982-05-01

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

  4. 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

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

    SciTech Connect

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

    1984-06-01

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

  6. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

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

    1982-01-01

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

  7. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

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

    1982-01-01

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

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

    SciTech Connect

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

    2006-07-01

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

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

    NASA Astrophysics Data System (ADS)

    Luo, Shanggeng; Sheng, Jiawei; Tang, Baolong

    2001-09-01

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

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

    SciTech Connect

    M. E. Krug; R. P. Shogan

    2005-09-30

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

  11. International experience with a multidisciplinary table top exercise for response to a PWR accident

    SciTech Connect

    Lakey, J.R.A.

    1996-06-01

    Table Top Exercises are used for the training of emergency response personnel from a wide range of disciplines whose duties range from strategic to tactical, from managerial to operational. The exercise reported in this paper simulates the first two or three hours of an imaginary accident on a generic PWR site (named Seaside or Lakeside depending on its location). It is designed to exercise the early response of staff of the utility, government, local authority and the media and some players represent the public. The relatively few scenarios used for this exercise are based on actual events scaled to give off-site consequences which demand early assessment and therefore stress the communication procedures. The exercise is applicable in different cultures and has been used in over 20 short courses held in the USA, UK, Sweden, Prague, and Hong Kong. There are two styles of support for players: a linear program which ensures that all players follow the desired path through the event and an open program which is triggered by umpires (who play the reactor crew from a script) and by requests from other players. In both cases the exercise ends with a Press Conference. Players have an initial briefing and are assigned to roles; those who must speak at interviews and at the Press Conference arc given separate briefing by an expert in Public Affairs. The exercise runs with up to six groups and the communication rate reaches about 30 to 40 messages per hour for each group. The exercise can be applied to test management and communication systems and to study human response to emergencies because the merits of individual players are highlighted in the relatively stressful conditions of the initial stage of an accident. For some players the exercise is the first time that they have been required to carry out their task in front of other people.

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

    SciTech Connect

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

    1993-06-01

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

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

    SciTech Connect

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

    2010-06-22

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

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

    NASA Astrophysics Data System (ADS)

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

    2010-06-01

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

  15. Syphilis - primary

    MedlinePlus

    Primary syphilis; Secondary syphilis; Late syphilis; Tertiary syphilis ... Syphilis is a sexually transmitted, infectious disease caused by the spirochete bacterium Treponema pallidum . This bacterium causes ...

  16. Hydrothermal synthesis of Ni 2FeBO 5 in near-supercritical PWR coolant and possible effects of neutron-induced 10B fission in fuel crud

    NASA Astrophysics Data System (ADS)

    Sawicki, Jerzy A.

    2011-08-01

    The hydrothermal synthesis of a nickel-iron oxyborate, Ni 2FeBO 5, known as bonaccordite, was investigated at pressures and temperatures that might occur at the surface of high-power fuel rods in PWR cores and in supercritical water reactors, especially during localized departures from nucleate boiling and dry-outs. The tests were performed using aqueous mixtures of nickel and iron oxides with boric acid or boron oxide, and as a function of lithium hydroxide addition, temperature and time of heating. At subcritical temperatures nickel ferrite NiFe 2O 4 was always the primary reaction product. High yield of Ni 2FeBO 5 synthesis started near critical water temperature and was strongly promoted by additions of LiOH up to Li/Fe and Li/B molar ratios in a range 0.1-1. The synthesis of bonaccordite was also promoted by other alkalis such as NaOH and KOH. The bonaccordite particles were likely formed by dissolution and re-crystallization by means of an intermediate nickel ferrite phase. It is postulated that the formation of Ni 2FeBO 5 in deposits of borated nickel and iron oxides on PWR fuel cladding can be accelerated by lithium produced in thermal neutron capture 10B(n,α) 7Li reactions. The process may also be aided in the reactor core by kinetic energy of α-particles and 7Li ions dissipated in the crud layer.

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

    SciTech Connect

    Grenouillet, Jean-Jacques

    2007-07-01

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

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

    SciTech Connect

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

    1994-07-01

    Reinforced concrete, prestressed concrete, and freestanding steel PWR containment structures and components have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits, inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these structures and components can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR containment structures and components for license renewal.

  19. LOCA hydroloads calculations with multidimensional nonlinear fluid/structure interaction. Volume 1: STEALTH 1D single-phase fluid studies. Final report. [PWR

    SciTech Connect

    Santee, G.E. Jr.; Mortensen, G.A.; Caraher, D.L.

    1980-04-01

    This report, which is the first in a series of reports for RP-1065, describes the first step in the stepwise approach for developing the methodology to assess the hydroloads on a large PWR during the subcooled portions of a hypothetical LOCA. The first step in the methodology considers enhancements and special modifications to the 1D STEALTH computer code in order that acoustic phenomena in piping and vessel networks may be simulated. The resulting code is termed 1D STEALTH-HYDRO. The 1D STEALTH-HYDRO enhancements consist of three control volume models to simulate area changes, orifices, and tees in piping networks. The theory of the control volume models is described.

  20. 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)

  1. 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. PMID:26047859

  2. Primary thrombocythemia

    MedlinePlus

    ... in which the bone marrow produces too many platelets. Platelets are a part of the blood that aids ... Primary thrombocythemia is caused by the overproduction of platelets. If untreated, this condition gets worse over time. ...

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

    SciTech Connect

    Chung, H. M.; Energy Technology

    2006-01-31

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

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

    SciTech Connect

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

    2013-07-01

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

  5. 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.

  6. LWR (Light Water Reactor) power plant simulations using the AD10 and AD100 systems

    SciTech Connect

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.; Institute of Nuclear Energy Research, Lung-Tan; Tawian Power Co., Taipei; Brookhaven National Lab., Upton, NY; Institute of Nuclear Energy Research, Lung-Tan )

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab.

  7. 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

  8. Elastic Vibrations in the Photosynthetic Bacterial Reaction Center Coupled to the Primary Charge Separation: Implications from Molecular Dynamics Simulations and Stochastic Langevin Approach.

    PubMed

    Milanovsky, Georgy E; Shuvalov, Vladimir A; Semenov, Alexey Yu; Cherepanov, Dmitry A

    2015-10-29

    Primary electron transfer reactions in the bacterial reaction center are difficult for theoretical explication: the reaction kinetics, almost unalterable over a wide range of temperature and free energy changes, revealed oscillatory features observed initially by Shuvalov and coauthors (1997, 2002). Here the reaction mechanism was studied by molecular dynamics and analyzed within a phenomenological Langevin approach. The spectral function of polarization around the bacteriochlorophyll special pair PLPM and the dielectric response upon the formation of PL(+)PM(-) dipole within the special pair were calculated. The system response was approximated by Langevin oscillators; the respective frequencies, friction, and energy coupling coefficients were determined. The protein dynamics around PL and PM were distinctly asymmetric. The polarization around PL included slow modes with the frequency 30-80 cm(-1) and the total amplitude of 130 mV. Two main low-frequency modes of protein response around PM had frequencies of 95 and 155 cm(-1) and the total amplitude of 30 mV. In addition, a slowly damping mode with the frequency of 118 cm(-1) and the damping time >1.1 ps was coupled to the formation of PL(+)PM(-) dipole. It was attributed to elastic vibrations of α-helices in the vicinity of PLPM. The proposed trapping of P excitation energy in the form of the elastic vibrations can rationalize the observed properties of the primary electron transfer reactions, namely, the unusual temperature and ΔG dependences, the oscillating phenomena in kinetics, and the asymmetry of the charge separation reactions. PMID:26148224

  9. Self-assembly of tetramers of 5,6-dihydroxyindole explains the primary physical properties of eumelanin: experiment, simulation, and design.

    PubMed

    Chen, Chun-Teh; Ball, Vincent; de Almeida Gracio, José Joaquim; Singh, Manoj Kumar; Toniazzo, Valérie; Ruch, David; Buehler, Markus J

    2013-02-26

    Eumelanin is a ubiquitous pigment in nature and has many intriguing physicochemical properties, such as broad-band and monotonous absorption spectrum, antioxidant and free radical scavenging behavior, and strong nonradiative relaxation of photoexcited electronic states. These properties are highly related to its structural and mechanical properties and make eumelanin a fascinating candidate for the design of multifunctional nanomaterials. Here we report joint experimental-computational investigation of the structural and mechanical properties of eumelanin assemblies produced from dopamine, revealing that the mass density of dry eumelanin is 1.55 g/cm³ and its Young's modulus is ≈5 GPa. We also find that wet eumelanin has a lower mass density and Young's modulus depending on the water-to-melanin ratio. Most importantly, our data show that eumelanin molecules tend to form secondary structures based on noncovalent π stacking in both dry and wet conditions, with an interlayer distance between eumelanin molecules of 3.3 Å. Corresponding transmission electron microscope images confirm the supramolecular organization predicted in our simulations. Our simulations show that eumelanin is an isotropic material at a larger scale when eumelanin molecules are randomly oriented to form secondary structures. These results are in good agreement with experimental observations, density functional theory calculations, and bridge the gap between earlier experimental and small-scale quantum mechanical studies of eumelanin. We use the knowledge acquired from the simulations to select a partner molecule, a cationic phthalocyanine, allowing us to produce layer-by-layer films containing eumelanin that display an electrical conductivity 5 orders of magnitudes higher than that of pure eumelanin films. PMID:23320483

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

    SciTech Connect

    1986-03-01

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

  11. Simulation of loss-of-decay heat removal with MAAP 3. 0B

    SciTech Connect

    Plys, M.G.; Paik, C.Y.

    1989-01-01

    Loss-of-decay heat removal (DHR) during nonpower operation and its potential consequences have been of increasing concern to US pressurized water reactor (PWR) owners and regulators. Regulatory action prompted by the April 10, 1987, loss of DHR at Diablo Canyon Unit 2 cited 37 instances of loss of DHR events attributed to inadequate reactor coolant system (RCS) water level. In the Diablo Canyon event, 7 days after shutdown for its first refueling outage, boiling and pressurization of the primary system began after {approximately}1 h after loss of DHR, which was reestablished after another 20 min. This paper describes the modification and application of the Modular Accident Analysis Program (MAAP) version 3.0B (Ref. 3) for simulation of loss of DHR scenarios. MAAP 30.B, maintained by the Electric Power Research Institute and available to most US and many foreign utilities, is designed to perform coupled thermal-hydraulic and fission product calculations for severe-accident analysis. In this application, the capabilities of MAAP are extended to allow simulation of a shutdown plant state to calculate the time to core uncovery and fuel damage. Operator actions to mitigate or terminate the accident can be simulated, and timing of key events given by MAAP is useful for the evaluation of procedures.

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

    NASA Astrophysics Data System (ADS)

    LaFleur, Adrienne M.; Menlove, Howard O.

    2015-05-01

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

  13. 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.

  14. LOCA hydroloads calculations with multidimensional nonlinear fluid/structure interaction. Volume 3. Fluid/structure interaction studies using 3-D STEALTH/WHAMSE. Final report. [PWR

    SciTech Connect

    Santee, G.E. Jr.; Chang, F.H.; Mortensen, G.A.; Brockett, G.F.; Gross, M.B.; Belytschko, T.B.

    1982-11-01

    This report, the third in a series of reports for RP-1065, describes the final step in the stepwise approach for developing the three-dimensional, nonlinear, fluid-structure interaction methodology to assess the hydroloads on a large PWR during the subcooled portions of a hypothetical LOCA. The final step in the methodology implements enhancements and special modifications to the STEALTH 3D computer program and the WHAMSE 3D computer program. After describing the enhancements, the individual and the coupled computer programs are assessed by comparing calculational results with either analytical solutions or with experimental data. The coupled 3D STEALTH/WHAMSE computer program is then applied to the simulation of HDR Test V31.1 to further assess the program and to investigate the role that fluid-structure interaction plays in the hydrodynamic loading of reactor internals during subcooled blowdown.

  15. 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

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

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

    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.

  17. Primary hyperparathyroidism

    PubMed Central

    Pallan, Shelley; Khan, Aliya

    2011-01-01

    Abstract Objective To discuss the presentation, diagnosis, and management of primary hyperparathyroidism (PHPT) in family medicine. Quality of evidence MEDLINE was searched from 2002 to 2009 using the terms presentation, diagnosis, and treatment of PHPT. Proceedings and guidelines from the Third International Workshop on Primary Hyperparathyroidism in May 2008 were reviewed in detail. Most studies offered level II and III evidence, although there were a number of single randomized controlled trials on PHPT (level I evidence). References from pertinent papers were also searched for relevant articles. Articles most relevant to family medicine and primary care practitioners are presented. Main message Primary hyperparathyroidism is the most common cause of hypercalcemia in outpatients. In the Western world, most patients with PHPT present with nonspecific symptoms such as fatigue, mood disturbances, and cognitive impairments. Diagnosis is established when intact parathyroid hormone levels are elevated or at the high end of the normal range in the setting of elevated total or ionized calcium levels (following exclusion of conditions that can mimic PHPT). Criteria for surgery have recently been modified. Surgery is always a suitable option in those with symptomatic PHPT and no contraindications. Those with contraindications or with asymptomatic PHPT not meeting the criteria for surgery can generally be safely monitored and considered for medical management. This might include treatment with bisphosphonates, hormone replacement therapy, raloxifene, or calcimimetic agents; however, there are currently no fracture data for any of these options. Conclusion The definitive therapy for symptomatic and asymptomatic PHPT is parathyroidectomy. In patients with asymptomatic PHPT not meeting the criteria for surgery, monitoring is safe and medical management designed to target skeletal protection or lower serum calcium is a suitable option. PMID:21321169

  18. 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.

  19. 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.

  20. 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)

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

    SciTech Connect

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

    2015-09-30

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

  2. 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.

  3. 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.

  4. 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.

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

    SciTech Connect

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

    1980-06-01

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

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

    SciTech Connect

    Willingham, C.E.

    1981-10-01

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

  7. 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.

  8. 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.

  9. 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

  10. 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)