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Sample records for 5x5 pwr rod

  1. Numerical evaluation of flow through a 5X5 PWR rod bundle: effect of the vane arrangement in a spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    Spacer grids along the fuel assembly of Pressurized Water Reactors (PWR) maintain rod bundles arranged in a regular square configuration. The mixing vanes present in the spacer grids promote cross and swirl flow between and within the subchannels, enhancing the heat transfer performance in the grid vicinity, but also causing an adverse increase of the pressure drop in the rod bundle due the constriction on the coolant flow area. Therefore, the thermal hydraulic design of the grid must allow for both low pressure loss and high coolant mixing, which means it is important to optimize the design of the grid in relation to the mixing vane. More recently, Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently as an auxiliary tool in the development of spacer grids. The influence of some geometric characteristics of spacer grids on the flow through a rod bundle have been numerically evaluated and are still a subject of discussion. This work analyses the influence of the vanes arrangement in the spacer grid on the flow through a PWR 5 x 5 rod bundle segment. The Numerical simulations were performed with the commercial code CFX 11.0. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the k- turbulence model with scalable wall function was used. Five different vane arrangements were simulated at reactor level power and flow characteristics. The same grid and vane geometry were used in all simulations. The results of this study were divided in two parts. In the first part the presence of peripheral vanes on 5 x 5 rod bundle spacer grid tests were evaluated. The results showed that peripheral vanes should be avoided in experiments and simulations in order to

  2. Evaluation of a numeric procedure for flow simulation of a 5X5 PWR rod bundle with a mixing vane spacer

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5x5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the {kappa}-{epsilon} turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and

  3. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sungkew; Jang, Hyungwook; Lim, Jongseon; Park, Eungjun; Nahm, Keeyil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively.

  4. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  5. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  6. Axial gas flow in irradiated PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Dagbjartsson, S.J.; Murdock, B.A.; Owen, D.E.; MacDonald, P.E.

    1977-09-01

    Transient and steady state axial gas flow experiments were performed on six irradiated, commercial pressurized water reactor fuel rods at ambient temperature and 533 K. Laminar flow equations, as used in the FRAP-T2 and SSYST fuel behavior codes, were used with the gas flow results to calculate effective fuel rod radial gaps. The results of these analyses were compared with measured gap sizes obtained from metallographic examination of one fuel rod. Using measured gap sizes as input, the SSYST code was used to calculate pressure drops and mass fluxes and the results were compared with the experimental gas flow data.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    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)

  8. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail: andre@inb.gov.br; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail: adolpho@technotest.com.br

    2009-07-01

    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  9. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

  10. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  11. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  12. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

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

    Science.gov (United States)

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

    2014-04-01

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

  14. Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

    Energy Technology Data Exchange (ETDEWEB)

    Ferroukhi, H.; Vasiliev, A. [Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland); Dufresne, A. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Chawla, R. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Paul Scherrer Institut (Switzerland)

    2012-07-01

    Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Lithium and boron analysis by LA-ICP-MS results from a bowed PWR rod with contact

    Directory of Open Access Journals (Sweden)

    Puranen Anders

    2017-01-01

    Full Text Available A previously published investigation of an irradiated fuel rod from the Ringhals 2 PWR, which was bowed to contact with an adjacent rod, identified a significant but highly localised thinning of the clad wall and increased corrosion. Rod fretting was deemed unlikely due to the adhering oxide covering the surfaces. Local overheating in itself was also deemed insufficient to account for the accelerated corrosion. Instead, an enhanced concentration of lithium due to conditions of local boiling was hypothesised to explain the accelerated corrosion. Studsvik has developed a hot cell coupled LA-ICP-MS (Laser Ablation Inductively Coupled Plasma Mass Spectrometer equipment that enables a flexible means of isotopic analysis of irradiated fuel and other highly active surfaces. In this work, the equipment was used to investigate the distribution of lithium (7Li and boron (11B in the outer oxide at the bow contact area. Depth profiling in the clad oxide at the opposite side of the rod to the point of contact, which is considered to have experienced normal operating conditions and which has a typical oxide thickness, evidenced levels of ∼10–20 ppm 7Li and a 11B content reaching hundreds of ppm in the outer parts of the oxide, largely in agreement with the expected range of Li and B clad oxide concentrations from previous studies. In the contact area, the 11B content was similar to the reference condition at the opposite side. The 7Li content in the outermost oxide closest to the contact was, however, found to be strongly elevated, reaching several hundred ppm. The considerable and highly localised increase in lithium content at the area of enhanced corrosion thus offers strong evidence for a case of lithium induced breakaway corrosion during power operation, when rod-to-rod contact and high enough surface heat flux results in a very local increase in lithium concentration.

  18. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-06-01

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

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

    Science.gov (United States)

    Walter, Daniel John

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

  1. Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

  2. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  3. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  4. 压水堆驱动线落棒历程计算%Calculation of Drop Course of Control Rod Assembly in PWR

    Institute of Scientific and Technical Information of China (English)

    周肖佳; 毛飞; 闵鹏; 林绍萱

    2013-01-01

    控制棒落棒性能验证是核电厂安全分析的重要部分,研制驱动线落棒历程计算程序有利于验证和改进控制棒驱动线设计。基于驱动线结构特点,分析运动组件的受力情况并进行分解,选择理论或数值方法逐一求取各分力的瞬态值,从而建立驱动线落棒历程的循环步进计算程序。利用秦山核电二期工程驱动线落棒性能试验数据对理论模型和程序计算结果进行对比验证。结果证明:所建立的驱动线落棒历程计算程序适用于压水堆驱动线系统,能正确地对运动组件落棒受力与运动历程进行模拟。%The validation of control rod drop performance is an important part of safety analysis of nuclear power plant .Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line .Based on structural features of the drive line ,the driving force on moving assembly was analyzed and decomposed ,the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method , and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved .The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR .It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA .

  5. PWR-UO{sub 2} nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Sousa, R.V.; Pereira, C., E-mail: claubia@nuclear.ufmg.br; Silva, C.A.M.; Costa, A.L.; Veloso, M.A.F.; Oliveira, A.H. de

    2013-10-15

    Highlights: • A three-dimensional model of a PWR fuel were simulated. • Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. • Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. • Results show very good agreement with those published by OECD. -- Abstract: Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO{sub 2} nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants.

  6. Effects of sleeve blockages on axial velocity and intensity of turbulence in an unheated 7 x 7 rod bundle. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1976-01-01

    An experimental study is described which was performed to investigate the turbulent flow phenomena near postulated sleeve blockages in a model nuclear fuel rod bundle. The sleeve blockages were characteristic of fuel clad ''swelling'' or ''ballooning'' which could occur during loss-of-coolant accidents (LOCA) in pressurized water reactors. The study was conducted to provide information relative to the flow phenomena near postulated blockages to support detailed safety analyses of LOCAs. The results of the study are especially useful for verification of the hydraulic treatment of reactor core computer programs such as COBRA.

  7. Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation

    Directory of Open Access Journals (Sweden)

    C. Mesado

    2012-01-01

    Full Text Available In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods.

  8. Monitoring of garbage with a 5 x 5 NaI (Tl) detector; Monitoreo de basura con un detector de NaI (Tl) de 5 x 5

    Energy Technology Data Exchange (ETDEWEB)

    Cortes P, A.; Becerril V, A.; Angeles C, A

    1991-12-15

    So far in that is carried out the first reload of nuclear fuel in the LVC, the monitoring of garbage has been carried out using monitors trade mark Eberline model RM 14. The procedure consists in manually monitoring each object and to separate of the considered 'clean' garbage the objects considered as contaminated, which register greater or equal counts to 100 cpm. This way to process was adequate under normal operation conditions, but not in the operation rhythm that implies a bigger maintenance since the time required for monitoring from 5 to 10 kg. of garbage is of the order of 0.5 hours and the production rhythm of this it ends up being a lot but high. Due to this necessity it was thought about the problem of looking by a more efficient monitoring method. In this work a method that uses a detector of NaI (Tl) of 5 x 5 inches is discussed. (Author)

  9. A comparison of the CFD simulation results in 5 x 5 sub-channels with mixing grids using different turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Yang, L.X.; Zhou, M.J.; Chao, Y.M. [Beijing Jiaotong Univ. (China). School of Mechanical Electronic and Control Engineering

    2016-07-15

    We evaluated the performance of various turbulence models, including eddy viscosity models and Reynolds stress models, when analyzing rod bundles in fuel assemblies using the Computational Fluid Dynamics (CFD) method. The models were assessed by calculating the pressure drop and Nusselt numbers in 5 x 5 rod bundles using the CFD software ANSYS CFX. Comparisons between the numerical and experimental results, as well as the swirl factor, cross-flow factor, and turbulence intensity utilized to evaluate the swirling and cross-flow, were used to analyze the inner relationship between the flow field and heat transfer. These comparisons allow the selection of the most appropriate turbulence model for modeling flow features and heat transfer in rod bundles.

  10. 压水堆燃料棒在轴向流作用下的随机振动响应研究%Random Response Analysis of PWR Fuel Rod Effect on Axial Flow

    Institute of Scientific and Technical Information of China (English)

    黄恒; 刘彤; 周跃民

    2015-01-01

    Based on random vibration theory ,the random response analysis method of PWR fuel rods under axial flow was established .The fluid force along the axial of rod was treated as a fluctuant random load ,and the mode shape method and power spectrum analysis method were used to derive the empirical formula of RMS response .This article provides a theoretical analysis method w hich does not rely on the flow induced vibration test of fuel assembly .The effects for the RMS response of fuel rods by the equivalent velocity ,turbulence intensity ,and correlation length factor were discussed .The method can meet the requirements of engineering analysis . The results show that the RMS response of fuel rods will increase with the equivalent velocity ,turbulence intensity and the correlation length factor .The response is more sensitive to the equivalent velocity and coefficient length factor changes ,and linearly with the turbulence intensity .In the operating condition of the pressurized water reactor (PWR) ,the RMS amplitude of fuel rods is about micrometers .%基于随机振动理论,建立了在轴向流作用下压水堆燃料棒随机响应的纯理论分析方法。将流体力考虑为沿燃料棒轴向位置的脉冲随机荷载,结合模态分析技术,从功率谱分析法推导出燃料棒振动均方根响应的表达式。提供了一套不依赖燃料组件流致振动实验的纯理论分析方法,重点分析了等效流速、湍流强度、相关长度系数等几个主要流场参数对结构均方根响应的影响。结果表明,本文计算模型的精度满足工程分析要求,燃料棒响应随等效流速、湍流强度和相关长度系数的增大而增大;其中响应对于等效流速和相关长度系数的变化较为敏感,而与湍流强度呈线性变化关系;在压水堆运行中的燃料棒均方根幅值约处在μm量级。

  11. Study of heat transfer in a eccentric fuel rods in a non stop planned shutdown of a PWR type reactor; Estudo da transferencia de calor em uma vareta combustivel excentrica, num desligamento nao planejado de um reator do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: deisedy@gmail.com, E-mail: diogosb@outlook.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to conduct a case study in which the fuel pellets are displaced related to the center coating. Therefore, it will be addressed, first, the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, at a time later, you can use the program to know the fuel rod behavior and coolant channel.

  12. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  13. Decision DGSNR/SD2/no.95/2005 Anomalies of rod clusters insertion in EDF PWR reactors; Decision DGSNR/SD2/no.95/2005 Anomalies d'insertion des grappes de commande des reacteurs a eau sous pression d'EDF

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-03-15

    Following reports of lengthening of the drop time of control rods on some PWR, in particular, with the deformation of the fuel assemblies, the Authority of Nuclear Safety asked, at the end of 2002, the operators to implement provisions of prevention and monitoring. In particular, this decision forced to carry out a measurement of the drop time of the rod clusters and prohibited to reload under rod clusters the assemblies during the last irradiation cycle. Since 2002, fuel assemblies with reinforced structure are gradually introduced allowing the limitation of the deformation under irradiation and a total improvement of the drop time. In 2004 on the favor of this favorable experience feedback, the ASN reduced the requirement. This favorable evolution continued. By the decision GGSNR/SD2/no.95/2005, the ASN authorizes the operator to charge fuel assemblies under rods during the last irradiation cycle and ends the obligation to carry out tests of drop time. (A.L.B.)

  14. 压水堆控制棒价值误差分析%Error Analysis for PWR Control Rod Integrate Worth

    Institute of Scientific and Technical Information of China (English)

    付学峰; 王磊; 郑继业; 蔡德昌; 张洪; 李冬生

    2013-01-01

    During physical startup test,the control rod integrate worth is a parameter which is most likely to be overstepped.This paper analyzes the factors which contribute to the control rod worth error and its feature.Typical example is also included.To reduce the control rod error,it is strongly suggested to create an accurate reflector model,select fuel assembly manufacture data as input and consider the control rod worth measurement method and condition when performing the calculation.To identify the main reason of control rod worth error,flux mapping test results,boron concentration and other core parameters should be analyzed comprehensively,combining with the control rod worth error distribution characteristics.%压水堆启动物理试验时,控制棒价值是比较容易超限的一个参数.本文系统分析了影响控制棒价值计算值与测量值偏差的主要因素以及各因素的影响特点、大小,并给出了部分实例分析,以期降低控制棒价值的误差,减少因控制棒价值超差对启动物理试验带来的不利影响,并在控制棒价值超差原因分析时提供帮助.分析表明,为降低控制棒价值误差,需要建立精确、合理的反射层模型,尽可能采用燃料组件的制造参数,控制棒的计算方法要考虑试验方法与工况;将注量率图试验结果、硼浓度和其他堆芯参数与控制棒价值误差分布特点相结合,进行原因查找.

  15. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod; Contribution a la modelisation du comportement mecanique des combustibles REP sous irradiation, avec en particulier le traitement de l`interaction pastille-gaine dans un crayon combustible

    Energy Technology Data Exchange (ETDEWEB)

    Hourdequin, N.

    1995-05-01

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR`s operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends.

  16. Self-assembly of conductive Cu nanowires on Si(111)'5 x 5 '-Cu surface

    Energy Technology Data Exchange (ETDEWEB)

    Tsukanov, Dmitry A; Ryzhkova, Maria V; Gruznev, Dimitry V; Utas, Oleg A; Kotlyar, Vasily G; Zotov, Andrey V; Saranin, Alexander A [Institute of Automation and Control Processes, 690041 Vladivostok (Russian Federation)], E-mail: tsukanov@iacp.dvo.ru

    2008-06-18

    Upon room-temperature deposition onto a Cu/Si(111)'5 x 5' surface in ultra-high vacuum, Cu atoms migrate over extended distances to become trapped at the step edges, where they form Cu nanowires (NWs). The formed NWs are 20-80 nm wide, 1-3 nm high and characterized by a resistivity of {approx}8 {mu}{omega} cm. The surface conductance of the NW array is anisotropic, with the conductivity along the NWs being about three times greater than that in the perpendicular direction. Using a similar growth technique, not only the straight NWs but also other types of NW-based structures (e.g. nanorings) can be fabricated.

  17. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83 (NCEI Accession 0131858)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution bathymetric surface for an area surrounding the mouth of Salt River Bay (SARI)St....

  18. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution relative seafloor reflectivity surface for an area surrounding the mouth of Salt...

  19. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83 (NCEI Accession 0131858)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution relative seafloor reflectivity surface for an area surrounding the mouth of Salt...

  20. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution bathymetric surface for an area surrounding the mouth of Salt River Bay (SARI)St....

  1. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  2. Standard PWR for Italy

    Energy Technology Data Exchange (ETDEWEB)

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  3. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  4. Metallography and Microanalysis of Qinshan PhaseⅠ NPP Spent Fuel Rods

    Institute of Scientific and Technical Information of China (English)

    QIAN; Jin; BIAN; Wei; GUO; Li-na; GUO; Yi-fan; CHU; Feng-min; LIANG; Zheng-qiang

    2015-01-01

    Qinshan PhaseⅠNPP is a first domestic commercial PWR and its fuel rods and fuel assembly were designed and manufactured by China.In order to assess the irradiation properties of the fuel rods,8spent fuel rods which were drew out from 3fuel assemblies were transferred to CIAE hot cells for post irradiation examination(PIE)in 2014.The cladding material of the fuel

  5. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  6. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  7. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  8. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  9. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  10. Irradiation Effects Test Series: Test IE-2. Test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C. M.; Croucher, D. W.; Ploger, S. A.; Mehner, A. S.

    1977-08-01

    The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m/sup 2/. After one hour at 68 kW/m, a power-cooling-mismatch sequence was initiated by a flow reduction at constant power. At a flow of 2550 kg/s-m/sup 2/, the onset of film boiling occurred on one rod, Rod IE-011. An additional flow reduction to 2245 kg/s-m/sup 2/ caused the onset of film boiling on the remaining three rods. Data are presented on the behavior of fuel rods during quasiequilibrium and during film boiling operation. The effects of initial gap size, cladding irradiation, rod power cycling, a rapid power increase, and sustained film boiling are discussed. These discussions are based on measured test data, preliminary postirradiation examination results, and comparisons of results with FRAP-T3 computer model calculations.

  11. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  12. SCOR 1000: an economic and innovative conceptual design PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)

    2007-07-01

    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  13. Morphoelastic rods

    CERN Document Server

    Tiero, Alessandro

    2014-01-01

    We propose a mechanical theory describing elastic rods which, like plant organs, can grow and can change their intrinsic curvature and torsion. The equations ruling accretion and remodeling are obtained by combining balance laws involving non-standard forces with constitutive prescriptions filtered by a dissipation principle that takes into account both standard and non-standard working.

  14. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Silva, Jose Eduardo Rosa da; Lucki, Georgi; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: myrthes@ipen.br

    2007-07-01

    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  15. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Energy Technology Data Exchange (ETDEWEB)

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  16. CONTROL ROD

    Science.gov (United States)

    Zinn, W.H.; Ross, H.V.

    1958-11-18

    A control rod is described for a nuclear reactor. In certaln reactor designs it becomes desirable to use a control rod having great width but relatively llttle thickness. This patent is addressed to such a need. The neutron absorbing material is inserted in a triangular tube, leaving volds between the circular insert and the corners of the triangular tube. The material is positioned within the tube by the use of dummy spacers to achleve the desired absorption pattern, then the ends of the tubes are sealed with suitable plugs. The tubes may be welded or soldered together to form two flat surfaces of any desired width, and covered with sheetmetal to protect the tubes from damage. This design provides a control member that will not distort under the action of outside forces or be ruptured by gases generated within the jacketed control member.

  17. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Science.gov (United States)

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki

    1992-06-01

    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  18. Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng

    2012-01-01

    <正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.

  19. Analysis of Subchannel and Rod Bundle PSBT Experiments with CATHARE 3

    Directory of Open Access Journals (Sweden)

    M. Valette

    2012-01-01

    Full Text Available This paper presents the assessment of CATHARE 3 against PWR subchannel and rod bundle tests of the PSBT benchmark. Noticeable measurements were the following: void fraction in single subchannel and rod bundle, multiple liquid temperatures at subchannel exit in rod bundle, and DNB power and location in rod bundle. All these results were obtained both in steady and transient conditions. Void fraction values are satisfactory predicted by CATHARE 3 in single subchannels with the pipe module. More dispersed predictions of void values are obtained in rod bundles with the CATHARE 3 3D module at subchannel scale. Single-phase liquid mixing tests and DNB tests in rod bundle are also analyzed. After calibrating the mixing in liquid single phase with specific tests, DNB tests using void mixing give mitigated results, perhaps linked to inappropriate use of CHF lookup tables in such rod bundles with many spacers.

  20. Progress and prospects of nuclear fuel development in Japan, (2). Progress and future plan of research and development on PWR fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Yoshiaki; Abeta, Sadaaki; Aisu, Hideo; Teranishi, Tomoyuki

    1982-06-01

    13 years have elapsed since the first PWR plant started the operation in Japan, and at present, 11 PWR plants are in operation. During this period, much results of use and experience have been accumulated for the PWR fuel. The improvement and development of the fuel have been performed to meet the supply of the fuel sufficiently adaptable to the severe environment in Japan. In this paper, the evaluation of soundness and the improvement of reliability of PWR fuel made so far are reported, and the response of fuel side to long cycle operation and load following-up operation, which will be required in near future, is explained. The inspection of fuel has been performed at reactor sites for the purpose of sufficiently observing the irradiation behavior of fuel and detecting the points out of order. Effort has been exerted to perform various inspections thoroughly on total number of fuel and reflect the results to the improved design. Fuel leak scarcely occurred from the beginning, accordingly, improvement has been made to reduce the bending of fuel rods. The change of PWR fuel design, the evaluation of soundness and the improvement of reliability of PWR fuel, and the improvement for the future are reported.

  1. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR; Evaluacion de la presencia de un absorbedor quemable en un ensamble 3x3 tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico)]. e-mail: mike_ipn_esfm@hotmail.com

    2008-07-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO{sub 2}. (Author)

  2. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

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

    2016-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

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

  4. Stress Analysis of Single Spacer Grid Support considering Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

  5. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Ok; Hwang, Mi Jin; Kim, Sung Woo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples.

  6. Neutron noise measurements on Bugey 2 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Marini, J.; Romy, D.; Spadi, J.C.; Assedo, R.; Castello, G.

    1982-01-01

    Following Bugey 2 PWR hot functional tests, dimension measurements of internals hold down spring led to suspect that vibration levels could change with time. Neutron noise measurements runs during the first cycle enabled describing vibration behaviour of internals. Comparisons with previous analytical and experimental results gained on the Safran model as well as on similar reactors were also made.

  7. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  8. Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, Saptarshi, E-mail: saptarshi.bhattacharjee@outlook.com [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)

    2017-06-15

    Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

  9. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  10. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  11. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  12. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  13. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  14. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V.; Rosenberg, R. [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  15. Tie rod insertion test

    CERN Multimedia

    B. LEVESY

    2002-01-01

    The superconducting coil is inserted in the outer vaccum tank and supported by a set of tie rods. These tie rods are made of titanium alloy. This test reproduce the final insertion of the tie rods inside the outer vacuum tank.

  16. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  17. Approaches to analyze the bowing of German PWR fuel assemblies; Ansaetze zur Analyse des Biegeverhaltens deutscher DWR-Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    Boeke, H.; Bauer, R.; Bloemeling, F.; Lawall, R. [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2012-11-01

    The analysis of the bowing behavior of PWR fuel elements is required in case of increased fuel element deformations that have been observed during the last years. In the contribution the following issues are discussed: fuel element properties (stiffness, constructive features), influence factors (guiding tubes, spacer), load transfer and its impact. Under consideration of external boundary conditions an evaluation scheme was developed, using analysis data (control rod drop time), friction force measurements, fuel element characteristics (fuel element deformation, bowing) and their ranking, and simulation models (fluid-structure interactions). The evaluation scheme allows the definition of appropriate measures. The suitability of the methodology was demonstrated.

  18. Radiative heat transfer modelling in a PWR severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Magali Zabiego; Florian Fichot [Institut de Radioprotection et de Surete Nucleaire - BP 3 - 13115 Saint-paul-Lez-Durance (France); Pablo Rubiolo [Westinghouse Science and Technology - 1344 Beulah Road - Pittsburgh - PA 15235 (United States)

    2005-07-01

    Full text of publication follows: The present study is devoted to the estimation of the radiative heat transfers during a severe accident sequence in a Pressurized Water Reactor. In such a situation, the residual nuclear power released by the fuel rods can not be evacuated and heats up the core. As a result, the cylindrical rods and the structures initially composing the core undergo a degradation process: swelling, breaking or melting of the rods and structures and eventual collapse to form a heap of fragments called a debris bed. As the solid matrix loses its original shape, the core geometry continuously evolves from standing, regularly-spaced cylinders to a non-homogeneous system including deformed remaining rods and structures and debris particles. To predict this type of sequence, the ICARE/CATHARE software [1] is developed by IRSN. Since the temperatures can reach values greater than 3000 K, it was of major interest to provide the code with an accurate radiative transfer model usable whatever the geometry of the system. Considering the size of a reactor core compared to the mean penetration length of radiation, the core can be seen as an optically thick medium. This observation led us to use the diffusion approximation to treat the radiation propagation. In this approach, the radiative flux is calculated in a way similar to thermal conduction: q{sub r} = [K{sub e}].{nabla}T where [K{sub e}] is the equivalent conductivity tensor of the system accounting for thermal and radiative transfer. An homogenization technique is applied to estimate the equivalent conductivity. Given the temperature level, the radiative contribution to the equivalent conductivity tensor quickly becomes dominant. This model was described earlier in [2] in which it was shown that an equivalent conductivity can be continuously calculated in the system when the geometry evolves from standing regular cylinder rods to swollen or broken ones, surrounded or not by a film of liquid materials, to

  19. Reflood experiments in rod bundles with flow blockages due to clad ballooning

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S.K.; Kim, J.; Kim, K.; Kim, B.J.; Park, J.K.; Youn, Y.J.; Choi, H.S.; Song, C.H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-07-15

    Clad ballooning and the resulting partial flow blockage are one of the major thermal-hydraulic concerns associated with the coolability of partially blocked cores during a loss-of-coolant accident (LOCA). Several in-pile tests have shown that fuel relocation causes a local power accumulation and a high thermal coupling between the clad and fuel debris in the ballooned regions. However, previous experiments in the 1980s did not take into account the fuel relocation phenomena and resulting local power increase in the ballooned regions. The present paper presents the results of systematic investigations on the coolability of rod bundles with flow blockages. The experiments were mainly performed in 5 x 5 rod bundles, 2 x 2 rod bundles and other test facilities. The experiments include a reflood heat transfer, single-phase convective heat transfer, flow redistributions phenomena, and droplet break-up behavior. The effects of the fuel relocation and resulting local power increase were investigated using a 5 x 5 rod bundle. The fuel relocation phenomena increase the peak cladding temperature.

  20. Effect of co-free valve on activity reduction in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, C.B.; Han, B.C.; Bum, J.S.; Hwang, I.S. [Department of Nuclear Engineering, Seoul National Univ. (Korea, Republic of); Lee, C.B. [Korea Atomic Energy Research Inst., Daejon (Korea, Republic of)

    2002-07-01

    Radioactive nuclei, such as {sup 68}Co and {sup 60}Co, deposited on out-of-core surfaces in a pressurized water reactor (PWR) primary coolant system, are major sources of occupational radiation exposure to plant maintenance personnel and act as costly impediment to prompt and effective repairs. Valve hardfacing alloys exposed to primary coolant are considered as one of the main Co sources. To evaluate the Co-free valve, such as NOREM 02 and Deloro 50, the candidates for the alternative to Stellite 6, in a simulated PWR primary condition, SNU corrosion test loop (SCOTL) was constructed. For gate valves hard-faced with made of NOREM 02 and Deloro 50 hot cycling tests were conducted for up to 2,000 on-off cycles with cold leak tests at 1,000 cycle interval. It was observed that the leak rate of NOREM 02 (Fe-base) did not satisfy the nuclear grade valve leak criteria. After 1000 cycles test, while there was no leakage in case of Deloro 50 (Ni-base). Also, Deloro 50 showed no leakage after 2000 cycles. To estimate the activity reduction effect, we modified CRUDSIM-MIT which modeled the effects of coolant chemistry on the crud transport and activity buildup in the primary system of PWR. In the new code, crud evaluation and assessment (CREAT), {sup 60}Co activity buildup prediction includes 1) Co-base valve replacement effect, 2) Co-base valve maintenance effect, and 3) control rod drive mechanism (CRDM) and main coolant pump (MCP) shaft contribution. CREAT predicted that the main contributor of Co activity buildup was the corrosion-induced release of Co from the steam generator (SG) tubing. With new SG's tubed with alloy 690, Korean Next Generation Reactor (APR-1400) is expected to have about 64% lower Co activity on SG surface. The use of all Co-free valves is expected to cut additional 8% of activity which is only marginal. (authors)

  1. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  2. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  3. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  4. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Energy Technology Data Exchange (ETDEWEB)

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)

    2009-07-15

    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

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

  6. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay

    2014-01-01

    situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal......The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...

  7. Telescopic drilling rod

    Energy Technology Data Exchange (ETDEWEB)

    Kagan, I.L.; Berezov, S.I.; Gavrilov, G.A.; Goykhman, Ya.A.; Makushkin, D.O.; Rachev, M.P.; Voynich, L.K.

    1981-09-07

    The telescopic drilling rod includes an inner section of the rod, in whose center cable has been passed and is attached a bearing assembly connecting it to the winch, outer section of rod along which there is pipeline connecting the working cavity formed by the inner section of rod and the housing, installed on the lower end of the outer section of rod, with cavity formed by framework of the guide swivel and end piece and connected to the hydraulic system of the machine by pipeline, as well as clamping elements. In order to drill wells to a depth greater than the length of the outer sectrion of the rod, the latter jointly with the inner section of rod is lowered into the extreme lower position until swivel rests on the feed mechanism. With further slipping of cable and the absence of pressure in the hydraulic system, clamping elements do not have an effect on the inner section of rod. It has the opportunity to freely move along the outer section of rod downwards to the face. When pressure is supplied on pipeline into cavity and further through pipeline into working cavity, the inner section of rod is clamped with feed of the outer section in the process of drilling, both sections move jointly. Because of the link between working cavity of sleeve installed on the lower end of the outer section of rod, and the hydraulic system of the machine through the swivel cavity, it is possible to fix the drilling rod in any mutual axial position of the section.

  8. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  9. Horizontal Drop of 21- PWR Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2001-04-26

    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. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. 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. 11) 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.

  10. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2015-10-15

    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  11. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  12. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  13. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

    2017-04-01

    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

  14. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  15. The PWR cores management; La gestion des coeurs REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    2000-01-25

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  16. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    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)

  17. Degraded core analysis for the PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-10-01

    The paper presents an analysis of the probability and consequences of degraded core accidents for the PWR. The article is based on a paper which was presented by the author to the Sizewell-B public inquiry. Degraded core accidents are examined with respect to:- the initiating events, safety plant failure, and processes with a bearing on containment failure. Accident types and frequencies are discussed, as well as the dispersion of radionuclides. Accident risks, i.e. individual and societal risks in degraded core accidents are assessed from:- the amount of radionuclides released, the weather, the population distribution, and the accident frequencies. Uncertainties in the assessment of degraded core accidents are also summarized. (U.K.).

  18. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  19. Prediction of CRUD deposition on PWR fuel using a state-of-the-art CFD-based multi-physics computational tool

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, Victor [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Kendrick, Brian K. [Theoretical Division (T-1, MS B221), Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Walter, Daniel [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Manera, Annalisa, E-mail: manera@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Secker, Jeffrey [Westinghouse Electric Company Nuclear Fuel Division, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-04-01

    In the present paper we report about the first attempt to demonstrate and assess the ability of state-of-the-art high-fidelity computational tools to reproduce the complex patterns of CRUD deposits found on the surface of operating Pressurized Water Reactors (PWRs) fuel rods. A fuel assembly of the Seabrook Unit 1 PWR was selected as the test problem. During Seabrook Cycle 5, CRUD induced power shift (CIPS) and CRUD induced localized corrosion (CILC) failures were observed. Measurements of the clad oxide thickness on both failed and non-failed rods are available, together with visual observations and the results from CRUD scrapes of peripheral rods. Blind simulations were performed using the Computational Fluid Dynamics (CFD) code STAR-CCM+ coupled to an advanced chemistry code, MAMBA, developed at Los Alamos National Laboratory. The blind simulations were then compared to plant data, which were released after completion of the simulations.

  20. Fluid structure interaction between rods and a cross flow - Numerical approach

    Energy Technology Data Exchange (ETDEWEB)

    Simoneau, Jan-patrice, E-mail: jan-patrice.simoneau@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Sageaux, Thomas, E-mail: thomas.sageaux@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Moussallam, Nadim, E-mail: nadim.moussallam@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Bernard, Olivier, E-mail: olivier.bernard1@areva.com [Areva, 1, Place J. Millet, F 92084 Paris la Defense (France)

    2011-11-15

    This paper presents a full coupled approach between fluid dynamics and structure analysis. It is conducted in order to further improve the assessment of fluid structure interaction problems, occurring in the nuclear field such as the behavior of PWR fuel rods, steam generators and other heat exchangers tubes, fast breeder fuel assemblies. The coupling is obtained by implementing a beam mechanical model in user routines of the CFD code Star-CD, and thanks to a moving grid procedure. The configurations considered are rods in a cross flow. The model is first validated on a single rod case. The lock-in effect is pointed out and both amplitude and frequency responses of the single rod are positively compared to experimental data. Secondly, the mutual influence of two rods, either in-line or parallely set, is investigated. Different behaviors, bounded by critical distances between the rods are highlighted. Finally, the stability of a 3 Multiplication-Sign 3 bundle is calculated for different impinging velocities. Stable and unstable areas are found when varying the impinging velocity. Above a limit, the vibrations amplify up to a contact between rods, this bound is found slightly greater than literature values for close configurations. It is therefore expected that further calculations, with model refinements, will bring valuable informations about bundle stability.

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

    Science.gov (United States)

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

    1997-09-01

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

  2. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  3. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Wilson

    2001-02-08

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  4. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth D. Wright

    1997-07-30

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  5. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  6. Seismic qualification of PWR plant auxiliary feedwater systems

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14.

  7. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Terttaliisa, E-mail: terttaliisa.lind@psi.c [Paul Scherrer Institut, Villigen (Switzerland); Csordas, Anna Pinter; Nagy, Imre [HAS KFKI Atomic Energy Research Institute, Budapest (Hungary); Stuckert, Juri [Forschungszentrum Karlsruhe, Karlsruhe (Germany)

    2010-02-15

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T approx 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T approx 1650 K, followed by a

  8. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Science.gov (United States)

    Lind, Terttaliisa; Csordás, Anna Pintér; Nagy, Imre; Stuckert, Juri

    2010-02-01

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ˜ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ˜ 1650 K, followed by a relatively

  9. The advanced main control console for next japanese PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, A. [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Ito, K. [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama (Japan); Yokoyama, M. [Mitsubishi Electric Corporation, Energy and Industrial Systems Center, Kobe (Japan)

    2001-07-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  10. Effect of Flow Blockage on the Coolability during Reflood in a 2 × 2 Rod Bundle

    Directory of Open Access Journals (Sweden)

    Kihwan Kim

    2014-01-01

    Full Text Available During the reflood phase of a large-break loss-of-coolant accident (LBLOCA in a pressurized-water reactor (PWR, the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

  11. LWR fuel rod behavior during reactor tests under loss-of-coolant conditions: Results of the FR2 in-pile tests

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.; Sepold, L.; Hofmann, P.; Petersen, C.; Schanz, G.; Zimmermann, H. (Kernforschungszentrum Karlsruhe G.m.b.H. (Germany, F.R.))

    1982-05-01

    Results of the FR2 in-pile tests on fuel rod behavior under loss-of-coolant accident (LOCA) conditions are presented. To investigate the possible influence of a nuclear environment on fuel rod failure mechanisms, unirradiated as well as irradiated (2500 to 35,000 MWd/tsub(U)) PWR-type test fuel rods were exposed to temperature transients simulating the second heatup phase of a LOCA. Loaded by internal overpressure, the cladding ballooned and ruptured. The burst data do not indicate major differences from results obtained out-of-pile with electrically heated fuel rod simulators, and do not show an influence of burnup. The fuel pellets in previously irradiated rods, already cracked during normal operation, crumbled completely in the regions with large cladding deformation. Post-test examinations revealed cladding mechanical behavior and oxidation to be comparable to out-of-pile results, with relatively little fission gas release during the transient.

  12. Assessment of PWR fuel degradation by post-irradiation examinations and modeling in DEGRAD-1 code; Avaliacao da degradacao de combustivel PWR por exames pos-irradiacao e modelagem no codigo DEGRAD-1

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Lucki, Georgi; Silva, Jose Eduardo Rosa da; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: myrthes@ipen

    2005-07-01

    On the majority of the cases, the inquiries on primary failures and secondary in PWR fuel rods are based on results of analysis were made use of the non-destructive examination results (coolant activities monitoring, sipping tests, visual examination). The complementary analysis methodology proposed in this work includes a modeling approach to characterization of the physical effects of the individual chemistry mechanisms that constitute the incubation phase of degradation phenomenon after primary failure that are integrated in the reactor operational history under stationary operational regime, and normal power transients. The computational program called DEGRAD-1 was developed based on this modeling approach. The practical outcome of the program is to predict cladding regions susceptible to massive hydriding. The applications presented demonstrate the validity of proposed method and models by actual cases simulation, which (primary and secondary) defects positions were known and formation time was estimated. By using the modeling approach, a relationship between the hydrogen concentration in the gap and the inner cladding oxide thickness has been identified which, when satisfied, will induce massive hydriding. The novelty in this work is the integrated methodology, which supplements the traditional analysis methods (using data from non-destructive techniques) with mathematical models for the hydrogen evolution, oxidation and hydriding that include refined approaches and criteria for PWR fuel, and using the FRAPCON-3 fuel performance code as the basic tool. (author)

  13. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  14. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  15. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  16. Advanced ion exchange resins for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, B. [Rohm and Haas Co. (United States); Tsuzuki, S. [Rohm and Haas Co. (Japan)

    2002-07-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  17. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: nolosesamuel@prodigy.net.mx

    2008-07-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  18. Morphoelastic rods. Part I: A single growing elastic rod

    KAUST Repository

    Moulton, D.E.

    2013-02-01

    A theory for the dynamics and statics of growing elastic rods is presented. First, a single growing rod is considered and the formalism of three-dimensional multiplicative decomposition of morphoelasticity is used to describe the bulk growth of Kirchhoff elastic rods. Possible constitutive laws for growth are discussed and analysed. Second, a rod constrained or glued to a rigid substrate is considered, with the mismatch between the attachment site and the growing rod inducing stress. This stress can eventually lead to instability, bifurcation, and buckling. © 2012 Elsevier Ltd. All rights reserved.

  19. Learning with Rods: One Account.

    Science.gov (United States)

    Cherry, Donald Esha

    This paper discusses one English as a Second Language (ESL) teacher's attempts to use cuisenaire rods as a language learning tool. Cuisenaire rods (sometimes called algebricks) vary in size from 1 x 1 x 10 centimeter sticks to 1 x 1 x 1 centimeter cubes, with each of the 10 sizes a different color. Although such rods have been used to teach…

  20. Experiment data report IFA-226 postirradiation examination. [PWR, BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bagger, C.; Carlsen, H.; Domanus, J.; Hougaard, H.; Larsen, E.; Larsen, N.

    1977-09-01

    IFA-226 contained twelve, mixed plutonium-uranium oxide fuel rods arranged in two, six-rod clusters. The assembly was designed to study fuel-cladding mechanical interaction, fuel thermal response, and fission gas release as a function of fuel density, initial fuel-to-cladding gap, rod power, and burnup. Data were obtained from fuel rod centerline thermocouples, fission gas pressure transducers, and cladding elongation sensors. Results of both nondestructive and destructive examinations are presented. The PIE indicated that one fuel rod failed during service as a result of internal hydriding of the end plug. Circumferential cladding ridges resulting from fuel-cladding interaction were present on all of the rods, with the largest ridges present on the rod with the smallest initial fuel-to-cladding gap. No incipient fuel rod failures were detected.

  1. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  2. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Energy Technology Data Exchange (ETDEWEB)

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.

    1999-05-15

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  3. Studies of a small PWR for onsite industrial power

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, O.H.; Smith, W.R.

    1977-04-19

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

  4. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  5. PWR fuel in Japan; The changes and trend for hereafter

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki

    1992-07-01

    As for the PWR fuel in Japan, much efforts have been exerted aiming at the high reliability since the start of operation of Mihama No. 1 plant of Kansai Electric Power Co., Inc. At the beginning of 1970s, the fuel made by Westinghouse in USA was imported, and since then, the pursuit of the causes of troubles and the countermeasures and the domestic production of fuel have been carried out, and the improvement of design and the strengthening of quality control have been advanced. As the results, the occurrence of troubles decreased rapidly. As the fuel improvement for hereafter, the economical improvement by higher burnup, the saving and effective use of uranium resources as well as the increase of reliability are emphasized. The changes in the PWR fuel by Westinghouse, the course of improvement in the PWR fuel in Japan, the improvement against the troubles of the fuel, the improved design, the verification of the performance of the PWR fuel, the trend of development of the fuel such as the heightening of burnup, the saving and effective use of uranium resources, and the improved type pressurized water reactors are reported. (K.I.).

  6. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  7. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  8. Research on PWR Core Performance With MOX Fuel Loading%MOX燃料对压水堆堆芯性能影响研究

    Institute of Scientific and Technical Information of China (English)

    李小生; 靳忠敏

    2013-01-01

    Use of MOX fuel in nuclear reactors is an effective way to dispose of plutonium .A large PWR reactor core with full core loading UO 2 fuel was referenced , the reactor core physics parameters of PWR with whole and part core loading MOX fuel were calculated by using DRAGON and DONJON codes ,and the reactivity worth of control rods and boron acid solution were researched under loading MOX fuel . The results show that PWR core with MOX fuel can achieve the desired cycle length and power distribution ,but loading MOX fuel will significantly decrease the reactivity worth of control rod and boron acid solution ,moreover ,the proportion of loading MOX fuel is positive to the decrease degree of reactivity worth .%在核反应堆中使用MOX燃料是处置钚的有效方式。以大型全UO2燃料压水堆堆芯设计作为参考,使用DRAGON、DONJON程序,计算在大型压水堆中全堆芯及部分堆芯装载MOX燃料后反应堆部分物理性能指标,研究加入MOX燃料后对控制棒与硼酸溶液的反应性价值的影响。结果表明,压水堆堆芯装载各比例MOX燃料均可达到与全UO2燃料堆芯相当的循环长度,功率分布也能满足相应的安全限值要求,但采用MOX燃料会造成控制棒与硼溶液的反应性价值降低,且降低程度与MOX燃料装载比例成正相关。

  9. Safety rod latch inspection

    Energy Technology Data Exchange (ETDEWEB)

    Leader, D.R.

    1992-02-01

    During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and metallography. During the examination of the L-Area latches, a failed latch mechanism from the 100 K reactor was added to the investigation. Fourteen latches that had a history of problems were removed from K-Area and sent to SRL for inclusion in this study the week after the original seven assemblies were examined, bringing the total of latch assemblies discussed in this report to twenty one. Results of the examination of the K-Area latch that initiated this study is not included in this report.

  10. Safety rod latch inspection

    Energy Technology Data Exchange (ETDEWEB)

    Leader, D.R.

    1992-02-01

    During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small ``button`` in the latch mechanism had broken off of the ``lock plunger`` and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and metallography. During the examination of the L-Area latches, a failed latch mechanism from the 100 K reactor was added to the investigation. Fourteen latches that had a history of problems were removed from K-Area and sent to SRL for inclusion in this study the week after the original seven assemblies were examined, bringing the total of latch assemblies discussed in this report to twenty one. Results of the examination of the K-Area latch that initiated this study is not included in this report.

  11. Evaluation of Fuel Performance Uncertainty in a PWR HFP RIA Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    Sensitivity and combined uncertainty studies based on the various kinds of uncertainty sources have been carried out in a PWR hot full power (HFP) condition. - Cladding inner diameter, fuel thermal conductivity, fuel thermal expansion and peak power have induced a significant impact to the fuel enthalpy and temperature. - Cladding hoop strain was strongly affected by the uncertainty parameters of cladding inner diameter, fuel thermal expansion, EPRI-1 CHF and peak power. - Above results are valid in the given analysis condition in this paper. Thereby, the analysis conditions, for example the peak linear heat rate before RIA or peak power and FWHM etc, are changed the results will be changed also. Approved analysis methodology for licensing application in the safety analysis of reactivity initiated accident (RIA) in Korea is based on a conservative approach. But newly introduced safety criteria, described in section 4.2 of NUREG-0800, tend to reduce the margins or depending on the reactor types rod failure is predicted due to the pellet-to-cladding mechanical interaction (PCMI) criteria. Thereby, licensee is trying to improve the margins by utilizing a less conservative approach.

  12. Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR

    Science.gov (United States)

    Wei, Guohai; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; He, Linfeng; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

  13. Qualification test of the EPR control rod drive mechanism in the full scale component test facility KOPRA

    Energy Technology Data Exchange (ETDEWEB)

    Herr, Wolfgang; Sykora, Alexander; Kleideiter, Ansgar [AREVA NP GmbH, Erlangen (Germany); Champomier, Francois [AREVA NP SAS, Paris (France)

    2009-07-01

    The control rod drive mechanism (CRDM) and the mobile set consisting of rod cluster control assembly (RCC-A) of the evolutionary power reactor (EPR) had to pass a full scale qualification test in representative site conditions. The KOPRA core test section in Erlangen is precisely designed for full scale tests on nuclear core components in respect to coolant temperature and volume flow of PWR site conditions. In the test channel the complete geometry of the central core position of the reactor pressure vessel is simulated with 1:1 scale. The performance test program has led to an optimized test sequence through small adjustments in operating parameters of CRDM. The endurance test program has demonstrated that all tested components, i.e. the CRDN, the control rod driveline and the components of the drop channel are able to function properly and to meet the specification goals.

  14. Decontamination of control rod housing from Palisades Nuclear Power Station.

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, M.D.; Nunez, L.; Purohit, A.

    1999-05-03

    Argonne National Laboratory has developed a novel decontamination solvent for removing oxide scales formed on ferrous metals typical of nuclear reactor piping. The decontamination process is based on the properties of the diphosphonic acids (specifically 1-hydroxyethane-1,1-diphosphonic acid or HEDPA) coupled with strong reducing-agents (e.g., sodium formaldehyde sulfoxylate, SFS, and hydroxylamine nitrate, HAN). To study this solvent further, ANL has solicited actual stainless steel piping material that has been recently removed from an operating nuclear reactor. On March 3, 1999 ANL received segments of control rod housing from Consumers Energy's Palisades Nuclear Plant (Covert, MI) containing radioactive contamination from both neutron activation and surface scale deposits. Palisades Power plant is a PWR type nuclear generating plant. A total of eight segments were received. These segments were from control rod housing that was in service for about 6.5 years. Of the eight pieces that were received two were chosen for our experimentation--small pieces labeled Piece A and Piece B. The wetted surfaces (with the reactor's pressurized water coolant/moderator) of the pieces were covered with as a scale that is best characterized visually as a smooth, shiny, adherent, and black/brown in color type oxide covering. This tenacious oxide could not be scratched or removed except by aggressive mechanical means (e.g., filing, cutting).

  15. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  16. Cone rod dystrophies

    Directory of Open Access Journals (Sweden)

    Hamel Christian P

    2007-02-01

    Full Text Available Abstract Cone rod dystrophies (CRDs (prevalence 1/40,000 are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP, also called the rod cone dystrophies (RCDs resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7. Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far. The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs, CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs, and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs. It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is

  17. ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

    Directory of Open Access Journals (Sweden)

    Endiah Puji Hastuti

    2015-04-01

    Full Text Available Turbulensi aliran pendingin pada proses perpindahan panas berfungsi untuk meningkatkan nilai koefisien perpindahan panas, tidak terkecuali aliran dalam kanal bahan bakar. Program CFD (CFD=computational fluid dynamics, FLUENT adalah program komputasi berbasis elemen hingga (finite element yang mampu memprediksi dan menganalisis fenomena dinamika aliran fluida secara teliti. Program perhitungan CFD dipilih dalam penelitian ini karena selain akurat juga dapat memberikan visualisasi dengan baik. Penelitian ini bertujuan untuk memahami karakteristika perpindahan panas, massa dan momentum dari dinding rod bahan bakar ke pendingin secara visual, pada medan temperatur, medan tekanan, dan medan energi kinetika pendingin, sebagai fungsi dinamika aliran di dalam kanal, pada kondisi tunak dan transien. Analisis dinamika aliran pada kanal bahan bakar PWR berbasis CFD dilakukan dengan menggunakan sampel data reaktor PWR dengan daya 1000 MWe dengan susunan bahan bakar 17x17. Untuk menguji sensitivitas persamaan aliran yang sesuai dengan model aliran turbulen pada kanal bahan bakar dilakukan pemodelan dengan menggunakan persamaan k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, dan Reynold stress model (RSM. Pada analisis sensitivitas aliran turbulen di dalam kanal digunakan model mesh hexahedral dengan memilih tiga geometri sel yang masing masing berukuran 0,5 mm; 0,2 mm dan 0,15 mm. Hasil analisis menunjukkan bahwa pada analisis kondisi tunak (steady state, terdapat hasil yang mirip pada model turbulen Ƙ-ε standard dan Ƙ-ω standard. Pengujian terhadap kriteria Dittus Boelter untuk bilangan Nusselt menunjukkan bahwa model Reynold stress model (RSM direkomendasikan. Analisis sensitivitas terhadap geometri mesh antara sel yang berukuran 0,5 mm, 0,2 mm dan 0,15 mm, menunjukkan bahwa geometri sel sebesar 0,5 mm telah mencukupi. Aliran turbulen berkembang penuh telah tercapai pada model LES dan DES, meskipun hanya dalam waktu singkat (3 s, model LES memerlukan waktu komputasi

  18. Assessment of PWR plutonium burners for nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, A J; Shapiro, N L

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  19. PWR fuel in Japan; Progress and future trends

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki (Mitsubishi Heavy Industries Ltd., Tokyo (Japan))

    1994-06-01

    Twenty years ago, in the early years of the Japanese civil nuclear power programme, the fuel used was imported from Westinghouse in the USA. However, it was always intended that there would be a move towards fuel fabrication in Japan and by the end of 1993 around 10,000 Mitsubishi PWR fuel assemblies had been supplied to 21 PWRs in Japan. The highest burnup achieved so far is 46 GWd/t. Design changes to reduce abnormalities have been made, reliability is improving all the time and further improvements in burnup are being developed. This progress in PWR cores and fuel including MOX fuel in Japan is charted and future research and development is outlined. (UK).

  20. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  1. Control of corrosion product transport in PWR secondary cycles

    Energy Technology Data Exchange (ETDEWEB)

    Sawochka, S.G.; Pearl, W.L. [NWT Corp., San Josa, CA (United States); Passell, T.O.; Welty, C.S. [Electric Power Research Institute, Palo Alto, CA (United States)

    1992-12-31

    Transport of corrosion products to PWR steam generators by the feedwater leads to sludge buildup on the tubesheets and fouling of tube-to-tube support crevices. In these regions, chemical impurities concentrate and accelerate tubing corrosion. Deposit buildup on the tubes also can lead to power generation limitations and necessitate chemical cleaning. Extensive corrosion product transport data for PWR secondary cycles has been developed employing integrating sampling techniques which facilitate identification of major corrosion product sources and assessments of the effectiveness of various control options. Plant data currently are available for assessing the impact of factors such as pH, pH control additive, materials of construction, blowdown, condensate treatment, and high temperature drains and feedwater filtration.

  2. Active Brownian rods

    Science.gov (United States)

    Peruani, Fernando

    2016-11-01

    Bacteria, chemically-driven rods, and motility assays are examples of active (i.e. self-propelled) Brownian rods (ABR). The physics of ABR, despite their ubiquity in experimental systems, remains still poorly understood. Here, we review the large-scale properties of collections of ABR moving in a dissipative medium. We address the problem by presenting three different models, of decreasing complexity, which we refer to as model I, II, and III, respectively. Comparing model I, II, and III, we disentangle the role of activity and interactions. In particular, we learn that in two dimensions by ignoring steric or volume exclusion effects, large-scale nematic order seems to be possible, while steric interactions prevent the formation of orientational order at large scales. The macroscopic behavior of ABR results from the interplay between active stresses and local alignment. ABR exhibit, depending on where we locate ourselves in parameter space, a zoology of macroscopic patterns that ranges from polar and nematic bands to dynamic aggregates.

  3. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  4. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  5. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

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

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available 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.

  7. PWR Cross Section Libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, Carolyn [Texas A& M University; Ilas, Germina [ORNL

    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.

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

    Science.gov (United States)

    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.

  9. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  10. PWR core stablity aganst xenon-induced spatial power oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H.J.; Han, K.I. (Korea Advanced Energy Research Inst., Seoul (Republic of Korea))

    1982-06-01

    Stability of a PWR core against xenon-induced axial power oscillation is studied using one-dimensional xenon transient analysis code, DD1D, that has been developed and verified at KAERI. Analyzed by DD1D utilizing the Kori Unit 1 design and operating data is the sensitivity of axial stability in a PWR core to the changes in core physical parameters including core power level, moderator temperature coefficient, core inlet temperature, doppler power coefficient and core average burnup. Through the sensitivity study the Kori Unit 1 core is found to be stable against axial xenon oscillation at the beginning of cycle 1. But, it becomes less stable as burnup progresses, and unstable at the end of cycle. Such a decrease in stability is mainly due to combined effect of changes in axial power distribution, moderator temperature coefficient and doppler power coefficient as core burnup progresses. It is concluded from the stability analysis of the Kori Unit 1 core that design of a large PWR with high power density and increased dimension can not avoid xenon-induced axial power instabilites to some extents, especially at the end of cycle.

  11. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  12. Control of Rod-Rod Interactions in Poly(3-alkylthiophenes)

    Science.gov (United States)

    Ho, Victor; Boudouris, Bryan W.; Segalman, Rachel A.

    2010-03-01

    Poly(3-hexylthiophene) is a commonly used semiconducting polymer because of its relatively high charge transport ability, low band gap, and solution processiblity. Strong intermolecular interactions lead to the formation of nanofibers during crystallization, which prevents long-range microstructural ordering. We show rod-rod interactions, parameterized by the Maier-Saupe parameter, can be controlled by rational polythiophene side chain design. Effects of side chain passivation are evidenced by a depressed melting temperature and the presence of a liquid crystalline region. Additionally, the Maier-Saupe parameters are estimated for poly(3-dodecylthiophene) and poly(3-ethylhexylthiophene); the relative magnitudes of each are related to the interchain spacings obtained by x-ray diffraction experiments. The systematic tuning of the rod-rod interactions in polythiophenes allows for manipulation of the ratio of Maier-Saupe to the Flory-Huggins parameter, a crucial value in obtaining long-range order in rod-coil block copolymer morphologies.

  13. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  14. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  15. Cuisenaire Rods Go to College.

    Science.gov (United States)

    Chinn, Phyllis; And Others

    1992-01-01

    Presents examples of questions and answers arising from a hands-on and exploratory approach to discrete mathematics using cuisenaire rods. Combinatorial questions about trains formed of cuisenaire rods provide the setting for discovering numerical patterns by experimentation and organizing the results using induction and successive differences.…

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

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    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. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  17. AgInCd control rod failure in the QUENCH-13 bundle test

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)], E-mail: leo.sepold@imf.fzk.de; Lind, T. [Paul Scherrer Institut, Laboratory for Thermalhydralics (LTH), Department of Nuclear Energy and Safety (NES), 5232 Villigen PSI (Switzerland); Csordas, A. Pinter [Fuel Materials Department, HAS KFKI AEKI, 1121 Budapest (Hungary); Stegmaier, U.; Steinbrueck, M.; Stuckert, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2009-09-15

    The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO{sub 2} pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI-Switzerland and AEKI-Hungary. Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g

  18. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Energy Technology Data Exchange (ETDEWEB)

    Hosler, J.

    1982-12-01

    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  19. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    CERN Document Server

    Bakosi, J; Lowrie, R B; Pritchett-Sheats, L A; Nourgaliev, R R

    2013-01-01

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3x3 and 5x5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carried out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the single-phase incompressible Navier-Stokes equations. The simulations explicitly resolve the la...

  20. Eulerian formulation of elastic rods

    Science.gov (United States)

    Huynen, Alexandre; Detournay, Emmanuel; Denoël, Vincent

    2016-06-01

    In numerous biological, medical and engineering applications, elastic rods are constrained to deform inside or around tube-like surfaces. To solve efficiently this class of problems, the equations governing the deflection of elastic rods are reformulated within the Eulerian framework of this generic tubular constraint defined as a perfectly stiff normal ringed surface. This reformulation hinges on describing the rod-deformed configuration by means of its relative position with respect to a reference curve, defined as the axis or spine curve of the constraint, and on restating the rod local equilibrium in terms of the curvilinear coordinate parametrizing this curve. Associated with a segmentation strategy, which partitions the global problem into a sequence of rod segments either in continuous contact with the constraint or free of contact (except for their extremities), this re-parametrization not only trivializes the detection of new contacts but also transforms these free boundary problems into classic two-points boundary-value problems and suppresses the isoperimetric constraints resulting from the imposition of the rod position at the extremities of each rod segment.

  1. Status of rod consolidation, 1988

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1989-01-01

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs.

  2. PWR safety and relief valve test program. Valve selection/juftification report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    NUREG 0578 required that full-scale testing be performed on pressurizer safety valves and relief valves representative of those in use or planned for use in PWR plants. To obtain valve performance data for the entire population of PWR plant valves, nine safety valves and ten relief valves were selected as a fully representative set of test valves. Justification that the selected valves represent all PWR plant valves was provided by each safety and relief valve manufacturer. Both the valve selection and justification work was performed as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of the PWR utilities in response to the recommendations of NUREG 0578 and the requirements of the NRC. Results of the Safety and Relief Valve Selection and Justification effort is documented in this report.

  3. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  4. Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

    Energy Technology Data Exchange (ETDEWEB)

    S. Mastilovic; A. Scheider; S.M. Bennett

    2001-01-29

    The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only.

  5. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    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.

  6. Integral Test Facility PKL: Experimental PWR Accident Investigation

    OpenAIRE

    2012-01-01

    Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circul...

  7. Topological mixing with ghost rods

    Science.gov (United States)

    Gouillart, Emmanuelle; Thiffeault, Jean-Luc; Finn, Matthew D.

    2006-03-01

    Topological chaos relies on the periodic motion of obstacles in a two-dimensional flow in order to form nontrivial braids. This motion generates exponential stretching of material lines, and hence efficient mixing. Boyland, Aref, and Stremler [J. Fluid Mech. 403, 277 (2000)] have studied a specific periodic motion of rods that exhibits topological chaos in a viscous fluid. We show that it is possible to extend their work to cases where the motion of the stirring rods is topologically trivial by considering the dynamics of special periodic points that we call “ghost rods”, because they play a similar role to stirring rods. The ghost rods framework provides a new technique for quantifying chaos and gives insight into the mechanisms that produce chaos and mixing. Numerical simulations for Stokes flow support our results.

  8. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

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

    Science.gov (United States)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

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

  10. ORNL rod-bundle heat-transfer test data. Volume 6. Thermal-hydraulic test facility experimental data report for test 3. 05. 5B - double-ended cold-leg break simulation

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-05-18

    Thermal-Hydraulic Test Facility (THTF) Test 3.05.5B was conducted by members of the ORNL PWR Blowdown Heat Transfer Separate-Effects Program on July 3, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.05.5B was designed to provide transient thermal-hydraulics data in rod bundle geometry under reactor accident-type conditions. Reduced instrument responses are presented. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  11. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  12. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  13. A particle assembly/constrained expansion (PACE) model for the formation and structure of porous metal oxide deposits on nuclear fuel rods in pressurized light water reactors

    Science.gov (United States)

    Brenner, Donald W.; Lu, Shijing; O'Brien, Christopher J.; Bucholz, Eric W.; Rak, Zsolt

    2015-02-01

    A new model is proposed for the structure and properties of porous metal oxide scales (aka Chalk River Unidentified Deposits (CRUD)) observed on the nuclear fuel rod cladding in Pressurized Water Reactors (PWR). The model is based on the thermodynamically-driven expansion of agglomerated octahedral nickel ferrite particles in response to pH and temperature changes in the CRUD. The model predicts that porous nickel ferrite with internal {1 1 1} surfaces is a thermodynamically stable structure under PWR conditions even when the free energy of formation of bulk nickel ferrite is positive. This explains the pervasive presence of nickel ferrite in CRUD, observed CRUD microstructures, why CRUD maintains its porosity, and variations in porosity within the CRUD observed experimentally. This model is a stark departure from decades of conventional wisdom and detailed theoretical analysis of CRUD chemistry, and defines new research directions for model validation, and for understanding and ultimately controlling CRUD formation.

  14. Response Surface Methodology Control Rod Position Optimization of a Pressurized Water Reactor Core Considering Both High Safety and Low Energy Dissipation

    Directory of Open Access Journals (Sweden)

    Yi-Ning Zhang

    2017-02-01

    Full Text Available Response Surface Methodology (RSM is introduced to optimize the control rod positions in a pressurized water reactor (PWR core. The widely used 3D-IAEA benchmark problem is selected as the typical PWR core and the neutron flux field is solved. Besides, some additional thermal parameters are assumed to obtain the temperature distribution. Then the total and local entropy production is calculated to evaluate the energy dissipation. Using RSM, three directions of optimization are taken, which aim to determine the minimum of power peak factor Pmax, peak temperature Tmax and total entropy production Stot. These parameters reflect the safety and energy dissipation in the core. Finally, an optimization scheme was obtained, which reduced Pmax, Tmax and Stot by 23%, 8.7% and 16%, respectively. The optimization results are satisfactory.

  15. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    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.

  16. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.

    1996-10-01

    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  17. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  18. PWR reactor vessel in-service-inspection according to RSEM

    Energy Technology Data Exchange (ETDEWEB)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)

    2006-07-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high

  19. CFD simulation of turbulent flow in a rod bundle with spacer grids (MATIS-H) using STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Cinosi, N., E-mail: n.cinosi@imperial.ac.uk; Walker, S.P.; Bluck, M.J.; Issa, R.

    2014-11-15

    Highlights: • CDF simulation of turbulent flow generated by a typical PWR spacer grid. • Benchmarking against the MATIS-H experiments run at KAERI in Daejeon, Korea. • Deployment of various steady RANS models to compute the turbulence. • Sensitivity analysis of hardware components. - Abstract: This paper presents the CFD simulation of the turbulent flow generated by a model PWR spacer grid within a rod bundle. The investigation was part of the MATIS-H benchmark exercise, organized by the OECD-NEA, with measurements performed at the KAERI facilities in Daejeon, Korea. The study employed the CD-Adapco code Star-CCM+. An initial sensitivity study was conducted to attempt to assess the importance to the overall flow of components such as the outlet plenum and the end support grid; these were shown to be able to be safely neglected, but the tapered end portion of the rods was found to be significant, and this was incorporated in the model analyzed. A RANS model using any of K-epsilon, K-omega and Reynolds-stress turbulence models was found to be adequate for the prediction of mean velocity profiles, but they all three underestimate the time-averaged turbulent velocity components. Vorticity seems to be better predicted, although the measured values of vorticity are only presented via colored contour plots, making quantitative comparison rather difficult. Circulation, calculated via an integral for each channel, seems to be well predicted by all three models.

  20. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

    2008-09-15

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

  1. PIE of the second fuel rod from the LOCA experiment (IFA-650.2)

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Jenssen, H.K.; Espeland, M.; Solum, N.O.

    2005-07-01

    The LOCA experiment on the second rod (IFA-650.2) a fresh, low-tin Zr-4, pressurised PWR rod was carried out in May 2004. The main objective was to produce ballooning, to determine the time to burst and to assess the material oxidation and hydriding kinetics. The rod pressure at hot conditions and peak PCT were 70 bar and 1050 C, respectively. To document the effect of the LOCA testing on the cladding, rod 2 was subjected in PIE to visual inspection, profilometry and metallography. On visual inspection the clad shows a typical balloon. In the region of max ballooning the clad shows a 35 mm long, < 20 mm burst opening. In the balloon region the outer clad diameter increased by <35% and locally the wall thickness reduction is >55%. The entire rod is covered with a black oxide layer. Below and above the split opening the continuous oxide layer was 40 to 50mum both on water and fuel side of the clad. At the locations of the upper and lower cladding thermocouples the oxide thickness was in the range 24-27 mum. Widmanstaetten structure is seen in the bulk of the clad and confirms the high temperature experienced. A some 40mum wide, hard and brittle zone with oxygen rich globular alpha-grains is found both at the outer and the inner edge of the clad in the balloon region. The zone is widest near the axial split (crack). In the clad few, arbitrary oriented hydride platelets are observed in the balloon area. (Author)

  2. CFD analyses of flow structures in air-ingress and rod bundle problems

    Science.gov (United States)

    Wei, Hong-Chan

    Two topics from nuclear engineering field are included in this dissertation. One study is the air-ingress phenomenon during a loss of coolant accident (LOCA) scenario, and the other is a 5-by-5 bundle assembly with a PWR design. The objectives were to investigate the Kelvin-Helmholtz instability of the gravity-driven stratified flows inside a coaxial pipe and the effects caused by two types of spacers at the downstream of the rod bundle. Richardson extrapolation was used for the grid independent study. The simulation results show good agreements with the experiments. Wavelet analysis and Proper Orthogonal Decomposition (POD) were used to study the flow behaviors and flow patterns. For the air-ingress phenomenon, Brunt-Vaisala frequency, or buoyancy frequency, predicts a frequency of 2.34 Hz; this is confirmed by the dominant frequency of 2.4 Hz obtained from the wavelet analysis between times 1.2 s and 1.85 s. For the rod bundle study, the dominant frequency at the center of the subchannel was determined to be 2.4 Hz with a secondary dominant frequency of 4 Hz and a much minor frequency of 6 Hz. Generally, wavelet analysis has much better performance than POD, in the air-ingress phenomenon, for a strongly transient scenario; they are both appropriate for the rod bundle study. Based on this study, when the fluid pair in a real condition is used, the time which air intrudes into the reactor is predictable.

  3. On the Minimum Safety Factor in Elastic Buckling of Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Kyu; Kim, Jae Yong; Yoon, Kyung Ho; Lee, Young Ho; Lee, Kang Hee; Kang, Heung Seok; Song, Kun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Elastic buckling of a thin tube is an instantaneous collapse phenomenon due to an external pressure. This should be prohibited for a PWR (Pressurized Water Reactor) fuel rod. There is an engineering formula of it; however, safety factor used to be applied to the calculation results since there will be uncertainty in the parameters of the formulae such as dimensional tolerances, environmental conditions and so forth. It is a designer's responsibility to determine an appropriate safety factor that is acceptably economically conservative. Mechanical properties of a material are usually adopted from a material handbook. However, they are usually different from the measured values of the material actually used. A local dimension anomaly critically affects the elastic buckling. Conventional safety factors against the elastic buckling seemed to be large (more than 3.5). However, the reason for this is rarely found. Engineering experience may be incorporated. Therefore, it is highly necessary to propose a minimum safety factor on the elastic buckling while accommodating the above mentioned uncertainties. It is so especially for the dual cooled fuel rod since it has never been used before. The primary purpose of this work is to quantify the aforementioned uncertainties of the parameters in the elastic buckling formula, especially for an outer cladding of the currently studied dual cooled fuel rod. It is extended from the previous theoretical and experimental study

  4. Experiments on silver-indium-cadmium control rod failure during severe accident sequences

    Energy Technology Data Exchange (ETDEWEB)

    Steinbrueck, M.; Stegmaier, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany)

    2010-05-15

    Silver-indium-cadmium (SIC) alloy is used as neutron absorber material in control rods (CR) of Pressurized Water Reactors (PWR). It is the material with the lowest melting temperature (approx. 1100 K) among all metallic and ceramic materials applied in nuclear reactors. During a hypothetical severe accident the SIC melt is kept in its stainless steel (SS) cladding tube as long as this is intact. After failure of the cladding tube by eutectic interaction with the Zircaloy-4 (Zry-4) guide tube or latest by reaching the SS melting temperature SIC elements are released and may interact with other core components. Furthermore, Ag-In-Cd are one of the main contributors to aerosol release in the reactor cooling system and may strongly influence nature and transport of fission products in the primary circuit and later on in the containment. The bundle experiment QUENCH-13 with prototypical SIC control rod as well as two series of single-rod tests with 10-cm long CR segments were performed at Karlsruhe Institute of Technology (KIT, former FZK) in order to improve the data base on SIC CR degradation and aerosol release. This paper concentrates on the degradation and failure mechanisms of SIC CRs as well as on the interaction between SIC absorber melt with other core components. (orig.)

  5. The Third ATLAS ROD Workshop

    CERN Multimedia

    Poggioli, L.

    A new-style Workshop After two successful ATLAS ROD Workshops dedicated to the ROD hardware and held at the Geneva University in 1998 and in 2000, a new style Workshop took place at LAPP in Annecy on November 14-15, 2002. This time the Workshop was fully dedicated to the ROD-TDAQ integration and software in view of the near future integration activities of the final RODs for the detector assembly and commissioning. More precisely, the aim of this workshop was to get from the sub-detectors the parameters needed for T-DAQ, as well as status and plans from ROD builders. On the other hand, what was decided and assumed had to be stated (like EB decisions and URDs), and also support plans. The Workshop gathered about 70 participants from all ATLAS sub-detectors and the T-DAQ community. The quite dense agenda allowed nevertheless for many lively discussions, and for a dinner in the old town of Annecy. The Sessions The Workshop was organized in five main sessions: Assumptions and recommendations Sub-de...

  6. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2017-02-15

    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  7. OPR1000 Control Rod Drop Accident Simulation using the SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chang Keun; Ha, Sang Jun; Moon, Chan Kook [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

    2012-05-15

    The Korea nuclear industry has developed a best estimated two-phase three-filed thermal-hydraulic analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plants), for safety analysis and design of a PWR (Pressurized Water Reactor). As the first phase, the demo version of the SPACE code was released in March 2010. The code has been verified and improved according to the Validation and Verification (V and V) matrix prepared for the SPACE code as the second phase of the development. In this study, a Control Rod Drop accident has been simulated using the SPACE code as one aspect of the V and V work. The results from this test were compared with tests of the RETRAN and CESEC codes

  8. Topological Optimization of Rod Mixers

    Science.gov (United States)

    Finn, Matthew D.; Thiffeault, Jean-Luc

    2006-11-01

    Stirring of fluid with moving rods is necessary in many practical applications to achieve homogeneity. These rods are topological obstacles that force stretching of fluid elements. The resulting stretching and folding is commonly observed as filaments and striations, and is a precursor to mixing. In a space-time diagram, the trajectories of the rods form a braid [1], and the properties of this braid impose a minimal complexity in the flow. We discuss how optimal mixing protocols can be obtained by a judicious choice of braid, and how these protocols can be implemented using simple gearing [2].[12pt] [1] P. L. Boyland, H. Aref, and M. A. Stremler, JFM 403, 277 (2000).[8pt] [2] J.-L. Thiffeault and M. D. Finn, http://arxiv.org/nlin/0603003

  9. Advanced gray rod control assembly

    Energy Technology Data Exchange (ETDEWEB)

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  10. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    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.

  12. PWR steam generator chemical cleaning, Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  13. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  14. Identifying thermal cycling mechanisms in PWR branch line piping

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Keller, J.D.; Bilanin, A.J. [Continuum Dynamics, Inc., Ewing, NJ (United States)

    2002-07-01

    Predicting the onset and the characteristics of thermal cycling in pressurized water reactor (PWR) branch line piping systems is critical to formulation of thermal fatigue screening tools. The complex nature of the underlying thermal-hydraulic phenomena, however, significantly complicates prediction using analytical models or direct numerical simulations. Instead, it is necessary to perform scaled experiments to identify the physical mechanisms and to gather data for formulation of semi-empirical models for the thermal cycling phenomena. Through the EPRI Materials Reliability Program a test program is underway to identify and develop semi-empirical correlations for the physical thermalhydraulic mechanisms that cause thermal cycling in dead-ended PWR branch line piping systems. Three series of tests are being performed in this test program: configuration tests on a representative up-horizontal (UH) branch line piping geometry, configuration tests on a representative down-horizontal (DH) branch line piping geometry, and high Reynolds number tests to assess penetration of secondary flow structures into a dead-ended branch line. Results from UH and DH configuration tests indicate that random turbulence penetration is not sufficient for thermal cycling to occur. Rather a swirling flow structure, representative of a large, 'corkscrew' vortical structure, is required for thermal cycling. Scale tests on the UH configuration have simulated cycling phenomena observed in full-scale plant data and have been used to determine parametric sensitivities in formulating a predictive model for the thermal cycling. Data indicate that the mechanism for thermal cycling in UH configurations is stochastic but scales with the leak rate from the valve. The critical dependent variables are reduced to several non-dimensional scaling curves, resulting in a semiempirical predictive model. This paper discusses the test program and the results obtained to date. Application of these

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Identification and evaluation of PWR in-vessel severe accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. CFD study of isothermal water flow in rod bundle with split-type spacer grid

    Science.gov (United States)

    Batta, A.; Class, A. G.

    2014-06-01

    The design of rod bundles in nuclear application nowadays is assessed by CFD (computational fluid dynamics). The accuracy of CFD models need validation. Within the OECD/NEA benchmark MATiS-H (Measurement and Analysis of Turbulent Mixing in Sub-channels - Horizontal) a single-phase water flow in a 5x5 rod bundle is studied. In the benchmark, two types of spacer grids are tested, the swirl type and the split type, where the current study focuses on the split type spacer grid. Comparison of CFD results obtained at Karlsruhe Institut of Technology (KIT) with experimental results of KAERI (Korea Atomic Energy Research Institute) are presented. In the benchmark velocities components along selected lines downstream of the spacer grid are measured and compared to CFD results. The CFD code STAR CCM+ with the Realized k-ɛ model is used. Comparisons with experimental results show quantitative and qualitative agreement for the averaged values of velocity components. Comparisons of results to other benchmark partners using different modeling show that the selected mesh size and models for the analysis of the current case gives relatively accurate results. However, the used turbulent model (Realized k-ɛ does not capture the turbulent intensity correctly. Computation shows that the flow has very high mixing due to the spacer grid, which does not decay within the measurements domain (z/ DH =0-10 downstream of spacer grid). The same conclusion can be drawn from experimental data.

  19. DESCRIPTION OF THE TRITIUM-PRODUCING BURNABLE ABSORBER ROD FOR THE COMMERCIAL LIGHT WATER REACTOR TTQP-1-015 Rev 19

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A.; Love, Edward F.; Thornhill, Cheryl K.

    2012-02-01

    Tritium-producing burnable absorber rods (TPBARs) used in the U.S. Department of Energy’s Tritium Readiness Program are designed to produce tritium when placed in a Westinghouse or Framatome 17x17 fuel assembly and irradiated in a pressurized water reactor (PWR). This document provides an unclassified description of the current design baseline for the TPBARs. This design baseline is currently valid only for Watts Bar reactor production cores. A description of the Lead Use TPBARs will not be covered in the text of the document, but the applicable drawings, specifications and test plan will be included in the appropriate appendices.

  20. Control rods in LMFBRs: a physics assessment

    Energy Technology Data Exchange (ETDEWEB)

    McFarlane, H.F.; Collins, P.J.

    1982-08-01

    This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B/sub 4/C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined.

  1. Solid-state-laser-rod holder

    Science.gov (United States)

    Gettemy, D.J.; Barnes, N.P.; Griggs, J.E.

    1981-08-11

    The disclosure relates to a solid state laser rod holder comprising Invar, copper tubing, and epoxy joints. Materials and coefficients of expansion of the components of the holder combine with the rod to produce a joint which will give before the rod itself will. The rod may be lased at about 70 to 80/sup 0/K and returned from such a temperature to room temperature repeatedly without its or the holder's destruction.

  2. 21 CFR 876.4270 - Colostomy rod.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out...

  3. Solitary waves on nonlinear elastic rods. II

    DEFF Research Database (Denmark)

    Sørensen, Mads Peter; Christiansen, Peter Leth; Lomdahl, P. S.

    1987-01-01

    In continuation of an earlier study of propagation of solitary waves on nonlinear elastic rods, numerical investigations of blowup, reflection, and fission at continuous and discontinuous variation of the cross section for the rod and reflection at the end of the rod are presented. The results...

  4. Phase behavior of colloidal silica rods

    NARCIS (Netherlands)

    Kuijk, A.; Byelov, D.; Petukhov, A.V.; van Blaaderen, A.; Imhof, A.

    2012-01-01

    Recently, a novel colloidal hard-rod-like model system was developed which consists of silica rods [Kuijk et al., JACS, 2011, 133, 2346]. Here, we present a study of the phase behavior of these rods, for aspect ratios ranging from 3.7 to 8.0. By combining real-space confocal laser scanning microscop

  5. Hydraulic Actuator for Ganged Control Rods

    Science.gov (United States)

    Thompson, D. C.; Robey, R. M.

    1986-01-01

    Hydraulic actuator moves several nuclear-reactor control rods in unison. Electromagnetic pump pushes liquid lithium against ends of control rods, forcing them out of or into nuclear reactor. Color arrows show lithium flow for reactor startup and operation. Flow reversed for shutdown. Conceived for use aboard spacecraft, actuator principle applied to terrestrial hydraulic machinery involving motion of ganged rods.

  6. Development of a program for the analysis on the free vibration of a fuel rod and its application

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seung; Yim, Jeong Sik [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Commercial Nuclear fuel burns more than 2 or three years in a core and it is essential that the fuels have a integrity without any failures during the burnup period. The factors that influence on the fuel integrity are classified as nuclear, mechanical, thermal and material factors and they are inter-related with complexity. Since the final integrity should be assured mechanically, the evaluation of the fuel rod mechanical integrity is important in a fuel design. The fuel rod for PWR is supported by spring of spacer grids to maintain its axial location and lateral space between fuel rods to get proper functions during the residence in a reactor. The long exposure duration makes the spring to be relax and loss the spring force that results in a fuel rod rattling which may cause fuel rod failure. The design criteria of the spring forces for various fuel vendors are similar each other but they are slightly different: require minimal spring force to prevent the spring from rattling at the end of life or the minimal gap between fuel rod and spring. However the spring force is relaxed due to the neutron irradiation and stress relaxation that suddenly decrease exponentially and the spring behave nonlinear by the initial spring deflection and the relaxation phenomenon. The objective of this study is to develop a finite element program to support the mechanical evaluation in view of the interaction between fuel rod and spacer spring. Here considering the spring behaviour as a function of burnup, the reaction forces of the springs are calculated by the finite element program, BEVIRA developed herein to aid the evaluation of the integrity of the fuel rod from fretting. A fuel rod is modelled as a beam to get natural frequencies and mode shapes supported by a rotational spring at each spacer spring. The results from the program are compared with previous data and those from ANSYS for the validation of the program and procedures. For the example calculation, the characteristics

  7. Study of power peak migration due to insertion of control bars in a PWR reactor; Estudo da migracao do pico de potencia em funcao da insercao das barras de controle em um reator refrigerado a agua

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Costa, Danilo Leite; Borges, Diogo da Silva; Lava, Deise Diana; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: danilolc26@gmail.com, E-mail: diogosb@outlook.com, E-mail: deisedy@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown.

  8. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Science.gov (United States)

    Young, J.N.

    1958-04-22

    An electromagnetic apparatus for moving a rod-like member in small steps in either direction is described. The invention has particular application in the reactor field where the reactor control rods must be moved only a small distance and where the use of mechanical couplings is impractical due to the high- pressure seals required. A neutron-absorbing rod is mounted in a housing with gripping uaits that engage the rod, and coils for magnetizing the gripping units to make them grip, shift, and release the rod are located outside the housing.

  9. Exploiting rod technology. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-06-01

    ROD development was proceeding apace until recent budgetary decisions caused funding support for ROD development to be drastically reduced. The funding which was originally provided by DARPA and the Balanced Technology Initiative (BTI) Office has been cut back to zero from $800K. To determine the aeroballistic coefficients of a candidate dart, ARDEC is currently supporting development out of its own 6.2 funds at about $100K. ARDEC has made slow progress toward achieving this end because of failures in the original dart during testing. It appears that the next dart design to be tested will diverge from the original concept visualized by DARPA and Science and Technology Associates (STA). STA, the design engineer, takes exception to these changes on the basis of inappropriate test conditions and insufficient testing. At this time, the full resolution of this issue will be difficult because of the current management structure, which separates the developer (ARDEC) from the designer (STA).

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

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    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

  11. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  12. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  13. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Directory of Open Access Journals (Sweden)

    Pavel N. Alekseev

    2012-01-01

    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

  14. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  15. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  16. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  17. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  18. SCC crack growth rate of cold-worked austenitic stainless steels in PWR primary water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Guerre, C.; Raquet, O.; Herms, E. [Commissariat a l' Energie Atomique (CEA), DEN/DPC/SCCME/LECA, Gif-sur-Yvette Cedex (France); Marie, S. [Commissariat a l' Energie Atomique (CEA), DEN/DM2S/SEMT/LISN, Gif-sur-Yvette Cedex (France); Le Calvar, M. [Inst. for Radiological Protection and Nuclear Safety (IRSN), DSR/SAMS, Fontenay-aux-Roses Cedex (France)

    2007-07-01

    Stress corrosion cracking (SCC) of stainless steels (SS) is a significant cause of failure in the pressurized water reactors (PWR). Most of the reported case history failures of SS in PWR can be attributed to pollutants (chloride, sulphate) and / or locally oxygenated environments, even to sensitisation of the SS. However, some failures have been attributed to heavy cold work (CW) of SS. In laboratory tests, SCC initiation of cold-worked SS has been obtained using slow strain rate tests (SSRT) in nominal PWR environment. This paper describes constant load and cyclic crack growth rate (CGR) tests on cold-worked SS, on CT specimens. 304L and 316L have been tested with a CW up to 60 %. CW 316L is more prone to cracking than 304L. Over 30 % of CW, 316L is susceptible to crack propagation under constant load. CW is the main controlling parameter for cracking. (author))

  19. High temperature control rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Vollman, Russell E. (Solana Beach, CA)

    1991-01-01

    A high temperature nuclear control rod assembly comprises a plurality of substantially cylindrical segments flexibly joined together in succession by ball joints. The segments are made of a high temperature graphite or carbon-carbon composite. The segment includes a hollow cylindrical sleeve which has an opening for receiving neutron-absorbing material in the form of pellets or compacted rings. The sleeve has a threaded sleeve bore and outer threaded surface. A cylindrical support post has a threaded shaft at one end which is threadably engaged with the sleeve bore to rigidly couple the support post to the sleeve. The other end of the post is formed with a ball portion. A hollow cylindrical collar has an inner threaded surface engageable with the outer threaded surface of the sleeve to rigidly couple the collar to the sleeve. the collar also has a socket portion which cooperates with the ball portion to flexibly connect segments together to form a ball and socket-type joint. In another embodiment, the segment comprises a support member which has a threaded shaft portion and a ball surface portion. The threaded shaft portion is engageable with an inner threaded surface of a ring for rigidly coupling the support member to the ring. The ring in turn has an outer surface at one end which is threadably engageably with a hollow cylindrical sleeve. The other end of the sleeve is formed with a socket portion for engagement with a ball portion of the support member. In yet another embodiment, a secondary rod is slidably inserted in a hollow channel through the center of the segment to provide additional strength. A method for controlling a nuclear reactor utilizing the control rod assembly is also included.

  20. Topological Mixing with Ghost Rods

    OpenAIRE

    2005-01-01

    Topological chaos relies on the periodic motion of obstacles in a two-dimensional flow in order to form nontrivial braids. This motion generates exponential stretching of material lines, and hence efficient mixing. Boyland et al. [P. L. Boyland, H. Aref, and M. A. Stremler, J. Fluid Mech. 403, 277 (2000)] have studied a specific periodic motion of rods that exhibits topological chaos in a viscous fluid. We show that it is possible to extend their work to cases where the motion of the stirring...

  1. Reactor control rod timing system. [LMFBR

    Science.gov (United States)

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  2. Automatic safety rod for reactors. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-03-23

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  3. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  4. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  5. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  6. Analyses of PWR boron dilution consequences with the Arrotta code

    Energy Technology Data Exchange (ETDEWEB)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  7. Continuous firefly algorithm applied to PWR core pattern enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Poursalehi, N., E-mail: npsalehi@yahoo.com [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)

    2013-05-15

    Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K{sub eff}) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable.

  8. French nuclear plants PWR vessel integrity assessment and life management

    Energy Technology Data Exchange (ETDEWEB)

    Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

    2001-07-01

    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  9. Acoustic loading effects on oscillating rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Lin, W.H.

    1980-01-01

    An analytical study of the interaction between an infinite acoustic medium and a cluster of circular rods is described. The acoustic field due to oscillating rods and the acoustic loading on the rods are first solved in a closed form. The acoustic loading is then used as a forcing function for rod responses, and the acousto-elastic couplings are solved simultaneously. Numerical examples are presented for several cases to illustrate the effects of various system parameters on the acoustic reaction force coefficients. The effect of the acoustic loading on the coupled eigenfrequencies are discussed.

  10. Stress corrosion cracking in the vessel closure head penetrations of French PWR`s; Fissuration par corrosion sous contrainte de penetrations de couvercle de cuve de reacteur nucleaire francais a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Buisine, D.; Cattant, F.; Champredonde, J.; Pichon, C.; Benhamou, C.; Gelpi, A.; Vaindirlis, M.

    1994-01-01

    During a hydrotest in September 1991, part of the statutory decennial in-service inspection, a leak was detected on the vessel head of Bugey 3, which is one of the first 900 MW 3-loop PWR`s in France. This leak was due to a cracked penetration used for a control rod drive mechanism. The investigations performed identified Primary Stress Corrosion Cracking of Alloy 600 as being the origin of this degradation. So a lot of the same design PWR`s are a concern due to this generic problem. In this case, PWSCC was linked to: - hot temperature of the vessel head; - high residual stresses due to the welding process between peripherical penetrations and the vessel head; - sensitivity of forged Alloy 600 used for penetration manufacturing. This following paper will present the cracked analysis based, in particular, on the main results obtained in France on each of these items. These results come from the operating experience, the destructive examinations and the programs which are running on stress analysis and metallurgical characterizations. (authors). 9 figs., 2 tabs.

  11. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has

  12. Development of a digital reactivity meter for criticality prediction and control rod worth evaluation in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuramoto, Renato Y.R.; Miranda, Anselmo F.; Valladares, Gastao Lommez; Prado, Adelk C. [Eletrobras Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil). Central Nuclear Almirante Alvaro Alberto], e-mail: kuramot@eletronuclear.gov.br

    2009-07-01

    In this work, we have proposed the development of a digital reactivity meter in order to monitor subcriticality continuously during criticality approach in a PWR. A subcritical reactivity meter can provide an easy prediction of the estimated critical point prior to reactor criticality, without complicated hand calculation. Moreover, in order to reduce the interval of the Physics Tests from the economical point of view, a subcritical reactivity meter can evaluate the control rod worth from direct subcriticality measurement. In other words, count rate of Source Range (SR) detector recorded during the criticality approach could be used for subcriticality evaluation or control rod worth evaluation. Basically, a digital reactivity meter is based on the inverse solution of the kinetic equations of a reactor with the external neutron source in one-point reactor model. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. When the Inverse Kinetic method is applied to a sufficiently high power level or to a core without an external neutron source, the neutron source term may be neglected. When applied to a lower power level or in the sub-critical domain, however, the source effects must be taken in account. Furthermore, some treatments are needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we have proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements in PWR. (author)

  13. Viscoelasticity of suspensions of long, rigid rods

    NARCIS (Netherlands)

    Dhont, Jan K.G.; Briels, W.J.

    2003-01-01

    A microscopic theory for the viscoelastic behaviour of suspensions of rigid rods with excluded volume interactions is presented, which is valid in the asymptotic limit of very long and thin rods. Stresses arising from translational and rotational Brownian motion and direct interactions are calculate

  14. Study of the rod style SFRFQ structure

    CERN Document Server

    Yan Xue Qing; Chen J

    2002-01-01

    There is a problem about upper limit of energy in the RFQ structure, although it is a wonderful low-energy-suited high current accelerating structure. After proposing an improved rod style SFRFQ structure without reversed field, the author studies its energy gain and transverse motion. The rod style SFRFQ structure is roughly compared with diaphragm SFRFQ structure

  15. Effects of axial power shapes on CHF locations in a single tube and in rod bundle assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, B.; Yang, B.W.; Zhang, H.; Zha, Y.; Zhang, Y. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology

    2016-07-15

    Currently, the prediction of rod bundle CHF is dependent on CHF correlations derived from CHF data. A simple correction factor, such as F-factor, is often used to account for the axial power shape differences based on a simple accumulated energy concept, which has totally no consideration on the impact of true local condition on CHF mechanism. Subsequently, as expected, large uncertainty is often associated with the CHF value and CHF location predictions. For the purpose of obtaining different power shapes effects on CHF, CFD calculated parameter values were used to predict the possible CHF occurrence location. The possible CHF location prediction method proposed in this paper is calculated void fraction, heat transfer coefficient (HTC), liquid temperature distribution and detailed local parameters. And the uniform and non-uniform CHF were analyzed. The prediction of possible CHF locations in a 5 x 5 rod bundle may provide useful information for the design of a full-length CHF test, enhance the accuracy of CHF and CHF location prediction, and reduce the costs of the experimentation.

  16. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  17. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    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 stor

  18. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    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 stor

  19. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables. (ACR)

  20. Nrl is required for rod photoreceptor development.

    Science.gov (United States)

    Mears, A J; Kondo, M; Swain, P K; Takada, Y; Bush, R A; Saunders, T L; Sieving, P A; Swaroop, A

    2001-12-01

    The protein neural retina leucine zipper (Nrl) is a basic motif-leucine zipper transcription factor that is preferentially expressed in rod photoreceptors. It acts synergistically with Crx to regulate rhodopsin transcription. Missense mutations in human NRL have been associated with autosomal dominant retinitis pigmentosa. Here we report that deletion of Nrl in mice results in the complete loss of rod function and super-normal cone function, mediated by S cones. The photoreceptors in the Nrl-/- retina have cone-like nuclear morphology and short, sparse outer segments with abnormal disks. Analysis of retinal gene expression confirms the apparent functional transformation of rods into S cones in the Nrl-/- retina. On the basis of these findings, we postulate that Nrl acts as a 'molecular switch' during rod-cell development by directly modulating rod-specific genes while simultaneously inhibiting the S-cone pathway through the activation of Nr2e3.

  1. Angra-1 reactor core simulation with reduced diameter fuel rods; Simulacao do nucleo de Angra-1 com combustiveis de menor diametro de vareta

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano M; Faria, Eduardo F.; Sakai, Massao; Gomes, Sydney da S. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil)

    2000-07-01

    From the neutronic point of view, it is advantageous to use fuel elements with narrower rod diameter at Angra-1 PWR, since the reactivity level increases, and that happens mainly for higher enrichments than the ones used up to now. This fact is due to the higher moderator/fuel ratio, leading to a stronger neutron thermalization. In order to quantify this effect, the nodal core MEDIUM/SAV90 has been employed to simulate Angra-1 cycles from the present until the equilibrium cycle. The actual fuel element design has been maintained in this report, with exception of fuel rods diameter, reduced to 9 mm. Results have shown a higher reactivity and final burnup for the reduced diameter fuel rods, producing less waste for final disposal. However, the combined effect of higher elements reactivity and burnup made difficult the cycle-by-cycle fuel reload optimization. Preliminary results show possible advantages of using reduced diameter fuel rods in reload schemes type 'stop and go', but not being recommendable for extended cycles. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. Engineering design feasibility of low boron concentration core in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, A. T.; Kim, M. H. [Kyung Hee University, Yongin-shi, Gyeonggi-do, 446-701 Republic of Korea (Korea, Republic of); Woo, I.; Shon, S. R., E-mail: atdaing@khu.ac.k [Korea Nuclear Fuel, 1047 Daedukdaero, Yuseong-gu, Daejeon, 305-353 Republic of Korea (Korea, Republic of)

    2010-10-15

    In pressurized water reactor operation, higher level of soluble boron concentration could contribute higher impact from boron dilution situations, higher amount of liquid waste, and higher radiation dose to operators from higher corrosion potential to cladding and structure. Two practical and feasible means to reduce the maximum boron concentration were investigated in this study. A technically straightforward, possible means, can be achieved either by implementation of enriched boric acid (Eba) or by increasing more shim rod (fixed burnable absorber) worth. A simplest option is that the Eba is applied into reference core (Ref) design, OPR-1000 design, Ulchin unit-5 by allowing use of same fuel assemblies and core design without changing any nuclear design methodology used in that Ref design. Although results of Eba option proved its favorable power distribution and peaking factor, its moderator temperature coefficient (MTC) value reached positive, 3.25 pcm/ C at 40 EFPD which is beyond the design safety limit. An alternative option with more shim rods in fuel assemblies was tried with four types of integral burnable absorbers: gadolinia, integral fuel burnable absorber (Ifba), erbium and alumina boron carbide. Four core design candidates have been developed by keeping major engineering designs and preserving equivalent fuel enrichment level used in Ref design. However, all optimal designs were targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between Ref and optimal core designs is presented here. One of them is suggested as the most promising and favorable low boron core (Lbc) design in this framework. The proper combination of axial and radial enrichment zoning pattern in Lbc design candidate with Ifba-bearing fuel assemblies at equilibrium cycle, could bring 2 times narrower axial offset variation than that of Ref design, and maintain acceptable power peaking factor around 23% lower than

  4. Phase behavior and structure formation of hairy-rod supramolecules

    NARCIS (Netherlands)

    Subbotin, A; Stepanyan, R; Knaapila, M; Ikkala, O; ten Brinke, G

    2003-01-01

    Phase behavior and microstructure formation of rod and coil molecules, which can associate to form hairy-rod polymeric supramolecules, are addressed theoretically. Association induces considerable compatibility enhancement between the rod and coil molecules and various microscopically ordered struct

  5. Influence of FIMA burnup on actinides concentrations in PWR reactors

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

  6. Eulerian Formulation of Spatially Constrained Elastic Rods

    Science.gov (United States)

    Huynen, Alexandre

    Slender elastic rods are ubiquitous in nature and technology. For a vast majority of applications, the rod deflection is restricted by an external constraint and a significant part of the elastic body is in contact with a stiff constraining surface. The research work presented in this doctoral dissertation formulates a computational model for the solution of elastic rods constrained inside or around frictionless tube-like surfaces. The segmentation strategy adopted to cope with this complex class of problems consists in sequencing the global problem into, comparatively simpler, elementary problems either in continuous contact with the constraint or contact-free between their extremities. Within the conventional Lagrangian formulation of elastic rods, this approach is however associated with two major drawbacks. First, the boundary conditions specifying the locations of the rod centerline at both extremities of each elementary problem lead to the establishment of isoperimetric constraints, i.e., integral constraints on the unknown length of the rod. Second, the assessment of the unilateral contact condition requires, in principle, the comparison of two curves parametrized by distinct curvilinear coordinates, viz. the rod centerline and the constraint axis. Both conspire to burden the computations associated with the method. To streamline the solution along the elementary problems and rationalize the assessment of the unilateral contact condition, the rod governing equations are reformulated within the Eulerian framework of the constraint. The methodical exploration of both types of elementary problems leads to specific formulations of the rod governing equations that stress the profound connection between the mechanics of the rod and the geometry of the constraint surface. The proposed Eulerian reformulation, which restates the rod local equilibrium in terms of the curvilinear coordinate associated with the constraint axis, describes the rod deformed configuration

  7. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  8. Morphoelastic rods Part II: Growing birods

    Science.gov (United States)

    Lessinnes, Thomas; Moulton, Derek E.; Goriely, Alain

    2017-03-01

    The general problem of determining the shape and response of two attached growing elastic Kirchhoff rods is considered. A description of the kinematics of the individual interacting rods is introduced. Each rod has a given intrinsic shape and constitutive laws, and a map associating points on the two rods is defined. The resulting filamentary structure, a growing birod, can be seen as a new filamentary structure. This kinematic description is used to derive the general equilibrium equations for the shape of the rods under loads, or equivalently, for the new birod. It is shown that, in general, the birod is not simply a Kirchhoff rod but rather, due to the internal constraints, new effects can appear. The two-dimensional restriction is then considered explicitly and the limit for small deformation is shown to be equivalent to the classic Timsohenko bi-metallic strip problem. A number of examples and applications are presented. In particular, the problem of two attached rods with intrinsic helical shape and uniform growth is computed in detail and a host of new interesting solutions and bifurcations are observed.

  9. Granular materials interacting with thin flexible rods

    Science.gov (United States)

    Neto, Alfredo Gay; Campello, Eduardo M. B.

    2017-04-01

    In this work, we develop a computational model for the simulation of problems wherein granular materials interact with thin flexible rods. We treat granular materials as a collection of spherical particles following a discrete element method (DEM) approach, while flexible rods are described by a large deformation finite element (FEM) rod formulation. Grain-to-grain, grain-to-rod, and rod-to-rod contacts are fully permitted and resolved. A simple and efficient strategy is proposed for coupling the motion of the two types (discrete and continuum) of materials within an iterative time-stepping solution scheme. Implementation details are shown and discussed. Validity and applicability of the model are assessed by means of a few numerical examples. We believe that robust, efficiently coupled DEM-FEM schemes can be a useful tool to the simulation of problems wherein granular materials interact with thin flexible rods, such as (but not limited to) bombardment of grains on beam structures, flow of granular materials over surfaces covered by threads of hair in many biological processes, flow of grains through filters and strainers in various industrial segregation processes, and many others.

  10. Magnetically controlled growing rods for scoliosis surgery.

    Science.gov (United States)

    Metkar, Umesh; Kurra, Swamy; Quinzi, David; Albanese, Stephen; Lavelle, William F

    2017-02-01

    Early onset scoliosis can be both a disfiguring as well as a life threatening condition. When more conservative treatments fail, pediatric spinal surgeons are forced to consider operative interventions. Traditionally, these interventions have involved the insertion of a variety of implants into the patient with a limited number of anchor points controlling the spine. In the past, these pediatric patients have had multiple surgeries for elective lengthening of these devices to facilitate their growth while attempting to control the scoliosis. These patients often experience a physical and emotional toll from their multiple repeated surgeries. Growing spine techniques have also had a noted high complication rate due to implant dislodgement and infections. Recently, the development of non-invasively, self-lengthening growing rods has occurred. These devices have the potential to allow for the devices to be lengthened magnetically in a conscious patient in the surgeon's office. Areas covered: This review summarized previously published articles in the English literature using a key word search in PubMed for: 'magnetically controlled growing rods', 'Magec rods', 'magnetic growing rods' and 'growing rods'. Expert commentary: Magnetically controlled growing rods have an advantage over growing rods in lengthening the growing spine in the absence of repetitive surgeries.

  11. Control Rod Malfunction at the NRAD Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  12. Estimation of irradiated control rod worth

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Antonopoulos-Domis, M. [School of Electrical and Computer Engineering, Aristotle University of Thessaloniki, Thessaloniki (Greece)

    2009-11-15

    When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

  13. Granular materials interacting with thin flexible rods

    Science.gov (United States)

    Neto, Alfredo Gay; Campello, Eduardo M. B.

    2016-01-01

    In this work, we develop a computational model for the simulation of problems wherein granular materials interact with thin flexible rods. We treat granular materials as a collection of spherical particles following a discrete element method (DEM) approach, while flexible rods are described by a large deformation finite element (FEM) rod formulation. Grain-to-grain, grain-to-rod, and rod-to-rod contacts are fully permitted and resolved. A simple and efficient strategy is proposed for coupling the motion of the two types (discrete and continuum) of materials within an iterative time-stepping solution scheme. Implementation details are shown and discussed. Validity and applicability of the model are assessed by means of a few numerical examples. We believe that robust, efficiently coupled DEM-FEM schemes can be a useful tool to the simulation of problems wherein granular materials interact with thin flexible rods, such as (but not limited to) bombardment of grains on beam structures, flow of granular materials over surfaces covered by threads of hair in many biological processes, flow of grains through filters and strainers in various industrial segregation processes, and many others.

  14. Tipping time of a quantum rod

    Energy Technology Data Exchange (ETDEWEB)

    Parrikar, Onkar [Birla Institute of Technology and Science-Pilani, Goa campus, Zuarinagar, Goa 4032726 (India)], E-mail: onkarsp@gmail.com

    2010-03-15

    The behaviour of a quantum rod, pivoted at its lower end on an impenetrable floor and restricted to moving in the vertical plane under the gravitational potential, is studied analytically under the approximation that the rod is initially localized to a 'small-enough' neighbourhood around the point of classical unstable equilibrium. It is shown that the rod evolves out of this neighbourhood. The time required for this to happen, i.e. the tipping time, is calculated using the semi-classical path integral. It is shown that equilibrium is recovered in the classical limit, and that our calculations are consistent with the uncertainty principle.

  15. High temperature control rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Vollman, R.E.

    1991-12-24

    This patent describes a control rod assembly for use in nuclear reactor control. It comprises segments, each the segment being made of a graphite composite material, each the segment having a chamber for containing neutron-absorbing material, wherein the chamber compromises a hollow cylindrical sleeve having a first end formed with an opening for receiving the neutron-absorbing material, and having a second end formed with a sleeve bore and an outer sleeve surface; a cylindrical weight-bearing support post positioned substantially centrally of the sleeve, the support post having a first end formed as a ball surface portion and a second end formed as a ball surface portion and a second end formed as a shaft, the shaft being engageable with the sleeve bore for rigidly coupling the support post axially within the hollow sleeve, a hollow cylindrical collar having a socket lip portion correspondingly shaped to receive the ball surface portion of an adjacent support post, and having an inner surface for engaging the outer sleeve surface on the second end of the sleeve to rigidly couple the collar to the sleeve.

  16. The Power-weakness Ratios (PWR as a Journal Indicator: Testing the “Tournaments” Metaphor in Citation Impact Studies

    Directory of Open Access Journals (Sweden)

    Loet Leydesdorff

    2016-09-01

    Full Text Available Purpose: Ramanujacharyulu developed the Power-weakness Ratio (PWR for scoring tournaments. The PWR algorithm has been advocated (and used for measuring the impact of journals. We show how such a newly proposed indicator can empirically be tested. Design/methodology/approach: PWR values can be found by recursively multiplying the citation matrix by itself until convergence is reached in both the cited and citing dimensions; the quotient of these two values is defined as PWR. We study the effectiveness of PWR using journal ecosystems drawn from the Library and Information Science (LIS set of the Web of Science (83 journals as an example. Pajek is used to compute PWRs for the full set, and Excel for the computation in the case of the two smaller sub-graphs: (1 JASIST+ the seven journals that cite JASIST more than 100 times in 2012; and (2 MIS Quart+ the nine journals citing this journal to the same extent. Findings: A test using the set of 83 journals converged, but did not provide interpretable results. Further decomposition of this set into homogeneous sub-graphs shows that—like most other journal indicators—PWR can perhaps be used within homogeneous sets, but not across citation communities. We conclude that PWR does not work as a journal impact indicator; journal impact, for example, is not a tournament. Research limitations: Journals that are not represented on the “citing” dimension of the matrix—for example, because they no longer appear, but are still registered as “cited” (e.g. ARIST—distort the PWR ranking because of zeros or very low values in the denominator. Practical implications: The association of “cited” with “power” and “citing” with “weakness” can be considered as a metaphor. In our opinion, referencing is an actor category and can be Metaphor in Citation Impact Studies in terms of behavior, whereas “citedness” is a property of a document with an expected dynamics very different from that of

  17. Impact of AD995 alumina rods

    Energy Technology Data Exchange (ETDEWEB)

    Chhabildas, L.C.; Furnish, M.D.; Reinhart, W.D. [Sandia National Labs., Albuquerque, NM (United States); Grady, D.E. [Applied Research Associates, Inc., Albuquerque, NM (United States)

    1997-10-01

    Gas guns and velocity interferometric techniques have been used to determine the loading behavior of an AD995 alumina rod 19 mm in diameter by 75 mm and 150 mm long, respectively. Graded-density materials were used to impact both bare and sleeved alumina rods while the velocity interferometer was used to monitor the axial-velocity of the free end of the rods. Results of these experiments demonstrate that (1) a time-dependent stress pulse generated during impact allows an efficient transition from the initial uniaxial strain loading to a uniaxial stress state as the stress pulse propagates through the rod, and (2) the intermediate loading rates obtained in this configuration lie between split Hopkinson bar and shock-loading techniques.

  18. Computer simulation of rod-sphere mixtures

    CERN Document Server

    Antypov, D

    2003-01-01

    Results are presented from a series of simulations undertaken to investigate the effect of adding small spherical particles to a fluid of rods which would otherwise represent a liquid crystalline (LC) substance. Firstly, a bulk mixture of Hard Gaussian Overlap particles with an aspect ratio of 3:1 and hard spheres with diameters equal to the breadth of the rods is simulated at various sphere concentrations. Both mixing-demixing and isotropic-nematic transition are studied using Monte Carlo techniques. Secondly, the effect of adding Lennard-Jones particles to an LC system modelled using the well established Gay-Berne potential is investigated. These rod-sphere mixtures are simulated using both the original set of interaction parameters and a modified version of the rod-sphere potential proposed in this work. The subject of interest is the internal structure of the binary mixture and its dependence on density, temperature, concentration and various parameters characterising the intermolecular interactions. Both...

  19. Bouncing Balls and Hot Rod Races.

    Science.gov (United States)

    Tibbs, Peggy; Sherrill, Donna

    This paper presents the Bouncing Ball Experiment which models quadratic and exponential functions, and the Hot Rod Races activity that explores velocity and acceleration. Activities include directions for the use of TI-82 and TI-83 calculators. (YDS)

  20. Double-clad nuclear fuel safety rod

    Science.gov (United States)

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  1. Microelectrophoresis of Silica Rods Using Confocal Microscopy.

    Science.gov (United States)

    Bakker, Henriëtte E; Besseling, Thijs H; Wijnhoven, Judith E G J; Helfferich, Peter H; van Blaaderen, Alfons; Imhof, Arnout

    2017-01-31

    The electrophoretic mobility and the zeta potential (ζ) of fluorescently labeled colloidal silica rods, with an aspect ratio of 3.8 and 6.1, were determined with microelectrophoresis measurements using confocal microscopy. In the case where the colloidal particles all move at the same speed parallel to the direction of the electric field, we record a xyz-stack over the whole depth of the capillary. This method is faster and more robust compared to taking xyt-series at different depths inside the capillary to obtain the parabolic flow profile, as was done in previous work from our group. In some cases, rodlike particles do not move all at the same speed in the electric field, but exhibit a velocity that depends on the angle between the long axis of the rod and the electric field. We measured the orientation-dependent velocity of individual silica rods during electrophoresis as a function of κa, where κ(-1) is the double layer thickness and a is the radius of the rod associated with the diameter. Thus, we determined the anisotropic electrophoretic mobility of the silica rods with different sized double layers. The size of the double layer was tuned by suspending silica rods in different solvents at different electrolyte concentrations. We compared these results with theoretical predictions. We show that even at already relatively high κa when the Smoluchowski limiting law is assumed to be valid (κa > 10), an orientation dependent velocity was measured. Furthermore, we observed that at decreasing values of κa the anisotropy in the electrophoretic mobility of the rods increases. However, in low polar solvents with κa < 1, this trend was reversed: the anisotropy in the electrophoretic mobility of the rods decreased. We argue that this decrease is due to end effects, which was already predicted theoretically. When end effects are not taken into account, this will lead to strong underestimation of the experimentally determined zeta potential.

  2. High Power Performance of Rod Fiber Amplifiers

    DEFF Research Database (Denmark)

    Johansen, Mette Marie; Michieletto, Mattia; Kristensen, Torben

    2015-01-01

    An improved version of the DMF rod fiber is tested in a high power setup delivering 360W of stable signal power. Multiple testing degrades the fiber and transverse modal instability threshold from >360W to ~290W.......An improved version of the DMF rod fiber is tested in a high power setup delivering 360W of stable signal power. Multiple testing degrades the fiber and transverse modal instability threshold from >360W to ~290W....

  3. IMPACT CONICAL ROD ON HARD LIMITER

    Directory of Open Access Journals (Sweden)

    Ulitin G.

    2014-12-01

    Full Text Available The problem is considered of longitudinal impact conical rod in article. A recommendation on the use of the approximate method of calculation is based on an analysis of the influence of design parameters on the value of the main oscillation frequency. There was obtained an equation of the displacement and stress of the rod. Engineering dependence has been proposed to determine the maximum force in the impact section.

  4. Self-diagnosing braided composite rod

    OpenAIRE

    Fangueiro, Raúl; Zdraveva, E.; Pereira, Cristiana Gonilho; Ferreira, A; Lanceros-Méndez, S.

    2010-01-01

    This paper presents the development of a braided reinforced composite rod (BCR) able to both reinforce and monitor the stress state of concrete structures. Carbon fibers have been used as sensing and reinforcing materials along with glass fiber. Various composites rods have been produced using an author patented technique based on a modified conventional braiding machine. The materials investigated were prepared with different carbon fiber content as follows: BCR2 (77% glass/23...

  5. Measurements of control rod efficiency in RBMK critical assembly upon dropping of the rods

    Energy Technology Data Exchange (ETDEWEB)

    Zhitarev, V. E., E-mail: vejitarev@nnrd.kiae.su; Kachanov, V. M.; Sergevnin, A. Yu.; Lebedev, G. V., E-mail: lgv2004@mail.ru [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15

    The efficiency of control rods in the RBMK critical assembly was measured in the case where one manual-control rod (MCR) is dropped from a steady critical state, and several other MCRs were additionally dropped after 44 s. The measured number of neutrons in the assembly during and after dropping of the rods was used to calculate the efficiency values of the rods by solution of the system of point kinetics equations. A series of methods of the initial data treatment for determination of the desired values of reactivity without the calculated corrections were used.

  6. Measurements of control rod efficiency in RBMK critical assembly upon dropping of the rods

    Science.gov (United States)

    Zhitarev, V. E.; Kachanov, V. M.; Sergevnin, A. Yu.; Lebedev, G. V.

    2014-12-01

    The efficiency of control rods in the RBMK critical assembly was measured in the case where one manual-control rod (MCR) is dropped from a steady critical state, and several other MCRs were additionally dropped after 44 s. The measured number of neutrons in the assembly during and after dropping of the rods was used to calculate the efficiency values of the rods by solution of the system of point kinetics equations. A series of methods of the initial data treatment for determination of the desired values of reactivity without the calculated corrections were used.

  7. High-throughput rod-induced electrospinning

    Science.gov (United States)

    Wu, Dezhi; Xiao, Zhiming; Teh, Kwok Siong; Han, Zhibin; Luo, Guoxi; Shi, Chuan; Sun, Daoheng; Zhao, Jinbao; Lin, Liwei

    2016-09-01

    A high throughput electrospinning process, directly from flat polymer solution surfaces induced by a moving insulating rod, has been proposed and demonstrated. Different rods made of either phenolic resin or paper with a diameter of 1-3 cm and a resistance of about 100-500 MΩ, has been successfully utilized in the process. The rod is placed approximately 10 mm above the flat polymer solution surface with a moving speed of 0.005-0.4 m s-1 this causes the solution to generate multiple liquid jets under an applied voltage of 15-60 kV for the tip-less electrospinning process. The local electric field induced by the rod can boost electrohydrodynamic instability in order to generate Taylor cones and liquid jets. Experimentally, it is found that a large rod diameter and a small solution-to-rod distance can enhance the local electrical field to reduce the magnitude of the applied voltage. In the prototype setup with poly (ethylene oxide) polymer solution, an area of 5 cm  ×  10 cm and under an applied voltage of 60 kV, the maximum throughput of nanofibers is recorded to be approximately144 g m-2 h-1.

  8. Rigid rod anchored to infinite membrane.

    Science.gov (United States)

    Guo, Kunkun; Qiu, Feng; Zhang, Hongdong; Yang, Yuliang

    2005-08-15

    We investigate the shape deformation of an infinite membrane anchored by a rigid rod. The density profile of the rod is calculated by the self-consistent-field theory and the shape of the membrane is predicted by the Helfrich membrane elasticity theory [W. Helfrich, Z. Naturforsch. 28c, 693 (1973)]. It is found that the membrane bends away from the rigid rod when the interaction between the rod and the membrane is repulsive or weakly attractive (adsorption). However, the pulled height of the membrane at first increases and then decreases with the increase of the adsorption strength. Compared to a Gaussian chain with the same length, the rigid rod covers much larger area of the membrane, whereas exerts less local entropic pressure on the membrane. An evident gap is found between the membrane and the rigid rod because the membrane's curvature has to be continuous. These behaviors are compared with that of the flexible-polymer-anchored membranes studied by previous Monte Carlo simulations and theoretical analysis. It is straightforward to extend this method to more complicated and real biological systems, such as infinite membrane/multiple chains, protein inclusion, or systems with phase separation.

  9. Long-Rod Moving-Plate Interaction

    Science.gov (United States)

    Partom, Y.

    2002-07-01

    Understanding the mechanics of interaction of a long rod projectile with a forward moving plate at an angle is essential to understanding long rod interaction with an explosive reactive armor cassette. To investigate the mechanics of such an interaction we use AUTODIN2D/EULER in plane geometry, although the problem is 3D. We assume that this is a satisfactory approximation, as we're only interested in the main features, and are not comparing fine details to experimental results. From the simulations we learn that the interaction never reaches steady state. Initially each material splits into two streams, and the interaction plane is perpendicular to the rod. But with time the interaction plane rotates slowly, until it becomes parallel to the rod, which is then able to continue moving forward without interruption. During this process interacting rod material of length DeltaL is diverted at an angle and becomes ineffective for penetrating the main target. We made many such runs to determine the dependence of DeltaL on the parameters of the problem. This dependence makes it possible to predict DeltaL for a variety of rod-plate situations.

  10. Topological optimisation of rod-stirring devices

    CERN Document Server

    Finn, Matthew D

    2011-01-01

    There are many industrial situations where rods are used to stir a fluid, or where rods repeatedly stretch a material such as bread dough or taffy. The goal in these applications is to stretch either material lines (in a fluid) or the material itself (for dough or taffy) as rapidly as possible. The growth rate of material lines is conveniently given by the topological entropy of the rod motion. We discuss the problem of optimising such rod devices from a topological viewpoint. We express rod motions in terms of generators of the braid group, and assign a cost based on the minimum number of generators needed to write the braid. We show that for one cost function -- the topological entropy per generator -- the optimal growth rate is the logarithm of the golden ratio. For a more realistic cost function,involving the topological entropy per operation where rods are allowed to move together, the optimal growth rate is the logarithm of the silver ratio, $1+\\sqrt{2}$. We show how to construct devices that realise th...

  11. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  12. Model of fracture for the Zry cladding of nuclear fuel rods included in the code DIONISIO 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: soba@cnea.gov.ar; Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-12-15

    The DIONISIO code describes most of the main phenomena occurring in a fuel rod during normal operation of a nuclear power reactor. Starting from the irradiation history, the code predicts the temperature distribution, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the internal rod volume, gas mixing, pressure increase, irradiation growth of the cladding, development of an oxide layer on its surface and hydrogen uptake, restructuring and grain growth in the pellet. This work presents the model of Zircaloy fracture included in the code DIONISIO 1.0. The model of pellet-cladding mechanical interaction (PCMI) provides the forces caused by the solid-solid contact which add to the changing internal pressure and to the constant external pressure. Besides, the program evaluates the effects of a corrosive atmosphere (stress corrosion cracking, SCC) internal or external. With these data, the code calculates the J integral around the tip of an initiated crack, and proceeds to analyze, according to the quantity of corrosive substance dissolved and the cladding stress field, if the crack remains unchanged, if it grows due to the I-SCC mechanism, or if propagation is ductile, following the R curve of the material. Results corresponding to different PHWR and PWR reactors are presented and compared with code results. In particular, good agreement is obtained in the simulation of MOX experiments, where the cladding failed due to propagation of cracks originated in SCC.

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

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

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

  14. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  15. PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-10-01

    Full Text Available ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe.  Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor.  Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Data meteorologi diambil setiap jam untuk jangka waktu satu tahun. Hasil perhitungan menunjukkan bahwa frekuensi kumulatif  untuk model input yang sama untuk Tapak pesisir Serang lebih tinggi dibandingkan dengan Semenanjung Muria. Untuk tapak yang sama, frekuensi kumulatif model input cyclic lebih tinggi dibandingkan model stratified. Model cyclic memberikan keleluasan dalam menentukan tingkat ketelitian perhitungan dan tidak membutuhkan data acuan dibandingkan dengan model stratified. Penggunaan model cyclic dan stratified melibatkan jumlah data yang besar dan pengulangan perhitungan  akan meningkatkan  ketelitian nilai-nilai statistika perhitungan. Kata kunci: dampak kecelakaan, PWR 1000-MWe,  probabilistik,  atmosferik, PC-Cosyma   ABSTRACT THE INFLUENCE OF ATMOSPHERIC CONDITIONS TO PROBABILISTIC CALCULATION OF IMPACT OF RADIOLOGY ACCIDENT ON PWR-1000MWe. The calculation of the radiological impact of the fission products releases due to potential accidents

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

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Moustafa, S.; Plagne, L.; Poncot, A. [EDF R and D, 1, Av du General de Gaulle, F92141 Clamart cedex (France)

    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)

  17. EPRI PWR Safety and Relief Value Test Program: safety and relief valve test report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A safety and relief valve test program was conducted by EPRI for a group of participating PWR utilities to respond to the USNRC recommendations documented in NUREG 0578 Section 2.1.2, and as clarified in NUREG 0737 Item II.D.1.A. Seventeen safety and relief valves representative of those utilized in or planned for use in participating domestic PWR's were tested under the full range of selected test conditions. This report contains a listing of the selected test valves and the corresponding as tested test matrices, valve performance data and principal observations for the tested safety and relief valves. The information contained in this report may be used by the participating utilities in developing their response to the above mentioned USNRC recommendations.

  18. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  19. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  20. Anti -corrosion Effect of ETA on Materials in Secondary Loop of PWR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    In the world, over sixty percent of nuclear power plant have used advanced amunes ETA(Ethanolamine) as pH control agent in secondary loop of PWR. There are eighty percent of nuclear powerplants using ETA in USA. The corrosion of materials in steam generator (SG) tube and secondary looppower water reactor have been inhibited, the life of SG and the economics of the plant are increasedbecause of using ETA.

  1. Modeling and Simulation of Release of Radiation in Flow Blockage Accident for Two Loops PWR

    OpenAIRE

    Khurram Mehboob; Cao Xinrong; Majid Ali

    2012-01-01

    In this study modeling and simulation of release of radiation form two loops PWR has been carried out for flow blockage accident. For this purpose, a MATLAB based program “Source Term Evaluator for Flow Blockage Accident” (STEFBA) has been developed, which uses the core inventory as its primary input. The TMI-2 reactor is considered as the reference plant for this study. For 1100 reactor operation days, the core inventory has been evaluated under the core design constrains at average reactor ...

  2. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Energy Technology Data Exchange (ETDEWEB)

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)

    1997-07-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  3. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru (Univ. of Tokyo (Japan)); Inoue, Akira (Tokyo Institute of Technology (Japan)); Miyazaki, Keiji (Osaka Univ. (Japan)); Abeta, Sadaaki (Mitsubishi, Tokyo (Japan)); Hori, Keiichi (Mitsubishi, Hyogo (Japan)); Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki

    1990-01-01

    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data.

  4. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Energy Technology Data Exchange (ETDEWEB)

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)

    2010-07-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  5. A Comparative analysis for control rod drop accident in RETRAN DNB and CETOP DNB Model

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chang Keun; Kim, Yo Han; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. In this paper, RETRAN DNB Model and CETOP DNB Model were analyzed by using comparative method.

  6. Local Fuel Rod Crud Prediction Tool Applications

    Energy Technology Data Exchange (ETDEWEB)

    Krammen, Michael A.; Karoutas, Zeses E.; Wang, Guoqiang; Young, Michael Y

    2009-06-15

    A code system with attendant methods has been developed for modeling local fuel rod crud. This tool is used to perform the Crud Induced Localized Corrosion (CILC) risk assessment recommended by the EPRI crud and corrosion guidelines, which were developed in response to the INPO zero fuel failures by 2010 initiatives. The methodology is in production use. This paper will describe the range of problems the methodology has already been applied to and the especial pertinence to low duty fuel applications. The methodology begins with Computational Fluid Dynamics (CFD) computations over a fuel assembly grid span. The CFD results provide detailed relative variations in local heat transfer coefficient over the grid span. These very local relative variations are used to determine very local thermal hydraulic conditions over the entire axial length of every fuel rod in a reactor core over the life of the rod in reactor. The expansion using the local relative variations is currently accomplished with the HIDUTYDRV code. The very local thermal hydraulic conditions are combined with reactor coolant crud concentrations derived from EPRI BOA analysis as input to models for predicting very local fuel rod crud deposition. The reactor coolant crud concentrations are determined over each reactor cycle by reactor system wide crud mass balance calculations. The reactor coolant crud concentrations are used to calculate local crud thickness using mass transfer models which are a function of the local thermal conditions. The advanced crud deposition models also include models for calculating local crud dryout. Local crud deposition and crud dryout are strongly dependent on very local boiling or steaming, which are predicted through the translation of the CFD results. The local crud thickness and degree of local crud dryout are key factors in determining the margin or risk for local fuel rod cladding crud induced fuel failure. The development and first application of these methods was in

  7. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, C. [Dept. of Nuclear Engineering, Texas A and M Univ., 3133 TAMU, College Station, TX 77843-3133 (United States); Ilas, G. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Naegeli, Robert Earl

    2004-06-01

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

  9. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  10. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    Science.gov (United States)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  11. Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-10-20

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

  13. Regulatory perspective on incomplete control rod insertions

    Energy Technology Data Exchange (ETDEWEB)

    Chatterton, M.

    1997-01-01

    The incomplete control rod insertions experienced at South Texas Unit 1 and Wolf Creek are of safety concern to the NRC staff because they represent potential precursors to loss of shutdown margin. Even before it was determined if these events were caused by the control rods or by the fuel there was an apparent correlation of the problem with high burnup fuel. It was determined that there was also a correlation between high burnup and high drag forces as well as with rod drop time histories and lack of rod recoil. The NRC staff initial actions were aimed at getting a perspective on the magnitude of the problem as far as the number of plants and the amount of fuel that could be involved, as well as the safety significance in terms of shutdown margin. As tests have been performed and data has been analyzed the focus has shifted more toward understanding the problem and the ways to eliminate it. At this time the staff`s understanding of the phenomena is that it was a combination of factors including burnup, power history and temperature. The problem appears to be very sensitive to these factors, the interaction of which is not clearly understood. The model developed by Westinghouse provides a possible explanation but there is not sufficient data to establish confidence levels and sensitivity studies involving the key parameters have not been done. While several fixes to the problem have been discussed, no definitive fixes have been proposed. Without complete understanding of the phenomena, or fixes that clearly eliminate the problem the safety concern remains. The safety significance depends on the amount of shutdown margin lost due to incomplete insertion of the control rods. Were the control rods to stick high in the core, the reactor could not be shutdown by the control rods and other means such as emergency boration would be required.

  14. Bent Telescopic Rods in Patients With Osteogenesis Imperfecta.

    Science.gov (United States)

    Lee, R Jay; Paloski, Michael D; Sponseller, Paul D; Leet, Arabella I

    2016-09-01

    Telescopic rods require alignment of 2 rods to enable lengthening. A telescopic rod converts functionally into a solid rod if either rod bends, preventing proper engagement. Our goal was to characterize implant bending as a mode of failure of telescopic rods used in the treatment of osteogenesis imperfecta in children. We conducted a retrospective review of our osteogenesis imperfecta database for patients treated with intramedullary telescopic rods at our institution from 1992 through 2010 and identified 12 patients with bent rods. The 6 boys and 6 girls had an average age at the time of initial surgery of 3.1 years (range, 1.8 to 8.3 y) and a total of 51 telescoping rods. Clinic notes, operative reports, and radiographs were reviewed. The rods were analyzed for amount of lengthening, characteristics of bending, presence of cut out, or disengagement from an anchor point. Bends in the rods were characterized by their location on the implant component. The bent and straight rods were compared. Data were analyzed with the Mann-Whitney test (statistical significance set at P≤0.05). Of the 51 telescoping rods, 17 constructs (33%) bent. The average interval between surgery and rod bending was 4.0 years (range, 0.9 to 8.2 y). Before bending, 11 of 17 telescoping rods had routine follow-up radiographs for review. In 10 of the rods, bending was present when early signs of rod failure were first detected. Rod bending did not seem to be related to rod size. There was no area on the rod itself that seemed more susceptible to bending. Rod bending can be an early sign of impending rod failure. When rod bending is first noted, it may predispose the rod to other subsequent failures such as loss of proximal and distal fixation and cut out. Rod bending should be viewed as an indicator for closer monitoring of the patient and discussions regarding future need for rod exchange. Level III-retrospective review.

  15. Axial Vibration Confinement in Nonhomogenous Rods

    Directory of Open Access Journals (Sweden)

    S. Choura

    2005-01-01

    Full Text Available A design methodology for the vibration confinement of axial vibrations in nonhomogenous rods is proposed. This is achieved by a proper selection of a set of spatially dependent functions characterizing the rod material and geometric properties. Conditions for selecting such properties are established by constructing positive Lyapunov functions whose derivative with respect to the space variable is negative. It is shown that varying the shape of the rod alone is sufficient to confine the vibratory motion. In such a case, the vibration confinement requires that the eigenfunctions be exponentially decaying functions of space, where the notion of spatial domain stability is introduced as a concept dual to that of the time domain stability. It is also shown that vibration confinement can be produced if the rod density and/or stiffness are varied with respect to the space variable while the cross-section area is kept constant. Several case studies, supporting the developed conditions imposed on the spatially dependent functions for vibration confinement in vibrating rods, are discussed. Because variation in the geometric and material properties might decrease the critical buckling loads, we also discuss the buckling problem.

  16. Wetting of a partially immersed compliant rod

    Science.gov (United States)

    Hui, Chung-Yuen; Jagota, Anand

    2016-11-01

    The force on a solid rod partially immersed in a liquid is commonly used to determine the liquid-vapor surface tension by equating the measured force required to remove the rod from the liquid to the vertical component of the liquid-vapor surface tension. Here, we study how this process is affected when the rod is compliant. For equilibrium, we enforce force and configurational energy balance, including contributions from elastic energy. We show that, in general, the contact angle does not equal that given by Young's equation. If surface stresses are tensile, the strain in the immersed part of the rod is found to be compressive and to depend only on the solid-liquid surface stress. The strain in the dry part of the rod can be either tensile or compressive, depending on a combination of parameters that we identify. We also provide results for compliant plates partially immersed in a liquid under plane strain and plane stress. Our results can be used to extract solid surface stresses from such experiments.

  17. Single Rod Vibration in Axial Flow

    Science.gov (United States)

    Weichselbaum, Noah; Wang, Shengfu; Bardet, Philippe

    2013-11-01

    Fluid structure interaction of a single rod in axial flow is a coupled dynamical system present in many application including nuclear reactors, steam generators, and towed antenna arrays. Fluid-structure response can be quantified thanks to detailed experimental data where both structure and fluid responses are recorded. Such datum deepen understanding of the physics inherent to the system and provide high-dimensionality quantitative measurements to validate coupled structural and CFD codes with various level of complexity. In this work, single rods fixed on both ends in a concentric pipe, are subjected to an axial flow with Reynolds number based on hydraulic diameter of Re =4000. Rods of varying material stiffness and diameter are utilized in the experiment resulting in a range of dimensionless U between 0.5 and 1, where U = (ρA/EI)1/2uL. Experimental measurements of the velocity field around the rod are taken with PIV from time-resolved Nd:YLF laser and a high speed CMOS camera. Three-dimensional and temporal vibration and deflection of the rod is recorded with shadowgraphy utilizing two sets of pulsed high power LED and dedicated CMOS camera. Through integration of these two diagnostics, it is possible to reconstruct the full FSI domain providing unique validation data.

  18. Dielectric rod feed for compact range reflector

    CERN Document Server

    Balabukha, Nikolay P; Shapkina, Natalia E

    2014-01-01

    A dielectric rod feed with a special radiation pattern of a tabletop form used for the compact range reflector is developed and analyzed. Application of this feed increases the size of the compact range quiet zone generated by the reflector. The feed consists of the dielectric rod made of polystyren, the rod is inserted into the circular waveguide with a corrugated flange. The waveguide is excited by the H11-mode. The rod is covered by the textolite biconical bushing and has a fluoroplastic insert in the vicinity of the bushing. Mathematical modeling was used to obtain the parameters of the feed for the optimal tabletop form of the radiation pattern. The problem of the electromagnetic radiation was solved for metal-dielectric bodies of rotation by method of integral equations with further solving of the problem of the synthesis for feed parameters. The dielectric rod feed was fabricated for the X-frequency range. Feed amplitude and phase patterns were measured in the frequency range 8.2-12.5 GHz. Presented re...

  19. Rod consolidation at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1986-12-01

    A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab.

  20. Magnetic switch for reactor control rod. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-09-30

    A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

  1. HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Lustman, B.; Losco, E.F.; Cohen, I.

    1961-07-11

    Nuclear reactor control rods comprised of highly compressed and sintered finely divided metal alloy panticles and fine metal oxide panticles substantially uniformly distributed theretbrough are described. The metal alloy consists essentially of silver, indium, cadmium, tin, and aluminum, the amount of each being present in centain percentages by weight. The oxide particles are metal oxides of the metal alloy composition, the amount of oxygen being present in certain percentages by weight and all the oxygen present being substantially in the form of metal oxide. This control rod is characterized by its high strength and resistance to creep at elevated temperatures.

  2. Sensitivity study of control rod depletion coefficients

    OpenAIRE

    Blomberg, Joel

    2015-01-01

    This report investigates the sensitivity of the control rod depletion coefficients, Sg, to different input parameters and how this affects the accumulated 10B depletion, β. Currently the coefficients are generated with PHOENIX4, but the geometries can be more accurately simulated in McScram. McScram is used to calculate Control Rod Worth, which in turn is used to calculate Nuclear End Of Life, and Sg cannot be generated in the current version of McScram. Therefore, it is also analyzed whether...

  3. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors; Elaboration et qualification des schemas de calcul de reference pour les absorbants dans les reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Blanc-Tranchant, P

    1999-11-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B{sub 4}C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  4. The difference between 5 x 5 doubly nonnegative and completely positive matrices

    NARCIS (Netherlands)

    Burer, Samuel; Anstreicher, Kurt M.; Duer, Mirjam

    2009-01-01

    The convex cone of n x n completely positive (CP) matrices and its dual cone of copositive matrices arise in several areas of applied mathematics, including optimization. Every CP matrix is doubly nonnegative (DNN), i.e., positive semidefinite and component-wise nonnegative, and it is known that, fo

  5. The difference between 5 x 5 doubly nonnegative and completely positive matrices

    NARCIS (Netherlands)

    Burer, Samuel; Anstreicher, Kurt M.; Duer, Mirjam

    2009-01-01

    The convex cone of n x n completely positive (CP) matrices and its dual cone of copositive matrices arise in several areas of applied mathematics, including optimization. Every CP matrix is doubly nonnegative (DNN), i.e., positive semidefinite and component-wise nonnegative, and it is known that,

  6. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  7. Modeling and simulation performance of sucker rod beam pump

    Science.gov (United States)

    Aditsania, Annisa; Rahmawati, Silvy Dewi; Sukarno, Pudjo; Soewono, Edy

    2015-09-01

    Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.

  8. Modeling and simulation performance of sucker rod beam pump

    Energy Technology Data Exchange (ETDEWEB)

    Aditsania, Annisa, E-mail: annisaaditsania@gmail.com [Department of Computational Sciences, Institut Teknologi Bandung (Indonesia); Rahmawati, Silvy Dewi, E-mail: silvyarahmawati@gmail.com; Sukarno, Pudjo, E-mail: psukarno@gmail.com [Department of Petroleum Engineering, Institut Teknologi Bandung (Indonesia); Soewono, Edy, E-mail: esoewono@math.itb.ac.id [Department of Mathematics, Institut Teknologi Bandung (Indonesia)

    2015-09-30

    Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.

  9. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  10. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  11. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  12. Valve inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Meliksetian, A.; Sklencar, A.M.

    1982-12-01

    The overpressure transients for Westinghouse-designed NSSSs are reviewed to determine the fluid conditions at the inlet to the PORV and safety valves. The transients considered are: licensing (FSAR) transients; extended operation of high pressure safety injection system; and cold overpressurization. The results of this review, presented in the form of tables and graphs, define the range of fluid conditions expected at the inlet to pressurized safety and power-operated relief valves utilized in Westinghouse-designed PWR units. These results will provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units.

  13. Spent nuclear fuel rods encapsulated in copper

    Energy Technology Data Exchange (ETDEWEB)

    Hanes, H.D.

    1984-04-01

    Using hot isostatic pressing, spent nuclear fuel rods and other radioactive wastes can be encapsulated in solid copper. The copper capsule which is formed is free of pores and cracks, and is highly resistant to attack by reducing ground waters. Such capsules should contain radioactive materials safely for hundreds of thousands of years in underground storage.

  14. Solitary waves on nonlinear elastic rods. I

    DEFF Research Database (Denmark)

    Sørensen, Mads Peter; Christiansen, Peter Leth; Lomdahl, P. S.

    1984-01-01

    Acoustic waves on elastic rods with circular cross section are governed by improved Boussinesq equations when transverse motion and nonlinearity in the elastic medium are taken into account. Solitary wave solutions to these equations have been found. The present paper treats the interaction between...

  15. ELECTRIC FIELD MEASUREMENT IN ROD-DISCONTINUED ...

    African Journals Online (AJOL)

    2014-06-30

    Jun 30, 2014 ... The used arrangement with homogeneous system is made up of a square metallic sheet ... This distance is considered positive when the rod is located ... in the case of the discontinuous earth which were defined according to ...

  16. Fabrication of preliminary fuel rods for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  17. Brownian rod scheme in microenvironment sensing

    Directory of Open Access Journals (Sweden)

    Ian Gralinski

    2012-03-01

    Full Text Available Fluctuations of freely translating spherical particles via Brownian motion should provide inexhaustible information about the micro-environment, but is beset by the problem of particles drifting away from the venue of measurement as well as colliding with other particles. We propose a scheme here to circumvent this in which a Brownian rod that lies in proximity to a cylindrical pillar is drawn in by a tuneable attractive force from the pillar. The force is assumed to act through the centre of each body and the motion exclusive to the x-y plane. Simulation studies show two distinct states, one in which the rod is moving freely (state I and the other in which the rod contacts the cylinder surface (state II. Information about the micro-environment could be obtained by tracking the rotational diffusion coefficient Dθ populating in either of these two states. However, the magnitude of the normalized charge product in excess of 6.3x104 was found necessary for a rod of 6.81 × 0.93 μm2 (length × diameter and 10μm diameter cylindrical pillar to minimize deviation errors. It was also found that the extent of spatial sensing coverage could be controlled by varying the charge level. The conditions needed to ascertain the rotational sampling for angle determination through the Hough transform were also discussed.

  18. Piston rod seal for a Stirling engine

    Science.gov (United States)

    Shapiro, Wilbur

    1984-01-01

    In a piston rod seal for a Stirling engine, a hydrostatic bearing and differential pressure regulating valve are utilized to provide for a low pressure differential across a rubbing seal between the hydrogen and oil so as to reduce wear on the seal.

  19. Adjustable solitary waves in electroactive rods

    Science.gov (United States)

    Wang, Y. Z.; Zhang, C. L.; Dai, H.-H.; Chen, W. Q.

    2015-10-01

    This paper presents an asymptotic analysis of solitary waves propagating in an incompressible isotropic electroactive circular rod subjected to a biasing longitudinal electric displacement. Several asymptotic expansions are introduced to simplify the rod governing equations. The boundary conditions on the lateral surface of the rod are satisfied from the asymptotic point of view. In the limit of finite-small amplitude and long wavelength, a set of ten simplified one-dimensional nonlinear governing equations is established. To validate our approach and the derivation, we compare the linear dispersion relation with the one directly derived from the three-dimensional linear theory in the limit of long wavelength. Then, by the reductive perturbation method, we deduce the far-field equation (i.e. the KdV equation). Finally, the leading order of the electroelastic solitary wave solution is presented. Numerical examples are provided to show the influences of the biasing electric displacement and material constants on the solitary waves. It is found that the biasing electric displacement can modulate the velocity of solitary waves with a prescribed amplitude in the electroactive rod, a very interesting result which may promote the particular application of solitary waves in solids with multi-field coupling.

  20. On contact numbers in random rod packings

    NARCIS (Netherlands)

    Wouterse, A.; Luding, Stefan; Philipse, A.P.

    2009-01-01

    Random packings of non-spherical granular particles are simulated by combining mechanical contraction and molecular dynamics, to determine contact numbers as a function of density. Particle shapes are varied from spheres to thin rods. The observed contact numbers (and packing densities) agree well

  1. VERIFIKASI KECELAKAAN HILANGNYA ALIRAN AIR UMPAN PADA REAKTOR DAYA PWR MAJU

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

    Full Text Available AP1000 adalah reaktor daya PWR maju dengan daya listrik 1154 MW yang didesain berdasarkan kinerja teruji dari desain PWR lain oleh Westinghouse. Untuk mempersiapkan peran Pusat Teknologi Reaktor dan Keselamatan Nuklir sebagai suatu Technical Support Organization (TSO dalam hal verifikasi keselamatan, telah dilakukan kegiatan verifikasi keselamatan untuk AP1000 yang dimulai dengan verifikasi kecelakaan kegagalan pendingin sekunder. Kegiatan dimulai dengan pemodelan fitur keselamatan teknis yaitu sistem pendinginan teras pasif yang terdiri dari sistem Passive Residual Heat Removal (PRHR, tangki core makeup tank (CMT, dan tangki In-containment Refueling Water Storage Tank (IRWST. Kecelakaan kegagalan pendingin sekunder yang dipilih adalah hilangnya aliran air umpan ke salah satu pembangkit uap yang disimulasikan menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Tujuan analisis adalah untuk memperoleh sekuensi perubahan parameter termohidraulika reaktor akibat kecelakaan dimana hasil analisis yang diperoleh divalidasi dan dibandingkan dengan hasil analisis menggunakan program perhitungan LOFTRAN di dalam dokumen desain keselamatan AP1000. Hasil verifikasi menunjukkan bahwa kejadian hilangnya suplai air umpan tidak berdampak pada kerusakan teras, sistem pendingin reaktor, maupun sistem sekunder. Penukar kalor PRHR telah terverifikasi kemampuannya dalam membuang kalor peluruhan teras setelah trip reaktor. Hasil validasi dengan dokumen pembanding menunjukkan kesesuaian pada sebagian besar parameter termohidraulika. Secara umum, model PWR maju yang dilengkapi dengan sistem pendinginan teras ciri pasif yang telah dikembangkan tetap selamat ketika terjadi kecelakaan kehilangan aliran pendingin sekunder. Kata kunci: Verifikasi, hilangnya aliran air umpan, AP1000   AP1000 is a PWR power reactor with 1154 MW of electrical power that is designed based on the proven performance of the other Westinghouse PWR designs. To prepare the role of Center for

  2. Stimulus-evoked outer segment changes in rod photoreceptors

    Science.gov (United States)

    Zhao, Xiaohui; Thapa, Damber; Wang, Benquan; Lu, Yiming; Gai, Shaoyan; Yao, Xincheng

    2016-06-01

    Rod-dominated transient retinal phototropism (TRP) has been recently observed in freshly isolated mouse and frog retinas. Comparative confocal microscopy and optical coherence tomography revealed that the TRP was predominantly elicited from the rod outer segment (OS). However, the biophysical mechanism of rod OS dynamics is still unknown. Mouse and frog retinal slices, which displayed a cross-section of retinal photoreceptors and other functional layers, were used to test the effect of light stimulation on rod OSs. Time-lapse microscopy revealed stimulus-evoked conformational changes of rod OSs. In the center of the stimulated region, the length of the rod OS shrunk, while in the peripheral region, the rod OS swung toward the center region. Our experimental observation and theoretical analysis suggest that the TRP may reflect unbalanced rod disc-shape changes due to localized visible light stimulation.

  3. Validation Test of CARR Safety Rod Driving Mechanism

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>CARR safety Rods are driven by hydraulic force. The safety rod driving mechanism is designed by Tsinghua University and manufactured by Shenyang LIMING factory. Two sets of the mechanism are used for the validation test.

  4. Longitudinal Vibrations of Rheological Rod With Variable Cross Section

    Institute of Scientific and Technical Information of China (English)

    Katica(Stevanovic)HEDRIH; AleksandarFILIPOVSKI

    1999-01-01

    Longitudinal vibrations of rheological rod with variable cross section are examined.Particular solutions and eigenfunction are accomplished for natural vibrations of the rod with hereditary material of standard hereditary body.Some examples are given.

  5. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Directory of Open Access Journals (Sweden)

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-01-01

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

  7. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  8. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  9. Computer simulation of rod-sphere mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Antypov, Dmytro

    2003-07-01

    Results are presented from a series of simulations undertaken to investigate the effect of adding small spherical particles to a fluid of rods which would otherwise represent a liquid crystalline (LC) substance. Firstly, a bulk mixture of Hard Gaussian Overlap particles with an aspect ratio of 3:1 and hard spheres with diameters equal to the breadth of the rods is simulated at various sphere concentrations. Both mixing-demixing and isotropic-nematic transition are studied using Monte Carlo techniques. Secondly, the effect of adding Lennard-Jones particles to an LC system modelled using the well established Gay-Berne potential is investigated. These rod-sphere mixtures are simulated using both the original set of interaction parameters and a modified version of the rod-sphere potential proposed in this work. The subject of interest is the internal structure of the binary mixture and its dependence on density, temperature, concentration and various parameters characterising the intermolecular interactions. Both the mixing-demixing behaviour and the transitions between the isotropic and any LC phases have been studied for four systems which differ in the interaction potential between unlike particles. A range of contrasting microphase separated structures including bicontinuous, cubic, and micelle-like arrangement have been observed in bulk. Thirdly, the four types of mixtures previously studied in bulk are subjected to a static magnetic field. A variety of novel phases are observed for the cases of positive and negative anisotropy in the magnetic susceptibility. These include a lamellar structure, in which layers of rods are separated by layers of spheres, and a configuration with a self-assembling hexagonal array of spheres. Finally, two new models are presented to study liquid crystal mixtures in the presence of curved substrates. These are implemented for the cases of convex and concave spherical surfaces. The simulation results obtained in these geometries

  10. Control rod reactivity measurement by rod-drop method at a fast critical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, L.; Yin, Y.; Lian, X.; Zheng, C. [Inst. of Nuclear Physics and Chemistry in CAEP, P. O. Box 919 210, Mianyang, Sichuan, 621900 (China)

    2012-07-01

    Rod-drop experiments were carried out to estimate the reactivity of the control rod of a fast critical assembly operated by CAEP. Two power monitor systems were used to obtain the power level and integration method was used to process the data. Three experiments were performed. The experimental results of the reactivity from the two power monitor systems were consistent and showed a reasonable range of reactivity compared to results from positive period method. (authors)

  11. Research on operation safety analyzing method of marine PWR%船用压水堆运行安全分析方法

    Institute of Scientific and Technical Information of China (English)

    陈玉清; 蔡琦; 赵新文

    2011-01-01

    通过对船用压水堆设计安全限值和运行限值的保守性分析,给出开展运行安全研究的理论依据,提出以概率论和确定论相结合的联合模拟分析方法进行船用压水堆运行安全研究,并以一束控制棒失控抽出事故为例进行了实例分析.结果表明,所提出运行安全分析方法可以准确描述船用压水堆事故后的动态响应图景,开展运行安全研究可以为船用压水堆事故时的应急处置提供依据.%The theoretical foundation on carrying out the marine PWR operating safety analysis is given in this paper through analysis on the conservativeness of designed safety limits and operating limits.The analyzing method by combining the determinate and probabilistic risk assessment is put forward. The accident of one bundle control rod uncontrolled draw is adopted as an example which indicates that the dynamic process after accident can be correctly described by using the analyzing method given in this paper.Therefore through operating safety analysis, theoretical foundation can be found for the emergency disposition.

  12. Evolution of reactor monitoring and protection systems for PWR; Evolution des systemes de surveillance et de protection des REP

    Energy Technology Data Exchange (ETDEWEB)

    Chaloin, B. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Mourlevat, J.L. [FRAMATOME ANP, 92 - Paris-La-Defence (France)

    2004-07-01

    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  13. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement

    1999-04-01

    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  14. Application of LBB to high energy piping systems in operating PWR

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    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.

  15. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  16. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT

    Directory of Open Access Journals (Sweden)

    Uwe Imke

    2012-01-01

    Full Text Available SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT. The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  17. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramsthaler, J. A.; Lime, J. F.; Sahota, M. S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A.

  18. The continued development of the MFM suite and its practical application on a PWR system

    DEFF Research Database (Denmark)

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies...... in physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  19. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Clayton, Daniel James [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

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

  1. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  2. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  3. Sucker rod string design of the pumping systems

    Directory of Open Access Journals (Sweden)

    Chun Hua Liu

    2015-08-01

    Full Text Available The existing design of sucker rod string mainly focuses on the simplifying assumptions that rod string was exposed to simple tension loading. And its goal was to have equal modified stress at the top of each taper. The improved rod design was to have the same degree of safety at each section, and it used a dynamic force distribution that was proportional along the whole string. However, the available procedures did not provide the desired accuracy of its pertinent analysis, and the operators could not identify the specific phenomena that occur in CBM wells. In this paper, the mathematical models of rod loads and string length were developed based on the cyclic nature of rod string loading; the fatigue endurance method is used to design the single rod string; and the tapered rod string is designed to have an equal equivalent stress at the top of each section. Its application characteristics are demonstrated by the example of CBM wells in Ordos Basin. The interpretations of results show that the previous design gave the single rods a larger diameter and the top rods in the string a greater percent than the proposed method. The calculation should concern about inertial, vibration and friction forces to illustrate the elastic force waves travelling in the rod material with the speed of sound. The single string should be designed using fatigue endurance ratings due to asymmetric pulsating tension of rod loading; and the tapered string should involve a balanced design by setting the fatigue endurance at each section equal. A shorter stroke length gives a greater rod taper percentage and an increased load capacity results to an enhanced rod diameter. The rod diameter increases with the pump size and load capacity for the single string, and the rod taper percentage of the top rod strings increases with plunger diameter for the tapered string. The proposed research improves efficiency of the pumping system, assures good operating conditions, and reduces

  4. Investigation of control rod worth and nuclear end of life of BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of {sup 10}B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% {sup 10}B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in {sup 10}B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming.

  5. Rod Has High Tensile Strength And Low Thermal Expansion

    Science.gov (United States)

    Smith, D. E.; Everton, R. L.; Howe, E.; O'Malley, M.

    1996-01-01

    Thoriated tungsten extension rod fabricated to replace stainless-steel extension rod attached to linear variable-differential transformer in gap-measuring gauge. Threads formed on end of rod by machining with special fixtures and carefully chosen combination of speeds and feeds.

  6. 77 FR 1504 - Stainless Steel Wire Rod From India

    Science.gov (United States)

    2012-01-10

    ... COMMISSION Stainless Steel Wire Rod From India Determination On the basis of the record \\1\\ developed in the... antidumping duty order on stainless steel wire rod From India would be likely to lead to continuation or... contained in USITC Publication 4300 (January 2012), entitled Stainless Steel Wire Rod From...

  7. Carbon Inverse Opal Rods for Nonenzymatic Cholesterol Detection.

    Science.gov (United States)

    Zhong, Qifeng; Xie, Zhuoying; Ding, Haibo; Zhu, Cun; Yang, Zixue; Gu, Zhongze

    2015-11-18

    Carbon inverse opal rods made from silica photonic crystal rods are used for nonenzymatic cholesterol sensing. The characteristic reflection peak originating from the physical periodic structure works as sensing signals for quantitatively estimating cholesterol concentrations. Carbon inverse opal rods work both in cholesterol standard solutions and human serum. They are suitable for practical use in clinical diagnose.

  8. Power-Cooling-Mismatch Test Series Test PCM-7. Experiment operating specifications. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sparks, D.T.; Smith, R.H.; Stanley, C.J.

    1979-02-01

    The experiment operating specifications for the Power-Cooling-Mismatch (PCM) Test PCM-7 to be conducted in the Power Burst Facility are described. The PCM Test Series was designed on the basis of a parametric evaluation of fuel behavior response with cladding temperature, rod internal pressure, time in film boiling, and test rod power being the variable parameters. The test matrix, defined in the PCM Experiment Requirements Document (ERD), encompasses a wide range of situations extending from pre-CHF (critical heat flux) PCMs to long duration operation in stable film boiling leading to rod failure.

  9. CARR辐照压水堆小组件热工水力分析%Thermal-hydraulic Analysis of PWR Small Assembly for Irradiation Test of CARR

    Institute of Scientific and Technical Information of China (English)

    尹皓; 邹耀; 刘兴民

    2015-01-01

    T he thermal‐hydraulic behaviors of the PWR 4 × 4 small assembly tested in the high temperature and high pressure loop of China Advanced Research Reactor were analyzed .The CFD method was used to carry out 3D simulation of the model ,thus detailed thermal‐hydraulic parameters were obtained .Firstly ,the simplified model was simulated to give the 3D temperature and velocity distributions and analyze the heat transfer process .Then the whole scale small assembly model was simulated and the simulation results were compared with those of simplified rod bundle model .Its flow behavior was studied and flow mixing characteristics of the grids were analyzed ,and the mixing factor of the grid was calculated and can be used for further thermal‐hydraulic study .It is show n that the highest temperature of the fuel rod meets the design limit and the mixing effect of the grid is obvious .%分析压水堆4×4小组件在CARR高温高压回路中进行辐照考验时的热工水力问题。利用计算流体动力学(C FD )软件对其进行三维数值模拟,以获得详细的热工水力参数。首先,模拟简化的燃料棒束模型,得出三维温度与速度分布,并分析了传热过程。然后,模拟全尺寸小组件,与棒束模型所得的结果进行对比分析,着重研究其流动,并分析了格架的搅混特性,得出可应用于一维热工水力程序的搅混因子。结果表明,燃料棒最高温度可满足安全性要求,且格架的搅混作用明显。

  10. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  11. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Energy Technology Data Exchange (ETDEWEB)

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.

    2014-07-01

    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  12. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    Institute of Scientific and Technical Information of China (English)

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  13. Accelerated IGA/SCC testing of Alloy 600 in contaminated PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Miglin, B.P.; Sarver, J.M. [Babcock & Wilcox R& D Division, Alliance, OH (United States); Aoki, K. [NFI, Osaka (Japan); Koch, D.W. [Babcock & Wilcox Nuclear Services, Lynchburg, VA (United States); Takamatsu, H. [Kansai Electric, Osaka (Japan)

    1992-12-31

    An accelerated corrosion test (360{degrees}C for 2000 hrs) was performed on C-ring specimens machined from one heat of Alloy 600 tubing in the mill-annealed condition. The specimens were exposed to secondary-side pressurized-water-reactor (PWR) solutions contaminated with lead, sulfur, silicon, and a combination of these contaminants. Where possible, MULTEQ calculations were performed to determine the chemical concentrations so that a constant elevated-temperature pH of 4.5 was achieved. This test was designed to examine the ability of these contaminants to cause intergranular attack and/or stress corrosion in stressed Alloy 600 tubing. The results from this test demonstrated that under the test conditions used, lead-contaminated PWR secondary water induces and propagates intergranular attack (IGA) and stress corrosion cracking (SCC) in Alloy 600. Attack was intergranular; the degree of attack did not vary in the liquid or vapor portions of the test environments. Although attack was more severe at higher stresses, significant attack was observed in samples stressed to the typical operating stress. Solutions of only sulfur and only silicon displayed no initiation or propagation of either IGA or SCC. However, the solution containing all three contaminants caused attack with identical morphology to that observed in the lead-contaminated solution.

  14. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  15. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

    Directory of Open Access Journals (Sweden)

    Mario Matijević

    2016-01-01

    Full Text Available The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS were demonstrated when applied to a realistic deep penetration Monte Carlo (MC shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR parameters is based on deterministic transport theory (SN method providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between SN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

  16. Characterization of Oxide Layer with Precipitates of HANA-6 Exposed in Simulated PWR Primary Water Environment

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hun; Lim, Jea Young; Lee, Sung Yong; Kim, Yoon Ho; Mok, Yong Kyoon [KEPCO NF, Daejeon (Korea, Republic of)

    2016-10-15

    The delayed oxidation behaviors of β-Nb ppts and their amorphization behaviors in HANA-6 and other Zr-base alloys have been frequently reported. On the other hand, although Zr(Nb,Fe)2 ppts could be formed in the HANA-6 alloy due to Fe impurities contained in Zrsponge, the oxidation behavior of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy has not been fully understood. In this study, oxide characteristics of HANA-6 corroded in simulated PWR environment for 165 and 315 days were investigated. And, oxidation behaviors of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy were investigated by TEM with EDS techniques. The superior corrosion property of HANA-6 has been confirmed through corrosion test in simulated PWR water for 387 days. By using TEM/EDS technique, the oxide characteristics with presence of β- Nb (or β-enriched), and ZrNbFe (possibly Zr(Nb,Fe){sub 2}) ppts have been characterized as follows. 1. Delayed oxidation behaviors of β-Nb and Zr(Nb,Fe){sub 2} ppts and their amorphization due to oxidation were observed from TEM/EDS analyses. 2. The oxide layers having crystallite and partially amorphous ppts were slightly increased with increasing corrosion test time from 165 days to 315 days. 3. In outer oxide layer, Fe in Zr(Nb,Fe){sub 2} ppt was depleted and dissolved to outer layer of ppt and bulk oxide layer.

  17. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  18. Vulnerability of a partially flooded PWR reactor cavity to a steam explosion

    Energy Technology Data Exchange (ETDEWEB)

    Cizelj, Leon [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)]. E-mail: leon.cizelj@ijs.si; Koncar, Bostjan [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia); Leskovar, Matjaz [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)

    2006-08-15

    When the hot molten core comes into contact with the water in the reactor cavity a steam explosion may occur. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and later, during the expansion of the water vapour, to production of missiles that may endanger surrounding structures. The purpose of the performed analysis is to provide an estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion, and to make an assessment of the vulnerabilities of the typical PWR cavity structures to steam explosions. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by comprehensive and reasonably conservative analyses of two typical ex-vessel steam explosion cases differing in the steam explosion energy conversion ratio. In particular, the vulnerability of the surrounding reinforced concrete walls to damage has been sought in both cases.

  19. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  20. Construction and utilization of linear empirical core models for PWR in-core fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k{sub {infinity}} profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results.

  1. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  3. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  4. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  5. Surface Oxidation Phenomena of Ni-Based Alloy 600 in PWR Primary Water Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Yun Soo; Hwang, Seong Sik; Kim, Sung Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    There is, nevertheless, growing evidence in support for the internal oxidation model by Scot, in which grain boundary oxidation is responsible for embrittlement and cracking. Grain boundaries can act as an enhanced diffusion path for oxidation, and grain boundary oxidation can be regarded as a precursor for crack initiation. Oxidation of the grain boundary in almost all nickel-based alloys exposed to primary water is known to be detrimental for grin boundary cohesion. Panter et al. showed that the crack initiation time is strongly reduced when the specimens are pre-exposed in a simulated PWR environment in the absence of applied stress. The changes of the grain boundary structure and chemistry owing to oxygen penetration can increase the sensitivity to PWSCC under a load since grain boundary oxidization significantly weakens the grain boundary strength. Most of the important experimental results obtained are believed to correlate with the oxidation penetration into the material. A spinel structure was detected by XRD in the oxide layers. Several different types of oxide scales were found by SEM examination on the corroded surface of Alloy 600 after an immersion test in the primary water environments. Surface grain boundaries were oxidized by oxygen penetration into the matrix through grain boundaries. Grain boundary oxidization is thought to be the main reason for intergranular cracking in this alloy in a primary water environment of a PWR.

  6. STUDY ON A HYDROPHOBIC-HYDROPHILIC GRADIENT ROD

    Institute of Scientific and Technical Information of China (English)

    Jun Ma; Bai-yu Li; Hai-yun Liu; Zhi-min Zheng; Jian Xu

    2004-01-01

    A hydrophobic-hydrophilic gradient rod with a length of 40 mm and a diameter of 3 mm was prepared by heating a polymethylsilsesquioxane rod in a cylindrical stove with temperature gradient. The rod was thus pyrolyzed under a temperature gradient condition. The organic end of the gradient rod appears hydrophobic with a contact angle of 109.9° while the other end is hydrophilic with a contact angle of 62.4°. The gradient chemical structure and the gradient microstructure along the rod were characterized by FTIR and SEM, respectively.

  7. Test Research on Special Sucker Rod for Screw Pump

    Institute of Scientific and Technical Information of China (English)

    Zhang Mingyi; Chen Mingzhan; Li Zhi

    2006-01-01

    @@ According to the statistics of straight thread sucker rods' application in screw pump in Daqing Oilfield before2000, the proportion of sucker rods' yearly breakaway reached to 41.6%, taking up 70% of the total wells that were checked. Thus it can be seen that the rods breakaway problem was becoming the main barrier restricting screw pump large-scale population and application. Since then,the development work on the special sucker rods for screw pump had been carried on. Through the analysis on the failure position and failure form of the sucker rods',the following conclusions arepresented:

  8. Photonic mesophases from cut rod rotators

    Energy Technology Data Exchange (ETDEWEB)

    Stelson, Angela C.; Liddell Watson, Chekesha M., E-mail: cml66@cornell.edu [Materials Science and Engineering, Cornell University, Ithaca, New York 14853 (United States); Avendano, Carlos [Chemical Engineering and Analytical Science, The University of Manchester, Manchester M13 9PL (United Kingdom)

    2016-01-14

    The photonic band properties of random rotator mesophases are calculated using supercell methods applied to cut rods on a hexagonal lattice. Inspired by the thermodynamic mesophase for anisotropic building blocks, we vary the shape factor of cut fraction for the randomly oriented basis. We find large, stable bandgaps with high gap isotropy in the inverted and direct structures as a function of cut fraction, dielectric contrast, and filling fraction. Bandgap sizes up to 34.5% are maximized at high dielectric contrast for rods separated in a matrix. The bandgaps open at dielectric contrasts as low as 2.0 for the transverse magnetic polarization and 2.25 for the transverse electric polarization. Additionally, the type of scattering that promotes the bandgap is correlated with the effect of disorder on bandgap size. Slow light properties are investigated in waveguide geometry and slowdown factors up to 5 × 10{sup 4} are found.

  9. Oligo(naphthylene–ethynylene) Molecular Rods

    DEFF Research Database (Denmark)

    Cramer, Jacob Roland; Ning, Yanxiao; Shen, Cai;

    2013-01-01

    Molecular rods designed for surface chirality studies have been synthesized in high yields. The molecules are composed of oligo(naphthylene–ethynylene) skeletons and functionalized at their two termini with carboxylic acids and hydrophobic groups. The molecular skeletons were constructed by means...... of palladium-catalyzed Sonogashira reactions between naphthyl halides and acetylenes. The triazene functionality was used as a protected iodine precursor to allow linear extension of the molecular rods during the synthe-ses. The carboxylic acid groups in the target molecules were protected as esters during...... the synthesis to keep the large aromatic molecules soluble during their syntheses. These rigid oligomers were designed to form lamella-like structures when adsorbed on a surface, through which multiple distinguishable surface conformations should be obtainable. Preliminary scanning tunneling microscopy imaging...

  10. [Rod of Asclepius. Symbol of medicine].

    Science.gov (United States)

    Young, Pablo; Finn, Bárbara C; Bruetman, Julio E; Cesaro Gelos, Jorge; Trimarchi, Hernán

    2013-09-01

    Symbolism is one of the most archaic forms of human thoughts. Symbol derives from the Latin word symbolum, and the latter from the Greek symbolon or symballo, which means "I coincide, I make matches". The Medicine symbol represents a whole series of historical and ethical values. Asclepius Rod with one serpent entwined, has traditionally been the symbol of scientific medicine. In a misconception that has lasted 500 years, the Caduceus of Hermes, entwined by two serpents and with two wings, has been considered the symbol of Medicine. However, the Caduceus is the current symbol of Commerce. Asclepius Rod and the Caduceus of Hermes represent two professions, Medicine and Commerce that, in ethical practice, should not be mixed. Physicians should be aware of their real emblem, its historical origin and meaning.

  11. Composites reinforcement by rods: a SAS study

    Energy Technology Data Exchange (ETDEWEB)

    Urban, V. [ESRF, BP220, 38043 Grenoble Cedex (France); Botti, A.; Pyckhout-Hintzen, W.; Richter, D. [IFF-Forschungszentrum Juelich, 52425 Juelich (Germany); Straube, E. [University of Halle, FB Physik, 06099 Halle (Germany)

    2002-07-01

    The mechanical properties of composites are governed by size, shape and dispersion degree of so-called reinforcing particles. Polymeric fillers based on thermodynamically driven microphase separation of block copolymers offer the opportunity to study a model system of controlled rod-like filler particles. We chose a triblock copolymer (PBPSPB) and carried out SAS measurements with both X-rays and neutrons, in order to characterize separately the hard phase and the cross-linked PB matrix. The properties of the material depend strongly on the way that stress is carried and transferred between the soft matrix and the hard fibers. The failure of the strain-amplification concept and the change of topological contributions to the free energy and scattering factor have to be addressed. In this respect the composite shows a similarity to a two-network system, i.e. interpenetrating rubber and rod-like filler networks. (orig.)

  12. Composites reinforcement by rods a SAS study

    CERN Document Server

    Urban, V; Pyckhout-Hintzen, W; Richter, D; Straube, E

    2002-01-01

    The mechanical properties of composites are governed by size, shape and dispersion degree of so-called reinforcing particles. Polymeric fillers based on thermodynamically driven microphase separation of block copolymers offer the opportunity to study a model system of controlled rod-like filler particles. We chose a triblock copolymer (PBPSPB) and carried out SAS measurements with both X-rays and neutrons, in order to characterize separately the hard phase and the cross-linked PB matrix. The properties of the material depend strongly on the way that stress is carried and transferred between the soft matrix and the hard fibers. The failure of the strain-amplification concept and the change of topological contributions to the free energy and scattering factor have to be addressed. In this respect the composite shows a similarity to a two-network system, i.e. interpenetrating rubber and rod-like filler networks. (orig.)

  13. Instabilities of a rotating helical rod

    Science.gov (United States)

    Park, Yunyoung; Ko, William; Kim, Yongsam; Lim, Sookkyung

    2016-11-01

    Bacteria such as Escherichia coli and Vibrio alginolyticus have helical flagellar filament. By rotating a motor, which is located at the bottom end of the flagellar filament embedded in the cell body, CCW or CW, they swim forward or backward. We model a left-handed helix by the Kirchhoff rod theory and use regularized Stokes formulation to study an interaction between the surrounding fluid and the flagellar filament. We perform numerical studies focusing on relations between physical parameters and critical angular frequency of the motor, which separates overwhiring from twirling. We are also interested in the buckling instability of the hook, which is very flexible elastic rod. By measuring buckling angle, which is an angle between rotational axis and helical axis, we observe the effects of physical parameters on buckling of the hook.

  14. Hollow Sucker Rod Applied in Production Engineering

    Institute of Scientific and Technical Information of China (English)

    Wang Tongbin; Liu Liandong; Hu Daoming; Jia Yanshan

    1997-01-01

    @@ Working Principle A positive cycle system or a working channel can be formed by means of hollow sucker rod and its mating parts in the oil tube ofa well, through which heat carriers (such as hot water,hot oil and steam), chemicals and heating cable can be pumped or put into the well so as to lower the viscosity of crude, dissolve the paraffin building-up and open the conduit, thus leading to the smooth oil flow out of well.

  15. Rod Driven Frequency Entrainment and Resonance Phenomena

    Directory of Open Access Journals (Sweden)

    Christina Salchow

    2016-08-01

    Full Text Available A controversy exists on photic driving in the human visual cortex evoked by intermittent photic stimulation. Frequency entrainment and resonance phenomena are reported for frequencies higher than 12 Hz in some studies while missing in others. We hypothesized that this might be due to different experimental conditions, since both high and low intensity light stimulation were used. However, most studies do not report radiometric measurements, which makes it impossible to categorize the stimulation according to photopic, mesopic, and scotopic vision. Low intensity light stimulation might lead to scotopic vision, where rod perception dominates. In this study, we investigated photic driving for rod-dominated visual input under scotopic conditions. Twelve healthy volunteers were stimulated with low intensity light flashes at 20 stimulation frequencies, leading to rod activation only. The frequencies were multiples of the individual alpha frequency (α of each volunteer in the range from 0.40–2.30*α. 306-channel whole head magnetoencephalography recordings were analyzed in time, frequency, and spatiotemporal domains with the Topographic Matching Pursuit algorithm. We found resonance phenomena and frequency entrainment for stimulations at or close to the individual alpha frequency (0.90–1.10*α and half of the alpha frequency (0.40–0.55*α. No signs of resonance and frequency entrainment phenomena were revealed around 2.00*α. Instead, on-responses at the beginning and off-responses at the end of each stimulation train were observed for the first time in a photic driving experiment at frequencies of 1.30–2.30*α, indicating that the flicker fusion threshold was reached. All results, the resonance and entrainment as well as the fusion effects, provide evidence for rod-dominated photic driving in the visual cortex.

  16. Rod Driven Frequency Entrainment and Resonance Phenomena

    Science.gov (United States)

    Salchow, Christina; Strohmeier, Daniel; Klee, Sascha; Jannek, Dunja; Schiecke, Karin; Witte, Herbert; Nehorai, Arye; Haueisen, Jens

    2016-01-01

    A controversy exists on photic driving in the human visual cortex evoked by intermittent photic stimulation. Frequency entrainment and resonance phenomena are reported for frequencies higher than 12 Hz in some studies while missing in others. We hypothesized that this might be due to different experimental conditions, since both high and low intensity light stimulation were used. However, most studies do not report radiometric measurements, which makes it impossible to categorize the stimulation according to photopic, mesopic, and scotopic vision. Low intensity light stimulation might lead to scotopic vision, where rod perception dominates. In this study, we investigated photic driving for rod-dominated visual input under scotopic conditions. Twelve healthy volunteers were stimulated with low intensity light flashes at 20 stimulation frequencies, leading to rod activation only. The frequencies were multiples of the individual alpha frequency (α) of each volunteer in the range from 0.40 to 2.30∗α. Three hundred and six-channel whole head magnetoencephalography recordings were analyzed in time, frequency, and spatiotemporal domains with the Topographic Matching Pursuit algorithm. We found resonance phenomena and frequency entrainment for stimulations at or close to the individual alpha frequency (0.90–1.10∗α) and half of the alpha frequency (0.40–0.55∗α). No signs of resonance and frequency entrainment phenomena were revealed around 2.00∗α. Instead, on-responses at the beginning and off-responses at the end of each stimulation train were observed for the first time in a photic driving experiment at frequencies of 1.30–2.30∗α, indicating that the flicker fusion threshold was reached. All results, the resonance and entrainment as well as the fusion effects, provide evidence for rod-dominated photic driving in the visual cortex. PMID:27588002

  17. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Science.gov (United States)

    Young, J.N.

    1957-08-20

    An electromagnetic device for moving an object in a linear path by increments is described. The device is specifically adapted for moving a neutron absorbing control rod into and out of the core of a reactor and consists essentially of an extension member made of magnetic material connected to one end of the control rod and mechanically flexible to grip the walls of a sleeve member when flexed, a magnetic sleeve member coaxial with and slidable between limit stops along the flexible extension, electromagnetic coils substantially centrally located with respect to the flexible extension to flex the extension member into gripping engagement with the sleeve member when ener gized, moving electromagnets at each end of the sleeve to attract the sleeve when energized, and a second gripping electromagnet positioned along the flexible extension at a distance from the previously mentioned electromagnets for gripping the extension member when energized. In use, the second gripping electromagnet is deenergized, the first gripping electromagnet is energized to fix the extension member in the sleeve, and one of the moving electromagnets is energized to attract the sleeve member toward it, thereby moving the control rod.

  18. Description and characterization of HBWR Series H-1 test rods

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, S.R.; Barner, J.O.; Welty, R.K.

    1979-06-01

    The as-built characterization results are presented for the HBWR Series H-1 test rods to be irradiated as part of the Fuel Performance Improvement Program (FPIP). The irradiation of these rods is to be conducted in the Halden Boiling Water Reactor (HBWR). Series H-1 consists of twelve rods for irradiation and six spares. Rod design types include (1) a reference dished pellet design, (2) an annular pellet design, (3) an annular pellet design combined with graphite-coated cladding, and (4) a packed-particle (vipac) design. The report, which describes the fabrication and detailed characterization results for the rods, is divided into four major sections: (1) experiment description, (2) process development required to fabricate the test rods, (3) methods and procedures used to fabricate and characterize the rods, and (4) a summary of the characterization results.

  19. Nuclear thermionic converter. [tungsten-thorium oxide rods

    Science.gov (United States)

    Phillips, W. M.; Mondt, J. F. (Inventor)

    1977-01-01

    Efficient nuclear reactor thermionic converter units are described which can be constructed at low cost and assembled in a reactor which requires a minimum of fuel. Each converter unit utilizes an emitter rod with a fluted exterior, several fuel passages located in the bulges that are formed in the rod between the flutes, and a collector receiving passage formed through the center of the rod. An array of rods is closely packed in an interfitting arrangement, with the bulges of the rods received in the recesses formed between the bulges of other rods, thereby closely packing the nuclear fuel. The rods are constructed of a mixture of tungsten and thorium oxide to provide high power output, high efficiency, high strength, and good machinability.

  20. Research on General Corrosion Property of 304L and 304NG Stainless Steels in Simulated PWR Primary Water

    Institute of Scientific and Technical Information of China (English)

    PENG; De-quan; HU; Shi-lin; ZHANG; Ping-zhu; WANG; Hui

    2012-01-01

    <正>The general corrosion behaviors of 304L and 304NG grade stainless steels in simulated pressurized water reactor (PWR) primary loop were studied using still autoclave, respectively, the corrosion test lasted for 1 680 hours. The corrosion oxide films were analyzed macroscopically and microscopically. The results are shown in Figs. 1, 2.

  1. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in BWR and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Haenninen, H.; Aho-Mantila, I.

    1981-05-01

    The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, has been examined both in O2-enriched BWR-conditions (8 ppm O2) and in typical PWR-conditions.

  2. Rod and Rod-driven Function in Achromatopsia and Blue Cone Monochromatism

    Science.gov (United States)

    Moskowitz, Anne; Hansen, Ronald M.; Akula, James D.; Eklund, Susan E.; Fulton, Anne B.

    2008-01-01

    Purpose To evaluate rod photoreceptor and postreceptor retinal function in pediatric patients with achromatopsia (ACHR) and blue cone monochromatism (BCM) using contemporary electroretinographic (ERG) procedures. Methods Fifteen patients (age 1 to 20 years) with ACHR and six patients (age 4 to 22 years) with BCM were studied. ERG responses to full-field stimuli were obtained in scotopic and photopic conditions. Rod photoreceptor (Srod, Rrod) and rod-driven postreceptor (log σ, Vmax) response parameters were calculated from the a-wave and b-wave. The ERG records were digitally filtered to demonstrate the oscillatory potentials (OPs); a sensitivity parameter, log SOPA1/2, and an amplitude parameter, SOPAmax, were used to characterize the OP response. Response parameters were compared to those of 12 normal control subjects. Results As expected, photopic responses were non-detectable in patients with ACHR and BCM. In addition, mean scotopic photoreceptor (Rrod) and postreceptor (Vmax and SOPAmax) amplitude parameters were significantly reduced compared to those in normal controls. The flash intensity required to evoke a half maximum b-wave amplitude (log σ) was significantly increased. Conclusions The results of this study provide evidence that deficits in rod and rod mediated function occur in the primary cone dysfunction syndromes, achromatopsia and blue cone monochromatism. PMID:18824728

  3. Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series); Anomalie de temps de chute des grappes de controle centrale de guang dong (daya bay) et centrales d`electricite de France (Palier N4-1450 MWE)

    Energy Technology Data Exchange (ETDEWEB)

    Olivera, J.J.; Naury, S.; Tricot, N.; Tran Dai, P.; Gama, J.M.

    1996-12-31

    The anomaly of control rod cluster drop time revealed at Guandong Nuclear Power Station in Daya Bay and in the Chooz B1 pilot unit for the N4 series, led to the replacement of the M1 type control rod cluster guide tubes with 1300 MWe PWR type guide tubes, adapted to the geometry of the Guandong reactors and the 1450 MWe reactors of the N4 series. The comparison of the drop times obtained with the 1300 MWe type control rod cluster guide 1300 MWe type control rod cluster guide tubes gave satisfactory results. These met the safety criterion for N4 series control rod cluster drop times (2.15 under hot shutdown conditions). The drop time tests which will be carried out in middle of and at the end of cycle 1 of Chooz B1 should make it possible to finally validate the solution already successfully implemented at Guandong. However, this anomaly has revealed the limits of representativeness of the experimental test loops with regard to the real reactor configuration. In view of this, it has been deemed necessary to ask Electricite de France to pursue its analysis both on the understanding of the phenomena which led to this anomaly and on the limits of the representativeness of the experimental test loops. (authors).

  4. Between a Map and a Data Rod

    Science.gov (United States)

    Teng, W. L.; Rui, H.; Strub, R. F.; Vollmer, B.

    2015-12-01

    A "Digital Divide" has long stood between how NASA and other satellite-derived data are typically archived (time-step arrays or "maps") and how hydrology and other point-time series oriented communities prefer to access those data. In essence, the desired method of data access is orthogonal to the way the data are archived. Our approach to bridging the Divide is part of a larger NASA-supported "data rods" project to enhance access to and use of NASA and other data by the Consortium of Universities for the Advancement of Hydrologic Science, Inc. (CUAHSI) Hydrologic Information System (HIS) and the larger hydrology community. Our main objective was to determine a way to reorganize data that is optimal for these communities. Two related objectives were to optimally reorganize data in a way that (1) is operational and fits in and leverages the existing Goddard Earth Sciences Data and Information Services Center (GES DISC) operational environment and (2) addresses the scaling up of data sets available as time series from those archived at the GES DISC to potentially include those from other Earth Observing System Data and Information System (EOSDIS) data archives. Through several prototype efforts and lessons learned, we arrived at a non-database solution that satisfied our objectives/constraints. We describe, in this presentation, how we implemented the operational production of pre-generated data rods and, considering the tradeoffs between length of time series (or number of time steps), resources needed, and performance, how we implemented the operational production of on-the-fly ("virtual") data rods. For the virtual data rods, we leveraged a number of existing resources, including the NASA Giovanni Cache and NetCDF Operators (NCO) and used data cubes processed in parallel. Our current benchmark performance for virtual generation of data rods is about a year's worth of time series for hourly data (~9,000 time steps) in ~90 seconds. Our approach is a specific

  5. Between a Map and a Data Rod

    Science.gov (United States)

    Teng, William; Rui, Hualan; Strub, Richard; Vollmer, Bruce

    2015-01-01

    A Digital Divide has long stood between how NASA and other satellite-derived data are typically archived (time-step arrays or maps) and how hydrology and other point-time series oriented communities prefer to access those data. In essence, the desired method of data access is orthogonal to the way the data are archived. Our approach to bridging the Divide is part of a larger NASA-supported data rods project to enhance access to and use of NASA and other data by the Consortium of Universities for the Advancement of Hydrologic Science, Inc. (CUAHSI) Hydrologic Information System (HIS) and the larger hydrology community. Our main objective was to determine a way to reorganize data that is optimal for these communities. Two related objectives were to optimally reorganize data in a way that (1) is operational and fits in and leverages the existing Goddard Earth Sciences Data and Information Services Center (GES DISC) operational environment and (2) addresses the scaling up of data sets available as time series from those archived at the GES DISC to potentially include those from other Earth Observing System Data and Information System (EOSDIS) data archives. Through several prototype efforts and lessons learned, we arrived at a non-database solution that satisfied our objectivesconstraints. We describe, in this presentation, how we implemented the operational production of pre-generated data rods and, considering the tradeoffs between length of time series (or number of time steps), resources needed, and performance, how we implemented the operational production of on-the-fly (virtual) data rods. For the virtual data rods, we leveraged a number of existing resources, including the NASA Giovanni Cache and NetCDF Operators (NCO) and used data cubes processed in parallel. Our current benchmark performance for virtual generation of data rods is about a years worth of time series for hourly data (9,000 time steps) in 90 seconds. Our approach is a specific implementation of

  6. Determination of the rod-wire transition length in colloidal indium phosphide quantum rods.

    Science.gov (United States)

    Wang, Fudong; Buhro, William E

    2007-11-21

    Colloidal InP quantum rods (QRs) having controlled diameters and lengths are grown by the solution-liquid-solid method, from Bi nanoparticles in the presence of hexadecylamine and other conventional quantum dot surfactants. These quantum rods show band-edge photoluminescence after HF photochemical etching. Photoluminescence efficiency is further enhanced after the Bi tips are selectively removed from the QRs by oleic acid etching. The QRs are anisotropically 3D confined, the nature of which is compared to the corresponding isotropic 3D confinement in quantum dots and 2D confinement in quantum wires. The 3D-2D rod-wire transition length is experimentally determined to be 25 nm, which is about 2 times the bulk InP exciton Bohr radius (of approximately 11 nm).

  7. Establishing baseline rod electroretinogram values in achromatopsia and cone dystrophy.

    Science.gov (United States)

    Wang, Isaac; Khan, Naheed W; Branham, Kari; Wissinger, B; Kohl, Susanne; Heckenlively, J R

    2012-12-01

    To establish the normal range of values for rod-isolated b-wave amplitudes in achromatopsia and cone dystrophies. We reviewed charts of 112 patients with various types of cone dystrophy, and compared their standardized electroretinographic rod b-wave amplitudes with age-matched normal controls. Twenty-six patients had known mutations in achromatopsia and cone dystrophy genes, while 53 were characterized by their inheritance pattern since they had yet to have their gene identified. Visual acuity information and scotomata were documented. We found that patients with achromatopsia and cone dystrophy had rod b-wave amplitudes that were significantly lower than age-matched controls, but found no evidence of rod amplitude progression nor loss of peripheral visual fields in the study group. We found that cone dystrophy patients of all types had depressed rod-isolated ERGs across the board. If typical diagnostic criteria are used, these patients might be considered to have "abnormal" rod-isolated electroretinographic values, and might be called "cone-rod dystrophy", even though the waveforms are stable for years. Patients with cone-rod dysfunction patterns on ERG can be better understood by also performing kinetic (Goldmann) visual fields, which will help to distinguish cone dystrophies from progressive cone-rod dystrophies by central scotomata size and progression over time in many forms of cone-rod dystrophy.

  8. Sucker-rod pumping handbook production engineering fundamentals and long-stroke rod pumping

    CERN Document Server

    Takacs, Gabor

    2015-01-01

    Sucker-Rod Pumping Handbook presents the latest information on the most common form of production enhancement in today's oil industry, making up roughly two-thirds of the producing oilwell operations in the world. The book begins with an introduction to the main features of sucker rod pumping and an explanation and comparison of lift methods. It goes on to provide the technical and practical knowledge needed to introduce the new and practicing production engineer and operator to the equipment, technology, and applications required to maintain optimum operating conditions.

  9. Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands` PINK programme (Joint project of ECN, IRI, KEMA and GKN)

    Energy Technology Data Exchange (ETDEWEB)

    Gruppelaar, H. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Klippel, H.T. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Verhagen, F.C.M. [Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands); Bruggink, J.C. [Gemeenschappelijke Kernenergiecentrale Nederland N.V., Dodewaard (Netherlands)

    1993-11-01

    Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

  10. Visual transduction in human rod photoreceptors.

    Science.gov (United States)

    Kraft, T W; Schneeweis, D M; Schnapf, J L

    1993-05-01

    1. Photocurrents were recorded with suction electrodes from rod photoreceptors of seven humans. 2. Brief flashes of light evoked transient outward currents of up to 20 pA. With increasing light intensity the peak response amplitude increased along an exponential saturation function. A half-saturating peak response was evoked by approximately sixty-five photoisomerizations. 3. Responses to brief dim flashes rose to a peak in about 200 ms. The waveform was roughly like the impulse response of a series of four to five low-pass filters. 4. The rising phases of the responses to flashes of increasing strength were found to fit with a biochemical model of phototransduction with an 'effective delay time' and 'characteristic time' of about 2 and 800 ms, respectively. 5. Spectral sensitivities were obtained over a wavelength range from 380 to 760 nm. The action spectrum, which peaked at 495 nm, followed the template described for photoreceptors in the macaque retina. Variation between rods in the position of the spectrum on the wavelength axis was small. 6. The scotopic luminosity function derived from human psychophysical experiments was found to agree well with the measured rod action spectrum after adjustments were made for lens absorption and photopigment self-screening in the intact eye. 7. Responses to steps of light rose monotonically to a maintained level, showing little or no relaxation. Nevertheless, the relationship between light intensity and steady-state response amplitude was shallower than that expected from simple response saturation. This is consistent with an adaptation mechanism acting on a rapid time scale. 8. Flash sensitivity fell with increasing intensities of background light according to Weber's law. Sensitivity was reduced twofold by lights evoking about 120 photoisomerizations per second. Background lights decreased the time to peak and the integration time of the flash response by up to 20%.

  11. Transient waves in finite viscoelastic rods

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, F. (Bologna Univ. (Italy). Ist. di Fisica); Nervosi, R. (Bologna Univ. (Italy))

    1980-11-29

    A method based on the Laplace transform is presented to compute wave-front expansions for transient waves in finite viscoelastic rods using the creep or the relaxation representation. The response is related to the basic solution of the semi-infinite problem, for which a series expansion is obtained by a recursive procedure. The convergence is guaranteed in any space-time domain if the material functions are entirely of exponential type. However, for numerical computation an acceleration of convergence is required and the Pade approximants turn out to be successful as shown by some examples.

  12. Rod-like nano-light harvester.

    Science.gov (United States)

    Ling, Jun; Zheng, Zhicheng; Köhler, Anna; Müller, Axel H E

    2014-01-01

    Imitating the natural "energy cascade" architecture, we present a single-molecular rod-like nano-light harvester (NLH) based on a cylindrical polymer brush. Block copolymer side chains carrying (9,9-diethylfluoren-2-yl)methyl methacrylate units as light absorbing antennae (energy donors) are tethered to a linear polymer backbone containing 9-anthracenemethyl methacrylate units as emitting groups (energy acceptors). These NLHs exhibit very efficient energy absorption and transfer. Moreover, we manipulate the energy transfer by tuning the donor-acceptor distance.

  13. Locally periodic Timoshenko rod: experiment and theory.

    Science.gov (United States)

    Díaz-de-Anda, A; Pimentel, A; Flores, J; Morales, A; Gutiérrez, L; Méndez-Sánchez, R A

    2005-05-01

    The flexural vibrations of a locally periodic rod, which consists of N unit cells, are discussed both from the experimental and theoretical points of view. Timoshenko's beam theory and the transfer matrix method are used to calculate the normal-mode frequencies and amplitudes. The theoretical values are then compared with the experimental ones, which are obtained using an electromagnetic acoustic transducer (EMAT). Good agreement between the numerical results and the experimental measurements is obtained. It is shown that as N grows, a band spectrum emerges.

  14. 4-rod RFQ linac for ion implantation

    Energy Technology Data Exchange (ETDEWEB)

    Fujisawa, Hiroshi; Hamamoto, Nariaki; Inouchi, Yutaka [Nisshin Electric Co. Ltd., Kyoto (Japan)

    1997-03-01

    A 34 MHz 4-rod RFQ linac system has been upgraded in both its rf power efficiency and beam intensity. The linac is able to accelerate in cw operation 0.83 mA of a B{sup +} ion beam from 0.03 to 0.91 MeV with transmission of 61 %. The rf power fed to the RFQ is 29 kW. The unloaded Q-value of the RFQ has been improved approximately 61 % to 5400 by copper-plating stainless steel cooling pipes in the RFQ cavity. (author)

  15. High-yield production of hydrophobins RodA and RodB from Aspergillus fumigatus in Pichia pastoris.

    Science.gov (United States)

    Pedersen, Mona Højgaard; Borodina, Irina; Moresco, Jacob Lange; Svendsen, Winnie Edith; Frisvad, Jens Christian; Søndergaard, Ib

    2011-06-01

    Hydrophobins are small fungal proteins with amphipatic properties and the ability to self-assemble on a hydrophobic/hydrophilic interface; thus, many technical applications for hydrophobins have been suggested. The pathogenic fungus Aspergillus fumigatus expresses the hydrophobins RodA and RodB on the surface of its conidia. RodA is known to be of importance to the pathogenesis of the fungus, while the biological role of RodB is currently unknown. Here, we report the successful expression of both hydrophobins in Pichia pastoris and present fed-batch fermentation yields of 200-300 mg/l fermentation broth. Protein bands of expected sizes were detected by SDS-PAGE and western blotting, and the identity was further confirmed by tandem mass spectrometry. Both proteins were purified using his-affinity chromatography, and the high level of purity was verified by silver-stained SDS-PAGE. Recombinant RodA as well as rRodB were able to convert a glass surface from hydrophilic to hydrophobic similar to native RodA, but only rRodB was able to decrease the hydrophobicity of a Teflon-like surface to the same extent as native RodA, while rRodA showed this ability to a lesser extent. Recombinant RodA and native RodA showed a similar ability to emulsify air in water, while recombinant RodB could also emulsify oil in water better than the control protein bovine serum albumin (BSA). This is to our knowledge the first successful expression of hydrophobins from A. fumigatus in a eukaryote host, which makes it possible to further characterize both hydrophobins. Furthermore, the expression strategy and fed-batch production using P. pastoris may be transferred to hydrophobins from other species.

  16. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  17. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Directory of Open Access Journals (Sweden)

    Kee-nam Song

    2012-01-01

    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  18. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  19. Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Thermal-hydraulic characteristic investigation on passive residual heat removal system(PRHRS)of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features.A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified.On the basis of theory analysis,a correlation of two-phase natural circulation was obtained,and relative errors of 95% test data were less than±16%.There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink,and its correlation of two-phase natural circulation system has been obtained.The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.

  20. Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

    CERN Document Server

    Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter

    2016-01-01

    A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...

  1. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  2. Computational analysis for prediction of pressure of PWR presurizer undertransient conditions

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A computer model has been developed for prediction of the pressure in thepressurizer undertransient conditions. In the model three separate thermodynamic regions which arenot required to be inthermal equilibrium have been considered. The mathematical model derived from the general conservation equations includesall of theimportant thermal-hydraulics phenomena occurring in the pressurizer,i.e., stratificationof the hot water andincoming cold water, bulk flashing and condensation, wall condensation, andinterfacial heat and masstransfer, etc. The bubble rising and rain-out models are developed to describe bulkflashing andcondensation, respectively. To obtain the wall condensation rate, a one-dimensionalheat conductionequation is solved by the pivoting method. The presented model will predict thepressure-time behaviorof a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreementwithavailable data on the CHASHMA nuclear power plant's pressurizer performance.

  3. PRETTA:A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS

    Institute of Scientific and Technical Information of China (English)

    阿谢德; 徐济鋆

    2001-01-01

    A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

  4. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-04-01

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

  6. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  7. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  8. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  9. Fatigue Crack Growth Rate of Type 347 Stainless Steel at the PWR Environment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Materials used in nuclear power plants are low alloy steel, stainless steel, and superalloy steel. Understanding the characteristics of these materials is important in the development of nuclear power plant related technology. Nb-stabilized Type 347 stainless steel is used for the coolant pressurizer surge line of Korea Standard Nuclear Power Plant (KSNPP). Surge line of PWR nuclear reactor are damaged by thermal fatigue due to thermal gradient during heat-up and cool-down, mechanical fatigue due to mechanical stress, and corrosion fatigue due to nuclear reactor water environment. Fatigue is an important factor which limits the life of structure. Fatigue crack growth rate curves in nuclear reactor environment are needed to evaluate the integrity of nuclear reactor structure but that result is not sufficient. In this study, fatigue crack growth rates at nuclear reactor environment are produced to evaluate integrity of nuclear power plant section 5

  10. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    CERN Document Server

    Kelly, G N; Charles, D; Hemming, C R

    1983-01-01

    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

  11. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  12. Stirring with ghost rods in a lid-driven cavity

    Science.gov (United States)

    Kumar, Pankaj; Chen, Jie; Stremler, Mark

    2009-11-01

    It has shown that passive fluid particles moving on periodic orbits can be used to `stir' a viscous fluid in a two-dimensional lid-driven cavity that exhibits a figure-eight flow pattern (Stremler & Chen 2007). Fluid motion in the vicinity of these particles produces ``ghost rod'' structures that behave like semi-permeable rods in the flow. Since these ghost rods are present due to the system dynamics, perturbations in the boundary conditions lead to variations in the existence and structure of the ghost rods. We discuss these variations and assess the role of ghost rods in mixing over a range of operating conditions for this system. The results suggest that ghost rods can play an important role in mixing for other counter-rotating flows.

  13. Vibration of the Package of Rods Linked by Spacer Grids

    Science.gov (United States)

    Zeman, V.; Hlaváč, Z.

    This paper deals with modelling and vibration analysis of the large package of identical parallel rods which are linked by transverse springs (spacer grids) placed on several level spacings. The vibration of rods is caused by the support plate motion. The rod discretization by FEM is based on Rayleigh beam theory. With respect to cyclic and central package rod symmetry, the system is decomposed to identical revolved rod segments. The modal synthesis method with condensation of the rod segments is used for modelling and determination of steady forced vibration of the whole system. The presented method is the first step to modelling of the nuclear fuel assembly vibration caused by kinematical excitation determined by motion of the support plates which are part of the reactor core.

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

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

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

  15. Developing and analyzing long-term fuel management strategies for an advanced Small Modular PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aeoi.org.ir

    2017-03-15

    Highlights: • Comprehensive introduction and supplementary concepts as a review paper. • Developing an integrated long-term fuel management strategy for a SMR. • High reliable 3-D core modeling over fuel pins against the traditional LRM. • Verifying the expert rules of large PWRs for an advanced small PWR. • Investigating large numbers of safety parameters coherently. - Abstract: In this paper, long-term fuel management (FM) strategies are introduced and analyzed for a new advanced Pressurized Light Water Reactor (PWR) type of Small Modular Reactors (SMRs). The FM strategies are developed to be safe and practical for implementation as much as possible. Safety performances, economy of fuel, and Quality Assurance (QA) of periodic equilibrium conditions are chosen as the main goals. Flattening power density distribution over fuel pins is the major method to ensure safety performance; also maximum energy output or permissible discharging burn up indicates economy of fuel fabrication costs. Burn up effects from BOC to EOC have been traced, studied, and highly visualized in both of transport lattice cell calculations and diffusion core calculations. Long-term characteristics are searched to gain periodical equilibrium characteristics. They are fissile changes, neutron spectrum, refueling pattern, fuel cycle length, core excess reactivity, average, and maximum burn up of discharged fuels, radial Power Peaking Factors (PPF), total PPF, radial and axial power distributions, batch effects, and enrichment effects for fine regulations. Traditional linear reactivity model have been successfully simulated and adapted via fine core and burn up calculations. Effects of high burnable neutron poison and soluble boron are analyzed. Different numbers of batches via different refueling patterns have been studied and visualized. Expert rules for large type PWRs have been influenced and well tested throughout accurate equilibrium core calculations.

  16. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    Energy Technology Data Exchange (ETDEWEB)

    Morlent, O.; Reuchet, J. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, 92 (France)

    2001-07-01

    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  17. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Energy Technology Data Exchange (ETDEWEB)

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)

    2009-07-01

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  18. Effect of aging on the PWR Chemical and Volume Control System

    Energy Technology Data Exchange (ETDEWEB)

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-06-01

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  19. Synthesis of homochiral tris-indanyl molecular rods

    DEFF Research Database (Denmark)

    Kjeldsen, Niels Due; Funder, Erik Daa; Gothelf, Kurt Vesterager

    2014-01-01

    Homochiral tris-indanyl molecular rods designed for supramolecular surface self-assembly were synthesized. The chiral indanol moiety was constructed via a Ti-mediated alkyne trimerization. Further manipulations resulted in a homochiral indanol monomer. This was employed as the precursor...... for successive Sonogashira and Ohira-Bestman reactions towards the homochiral tris-indanyl molecular rods. The molecular rods will be applied for scanning tunnelling microscopy studies of their surface self-assembly and chirality....

  20. Do rod signals control stimulus field prevalence in binocular rivalry?

    Science.gov (United States)

    Young, R S; Young, G A

    1979-02-01

    We tested the hypothesis that rod signals mediate the effect of luminance on field prevalence in binocular rivalry. Results show that (1) field prevalence saturates at a luminance that is less than 0.001 that of the value reported for rod saturation, (2) the spectral sensitivity inferred from the prevalence-luminance relationship is not scotopic, and (3) the prevalence-luminance relationship does not behave univariantly under all chromatic conditions. We conclude that rod signals alone do not control field prevalence.

  1. Vortex patterns in a superconducting-ferromagnetic rod

    Energy Technology Data Exchange (ETDEWEB)

    Romaguera, Antonio R. de C, E-mail: antonio.romaguera@df.ufrpe.b [Departamento de Fi' sica, Universidade Federal Rural de Pernambuco, 52171-900 Recife, Pernambuco (Brazil); Doria, Mauro M. [Departamento de Fi' sica dos Solidos, Universidade Federal do Rio de Janeiro, 21941-972 Rio de Janeiro (Brazil); Peeters, Francois M. [Departement Fysica, Universiteit Antwerpen, Groenenborgerlaan 171, B-2020 Antwerpen (Belgium)

    2010-10-01

    A superconducting rod with a magnetic moment on top develops vortices obtained here through 3D calculations of the Ginzburg-Landau theory. The inhomogeneity of the applied field brings new properties to the vortex patterns that vary according to the rod thickness. We find that for thin rods (disks) the vortex patterns are similar to those obtained in presence of a homogeneous magnetic field instead because they consist of giant vortex states. For thick rods novel patterns are obtained as vortices are curve lines in space that exit through the lateral surface.

  2. Biophysical mechanism of transient retinal phototropism in rod photoreceptors

    Science.gov (United States)

    Zhao, Xiaohui; Thapa, Damber; Wang, Benquan; Gai, Shaoyan; Yao, Xincheng

    2016-03-01

    Oblique light stimulation evoked transient retinal phototropism (TRP) has been recently detected in frog and mouse retinas. High resolution microscopy of freshly isolated retinas indicated that the TRP is predominated by rod photoreceptors. Comparative confocal microscopy and optical coherence tomography (OCT) revealed that the TRP predominantly occurred from the photoreceptor outer segment (OS). However, biophysical mechanism of rod OS change is still unknown. In this study, frog retinal slices, which open a cross section of retinal photoreceptor and other functional layers, were used to test the effect of light stimulation on rod OS. Near infrared light microscopy was employed to monitor photoreceptor changes in retinal slices stimulated by a rectangular-shaped visible light flash. Rapid rod OS length change was observed after the stimulation delivery. The magnitude and direction of the rod OS change varied with the position of the rods within the stimulated area. In the center of stimulated region the length of the rod OS shrunk, while in the peripheral region the rod OS tip swung towards center region in the plane perpendicular to the incident stimulus light. Our experimental result and theoretical analysis suggest that the observed TRP may reflect unbalanced disc-shape change due to localized pigment bleaching. Further investigation is required to understand biochemical mechanism of the observed rod OS kinetics. Better study of the TRP may provide a noninvasive biomarker to enable early detection of age-related macular degeneration (AMD) and other diseases that are known to produce retinal photoreceptor dysfunctions.

  3. Grey Rod Test in HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    Westinghouse/KAERI/KNF agreed to perform an irradiation test in the HANARO reactor to obtain irradiation data on the new grey rods that will be part of an AP1000 system. As a preliminary test, two samples containing pure Ag (Reference) and Ag-In-Cd materials provided by Westinghouse Electric Company (WEC) were inserted in a KNF irradiation capsule of 07M-13N. The specimens were irradiated for 95.19days (4 cycles) in the CT test hole of the HANARO of a 30MW thermal output to have a fast neutron fluence of 1.11x10{sup 21}(n/cm{sup 2}) (E>1.0MeV). This report provides all the test conditions and data obtained during the irradiation test of the grey rods in HANARO requested by Westinghouse. The test was prepared according to the meeting minutes (June 26, 2007) and the on-going subject test was stopped midway by the request of Westinghouse.

  4. Plastic Guidance Fins for Long Rod Projectiles .

    Directory of Open Access Journals (Sweden)

    Mark L. Bundy

    1997-10-01

    Full Text Available Projectile tail fins on long rod kinetic energy (KE penetrators serve the same purpose as fletchings (feathers on an arrow, namely, they help align the projectile axis with its velocity vector. This reduces the projectile's yaw and hence reduces its aerodynamic drag. In addition, a low yaw angle at target impact helps to maximise the projectile's target penetration. It is typical for projectiles to exit the gun muzzle and enter free flight at some ndn-zero yaw angle. Aerodynamic forces acting on yawed tail fins create a stabilising torque about the projectile's centre of gravity (CG. This torque can be increased by making the fin material lighter. Most conventional long rod penetrators fired from high performance guns have tail fins made from aluminium. However, aluminium can undergo catastrophic oxidation (rapid burning in-bore. Coating aluminium with Al/sub 2/O/sub 3/ {hardcoat prevents ignition of the substrate, provided solid propellant grain impacts do not chip the brittle hardcoat off the surface. Plastic is lighter than aluminium and less exothermic when oxidized. Therefore, other factors aside, it is conceivable that plastic fins could increase projectile stability while incurring less thermal erosion than aluminium. However, thermal loads are not the only concern when considering plastic as an alternative tail fin material. The mechanical strength of plastic is also a critical factor. This paper discusses some of the successes and failures of plastic fins, at least relatively thin fins, for use as KE stabilisers.

  5. Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1978-10-01

    This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor (BRPR).

  6. Rolls-Royce digital Rod Control System

    Energy Technology Data Exchange (ETDEWEB)

    Pouillot, M. [Rolls-Royce Civil Nuclear SAS (France)

    2010-07-01

    Full text of publication follows: Rolls-Royce has developed a new generation of Rod Control System, based on 40 years of experience. The fifth-generation Rod Control System (RCS) from Rolls-Royce offers a reliable, modular design with adaptability to your preferred platform, for modernization projects or new reactors. Flexible implementation provides the option for you to keep existing cabinets, which permits you to optimize installation approach. Main features for the power part: - Control Rod Drive Mechanism (CRDM) type: 3-coil. - Independent control of each sub-bank. - Each sub-bank is controlled by a cycler unit and 3 identical power racks, each including 4 identical power modules and a common power-supply module. - Coil-per-coil digital control: each power module embeds power-conversion, current-control, and current-monitoring functions for one coil. Control and monitoring are carried out by separate electronics in the module. Current is digitized and fully monitored by means of min-max templates. - A double-hold function is included: a power module assigned to a gripper will activate its coil if a fault risking to cause a reactor trip occurs. - Power modules are standardized, hot-pluggable and self-configured: a power module includes a set of parameters for each type of coil SG, MG, LC. The module recognizes the rack it is plugged in, and chooses automatically parameters to be used. Main benefits: - Reduced operational, maintenance, training, and inventory costs: standardization of power modules and integration of control and monitoring on the same PC-card lead to a drastic reduction of spare part types, and simplification of the system. - Easy maintenance: - Replacement of a power module solves nearly all failures due to current control or monitoring for a coil. It is done instantly thanks to hot-plug capability. - On the front plate of power-modules, LEDs provide useful information for diagnostic: current setpoint from cycler, output current bar

  7. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  8. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    OpenAIRE

    Alroumi Fawaz; Kim Donghoon; Schow Ryan; Jevremovic Tatjana

    2016-01-01

    Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is ...

  9. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-07-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  10. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  11. Biomechanical comparison of cervical fixation via transarticular facet screws without rods versus lateral mass screws with rods.

    Science.gov (United States)

    Yi, Seong; Rim, Dae-Cheol; Nam, Ki-Se; Keem, Sang-Hyun; Murovic, Judith A; Lim, Jesse; Park, Jon

    2015-04-01

    Transarticular facet screws restore biomechanical stability to the cervical spine when posterior cervical anatomy has been compromised. This study compares the more recent, less invasive, and briefer transarticular facet screw system without rods with the lateral mass screw system with rods. For this study, 6 human cervical spines were obtained from cadavers. Transarticular facet screws without rods were inserted bilaterally into the inferior articular facets at the C5-C6 and C5-C6-C7 levels. Lateral mass screws with rods were inserted bilaterally at the same levels using Magerl's technique. All specimens underwent range of motion (ROM) testing by a material testing machine for flexion, extension, lateral bending, and axial rotation. Both fixation methods, transarticular facet screws without rods and lateral mass screws with rods, reduced all ROM measurements and increased spinal stiffness. No statistically significant differences between the 2 stabilization methods were found in ROM measurements for 1-level insertions. However, in 2-level insertions, ROM for the nonrod transarticular facet screw group was significantly increased for flexion-extension and lateral bending. Transarticular facet screws without rods and lateral mass screws with rods had similar biomechanical stability in single-level insertions. For 2-level insertions, transarticular facet screws without rods are a valid option in cervical spine repair. Copyright © 2015 Elsevier Inc. All rights reserved.

  12. Patterns of rod proliferation in deep-sea fish retinae.

    Science.gov (United States)

    Fröhlich, E; Negishi, K; Wagner, H J

    1995-07-01

    In a sample of 37 species of deep-sea fish species from the sea floor of the Porcupine Seabight and the Gobal spur (North Atlantic) we investigated the overall structure of the retina with special respect for the organization of rods, their length and their arrangement in multiple banks. Using an immunocytochemical marker for cell proliferation (PCNA) we studied the mechanisms of rod proliferation, and, by means of serial section reconstruction followed their integration into the existing population of rods. Furthermore, in three different species we have observed growth related changes in retinal thickness, rod density and proliferation activity. We found evidence for two different principles for the organization of rods in these deep-sea fish retinae. In the first group of species represented by Nematonurus armatus and Coryphaenoides guentheri we found rods to be rather short (20-30 microns) and arranged in three and more banks. In these species rod proliferation occurred in a single band of cells immediately vitread of the external limiting membrane, thus showing a high degree of spatial and temporal order. In these species, young rods are inserted just sclerad of the external limiting membrane and the older outer segments pushed away from the incoming light towards the back of the eye. This may be linked to a progressive loss of function of the older rods and might represent an alternative mechanism to the disk shedding in other vertebrates. In the second population (e.g. Conocara macroptera, Alepocephalus agassizii) we observed considerably longer rod outer segments (60-80 microns) forming no more than two layers. These retinae had rod precursors arranged in disseminated clusters throughout the outer nuclear layer indicating the lack of clear spatio-temporal order in mitotic activity along with a more statistical pattern of integration of the newly formed outer segments. In our sample of species both populations were of about equal size suggesting that the two

  13. Investigation of axial power gradients near a control rod tip

    Energy Technology Data Exchange (ETDEWEB)

    Loberg, John, E-mail: John.Loberg@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Osterlund, Michael, E-mail: Michael.Osterlund@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Bejmer, Klaes-Hakan, E-mail: Klaes-Hakan.Bejmer@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Blomgren, Jan, E-mail: Jan.Blomgren@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Kierkegaard, Jesper, E-mail: Jesper.Kierkegaar@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden)

    2011-07-15

    Highlights: > Pin power gradients near BWR control rod tips have been investigated. > A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. > Small nodes increases pin power gradients; standard nodes underestimates gradients. > The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, {approx}15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

  14. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks; Analise qualitativa da politica de manutencoes dos sistemas de um PWR tipico por redes neurais artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lourenco, Victor Hugo Moreno

    2010-02-15

    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  15. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

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

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-02-01

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

  18. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2015-09-01

    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  19. PFM Analysis for Pre-Existing Cracks on Alloy 182 Weld in PWR Primary Water Environment using Monte Carlo Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.

  20. Theoretical estimation of the impact velocity during the PWR spent drop in water condition

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Oh Joon; Park, Nam Gyu; Lee, Seong Ki; Kim, Jae Ik [KEPCO NF, Daejeon (Korea, Republic of)

    2016-06-15

    The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the 3×3 short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.

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

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

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

  2. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  3. Switch isotropic/anisotropic wettability via dual-scale rods

    Directory of Open Access Journals (Sweden)

    Yang He

    2014-10-01

    Full Text Available It is the first time to demonstrate the comparison of isotropic/anisotropic wettability between dual-scale micro-nano-rods and single-scale micro-rods. Inspired by the natural structures of rice leaf, a series of micro-nano-rods and micro-rods with different geometric parameters were fabricated using micro-fabrication technology. Experimental measured apparent contact angles and advancing and receding contact angles from orthogonal orientations were characterized. The difference of contact angles from orthogonal orientation on dual-scale rods was much smaller than those on single-scale rods in both static and dynamic situation. It indicated that the dual-scale micro-nano-rods showed isotropic wettability, while single-scale micro-rods showed anisotropic wettability. The switch of isotropic/anisotropic wettability could be illustrated by different wetting state and contact line moving. It offers a facial way to switch isotropic/anisotropic wettability of the surface via dual-scale or single-scale structure.

  4. Switch isotropic/anisotropic wettability via dual-scale rods

    Science.gov (United States)

    He, Yang; Jiang, Chengyu; Wang, Shengkun; Ma, Zhibo; Yuan, Weizheng

    2014-10-01

    It is the first time to demonstrate the comparison of isotropic/anisotropic wettability between dual-scale micro-nano-rods and single-scale micro-rods. Inspired by the natural structures of rice leaf, a series of micro-nano-rods and micro-rods with different geometric parameters were fabricated using micro-fabrication technology. Experimental measured apparent contact angles and advancing and receding contact angles from orthogonal orientations were characterized. The difference of contact angles from orthogonal orientation on dual-scale rods was much smaller than those on single-scale rods in both static and dynamic situation. It indicated that the dual-scale micro-nano-rods showed isotropic wettability, while single-scale micro-rods showed anisotropic wettability. The switch of isotropic/anisotropic wettability could be illustrated by different wetting state and contact line moving. It offers a facial way to switch isotropic/anisotropic wettability of the surface via dual-scale or single-scale structure.

  5. Probabilistic thermo-chemical analysis of a pultruded composite rod

    NARCIS (Netherlands)

    Baran, Ismet; Tutum, Cem C.; Hattel, Jesper H.

    2012-01-01

    In the present study the deterministic thermo-chemical pultrusion simulation of a composite rod taken from the literature [7] is used as a validation case. The predicted centerline temperature and cure degree profiles of the rod match well with those in the literature [7]. Following the validation c

  6. Probabilistic thermo-chemical analysis of a pultruded composite rod

    DEFF Research Database (Denmark)

    Baran, Ismet; Tutum, Cem Celal; Hattel, Jesper Henri

    2012-01-01

    In the present study the deterministic thermo-chemical pultrusion simulation of a composite rod taken from the literature [7] is used as a validation case. The predicted centerline temperature and cure degree profiles of the rod match well with those in the literature [7]. Following the validation...

  7. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  8. Directed Assembly of colloidal rods, spheres and their mixtures

    NARCIS (Netherlands)

    Bakker, H.E.

    2017-01-01

    The directed assembly of colloidal rods, spheres and their mixtures in external fields such as gravity, electric fields and shear flow was investigated. Using confocal microscopy we determined the experimental phase diagram of a binary mixture of colloidal silica rods and spheres that form a binary

  9. Isotropic-to-nematic nucleation in suspensions of colloidal rods

    NARCIS (Netherlands)

    Cuetos, A.; van Roij, R.H.H.G.; Dijkstra, M.

    2008-01-01

    Using computer simulations, we study the isotropic-to-nematic nucleation in a fluid of colloidal hard rods as well as in a mixture of colloidal rods and non-adsorbing polymer. In order to follow the transformation of the system from the isotropic to the nematic phase, we use a new cluster criterion

  10. Numerical investigation of flow past a row of rectangular rods

    Directory of Open Access Journals (Sweden)

    S.Ul. Islam

    2016-09-01

    Full Text Available A numerical study of uniform flow past a row of rectangular rods with aspect ratio defined as R = width/height = 0.5 is performed using the Lattice Boltzmann method. For this study the Reynolds number (Re is fixed at 150, while spacings between the rods (g are taken in the range from 1 to 6. Depending on g, the flow is classified into four patterns: flip-flopping, nearly unsteady-inphase, modulated inphase-antiphase non-synchronized and synchronized. Sudden jumps in physical parameters were observed, attaining either maximum or minimum values, with the change in flow patterns. The mean drag coefficient (Cdmean of middle rod is higher than the second and fourth rod for flip-flopping pattern while in case of nearly unsteady-inphase the middle rod attains minimum drag coefficient. It is also found that the Strouhal number (St of first, second and fifth rod decreases as g increases while that of other two have mixed trend. The results further show that there exist secondary interaction frequencies together with primary vortex shedding frequency due to jet in the gap between rods for 1 ⩽ g ⩽ 3. For the average values of Cdmean and St, an empirical relation is also given as a function of gap spacing. This relation shows that the average values of Cdmean and St approach to those of single rectangular rod with increment in g.

  11. Approximation in LQG control of a thermoelastic rod

    Science.gov (United States)

    Gibson, J. S.; Rosen, I. G.; Tao, G.

    1989-01-01

    Control and estimator gains are computed for linear-quadratic-Gaussian (LQG) optimal control of the axial vibrations of a thermoelastic rod. The computations are based on a modal approximation of the partial differential equations representing the rod, and convergence of the approximations to control and estimator gains is the main issue.

  12. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  13. Chitosan rod reinforced by self-crosslinking through thermal treatment

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Chitosan (CS) rods were reinforced at high temperatures to form network structure by self-crosslinking of amino groups.Properties of treated CS rods were studied by FTIR spectroscopy,intrinsic viscosity measurement,mechanical properties testing and water absorption measurement.The FTIR spectra indicated that the CS configuration was transformed from β-CS for untreated CS rods to α-CS for thermally treated CS rods.Meanwhile,the crosslinking also occurred between amino groups of CS.Due to the increase in the crosslinking degree,the intrinsic viscosity increased with the rising of temperature.It was found that the network structure enhanced the bending strength of CS rods,which reached 154.8 MPa when CS rods were treated at 140℃ for 2 h.Thermal treatment also reduced the water absorption of CS rods.Due to the improved mechanical properties,thermally treated CS rods could be used as a novel device for internal fixation of bone fracture.

  14. 49 CFR 236.794 - Rod, up-and-down.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Rod, up-and-down. 236.794 Section 236.794 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION..., up-and-down. A rod used for connecting the semaphore arm to the operating mechanism of a signal....

  15. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. the combinatorial optimization problem of grouping fuel rods into a given number of rod groups with the same enrichment, and the problem of determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by the linear combination C{sub 1}X + C{sub 2}X{sub G}, where X and X{sub G} stand, respectively, for control variables giving constraint to the local power peaking factor and the gadolinium rod power. C{sub 1} and C{sub 2} are user-definable weighting factors to accommodate design preferences. The algorithm for solving this combinatorial optimization problem starts by finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering. This latter problem is solved using the method of approximation programming. A practical application is shown for a contemporary 8 x 8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  16. Biomolecular rods and tubes in nanotechnology

    Science.gov (United States)

    Bittner, Alexander M.

    2005-02-01

    Biomolecules are vitally important elements in nanoscale science and also in future nanotechnology. Their shape and their chemical and physical functionality can give them a big advantage over inorganic and organic substances. While this becomes most obvious in proteins and peptides, with their complicated, but easily controlled chemistry, other biomolecular substances such as DNA, lipids and carbohydrates can also be important. In this review, the emphasis is on one-dimensional molecules and on molecules that self-assemble into linear structures, and on their potential applications. An important aspect is that biomolecules can act as templates, i.e. their shape and chemical properties can be employed to arrange inorganic substances such as metals or metal compounds on the nanometre scale. In particular, rod- and tube-like nanostructures can show physical properties that are different from those of the bulk material, and thus these structures are likely to be a basis for new technology.

  17. Electric Fuel Rod Simulator Fabrication at ORNL

    Science.gov (United States)

    Ott, Larry J.; McCulloch, Reg

    2004-02-01

    Commercial vendors could not supply the high-quality, highly instrumented electric fuel rod simulators (FRS) required for large thermal-hydraulic safety-oriented experiments at the Oak Ridge National Laboratory (ORNL) in the 1970s and early 1980s. Staff at ORNL designed, developed, and manufactured the simulators utilized in these safety experiments. Important FRS design requirements include (1) materials of construction, (2) test power requirements and availability, (3) experimental test objectives, (4) supporting thermal analyses, and (5) extensive quality control throughout all phases of FRS fabrication. This paper will present an overview of these requirements (design, analytics, and quality control) as practiced at ORNL to produce a durable high-quality FRS.

  18. Axial thermal rotation of slender rods.

    Science.gov (United States)

    Li, Dichuan; Fakhri, Nikta; Pasquali, Matteo; Biswal, Sibani Lisa

    2011-05-06

    Axial rotational diffusion of rodlike polymers is important in processes such as microtubule filament sliding and flagella beating. By imaging the motion of small kinks along the backbone of chains of DNA-linked colloids, we produce a direct and systematic measurement of axial rotational diffusivity of rods both in bulk solution and near a wall. The measured diffusivities decrease linearly with the chain length, irrespective of the distance from a wall, in agreement with slender-body hydrodynamics theory. Moreover, the presence of small kinks does not affect the chain's axial diffusivity. Our system and measurements provide insights into fundamental axial diffusion processes of slender objects, which encompass a wide range of entities including biological filaments and linear polymer chains.

  19. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, T.A.

    1997-06-01

    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  20. Study of the anti-sand sucker rod pump

    Science.gov (United States)

    Wei, Hangxin; Lv, Bingxin; Xi, Wenkui; Yi, Peng

    2017-06-01

    In order to solve the problem of sand stuck in the sucker rod pump, an anti-sand sucker rod pump is designed. The anti-sand sucker rod pump includes the conventional sucker rod pump and the swirl flow device. The sand particles can be separated from the oil in the swirl flow device, so the plunger of the sucker rod pump cannot be stuck. The motion equation of the sand particles in oil is deduced. The virtual model of the swirl flow device is built in GAMBIT software. And simulation of solid-liquid two phase flow is simulated in software FLUENT. The simulation results show that the swirl flow device can realize the sand particles separation from the oil completely. So the pump can have the effect of anti-sands.

  1. Investigating the optical XNOR gate using plasmonic nano-rods

    Science.gov (United States)

    Akhlaghi, Majid; Kaboli, Milad

    2016-04-01

    In this paper, a coherent perfect absorption (CPA)-type XNOR gate based on plasmonic nano particle is proposed. It consists of two plasmonic nano rod arrays on top of two parallel arms with quartz substrate. The operation principle is based on the absorbable formation of a conductive path in the dielectric layer of a plasmonic nano-particles waveguide. Since the CPA efficiency depends strongly on the number of plasmonic nano-rod and the nano rod location, an efficient binary optimization method based the Particle Swarm Optimization (PSO) algorithm is used to design an optimized array of the plasmonic nano-rod in order to achieve the maximum absorption coefficient in the 'off' state and the minimum absorption coefficient in the 'on' state. In Binary PSO (BPSO), a group of birds consists a matrix with binary entries, control the presence ('1‧) or the absence ('0‧) of nano rod in the array.

  2. Semiconductor Quantum Rods as Single Molecule FluorescentBiological Labels

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Aihua; Gu, Weiwei; Boussert, Benjamine; Koski, Kristie; Gerion, Daniele; Manna, Liberato; Le Gros, Mark; Larabell, Carolyn; Alivisatos, A. Paul

    2006-05-29

    In recent years, semiconductor quantum dots have beenapplied with great advantage in a wide range of biological imagingapplications. The continuing developments in the synthesis of nanoscalematerials and specifically in the area of colloidal semiconductornanocrystals have created an opportunity to generate a next generation ofbiological labels with complementary or in some cases enhanced propertiescompared to colloidal quantum dots. In this paper, we report thedevelopment of rod shaped semiconductor nanocrystals (quantum rods) asnew fluorescent biological labels. We have engineered biocompatiblequantum rods by surface silanization and have applied them fornon-specific cell tracking as well as specific cellular targeting. Theproperties of quantum rods as demonstrated here are enhanced sensitivityand greater resistance for degradation as compared to quantum dots.Quantum rods have many potential applications as biological labels insituations where their properties offer advantages over quantumdots.

  3. Fatigue Life Improving of Drill Rod by Inclusion Control

    Science.gov (United States)

    Wang, Linzhu; Yang, Shufeng; Li, Jingshe; Liu, Wei; Zhou, Yinghao

    2016-08-01

    Large and hard inclusions often deteriorate the service performance and reduce the fatigue lifetime of drill rods. In this paper, the main reasons of the rupture of drill rods were analyzed by the examination of their fracture and it is found that the large inclusions were the main reason of breakage of rod drill. The inclusions were high of Ca content or Al2O3 rich. Smaller and better deformability inclusions were obtained by the optimization of refining slag, calcium treatment process and the flow control devices of tundish. Results of industrial experiment after optimization show that total oxygen content of drill rods decreased by more than 50%, macro-inclusions weight fraction decreased from about 4 mg/10 kg to about 0.3 mg/10 kg and the micro-inclusions average size decreased from 6 to 3.6 μm. The average using times of drill rods after optimization were increased by about 60%.

  4. Control wetting state transition by micro-rod geometry

    Science.gov (United States)

    He, Yang; Jiang, Chengyu; Wang, Shengkun; Yin, Hengxu; Yuan, Weizheng

    2013-11-01

    Understanding the effect of micro-structure geometry on wetting state transition is important to design and control surface wettability. Micro-rod model was proposed and the relationship between micro-rod geometry and wetting state was investigated in the paper taking into account only the surface roughness and neglecting the chemistry interaction. Micro-rods with different geometric parameters were fabricated using micro-fabrication technology. Their contact angles were measured and compared with theoretical ones. The experimental results indicated that increasing the height and decreasing the space of micro-rod may result in Cassie wetting state, while decreasing the height and increasing the space may result in Wenzel wetting state. A suspended wetting state model due to scallops was proposed. The wetting state transition was interpreted by intruding height, de-pinning and sag mechanism. It may offer a facile way to control the surface wetting state transition by changing the geometry of micro-rod.

  5. Controllable single photon stimulation of retinal rod cells

    CERN Document Server

    Phan, Nam Mai; Bessarab, Dmitri A; Krivitsky, Leonid A

    2013-01-01

    Retinal rod cells are commonly assumed to be sensitive to single photons [1, 2, 3]. Light sources used in prior experiments exhibit unavoidable fluctuations in the number of emitted photons [4]. This leaves doubt about the exact number of photons used to stimulate the rod cell. In this letter, we interface rod cells of Xenopus laevis with a light source based on Spontaneous Parametric Down Conversion (SPDC) [5], which provides one photon at a time. Precise control of generation of single photons and directional delivery enables us to provide unambiguous proof of single photon sensitivity of rod cells without relying on the statistical assumptions. Quantum correlations between single photons in the SPDC enable us to determine quantum efficiency of the rod cell without pre-calibrated reference detectors [6, 7, 8]. These results provide the path for exploiting resources offered by quantum optics in generation and manipulation of light in visual studies. From a more general perspective, this method offers the ult...

  6. Entropical Colloidal Interaction in Rod-like Molecules Solution

    Science.gov (United States)

    Hohlfeld, Evan B.; Lin, Keng-Hui; Zeri, Ana Carolina; Crocker, John C.; Yodh, Arjun G.

    2001-03-01

    We report direct measurements of the functional form of the depletion interaction between two colloidal spheres in a rod-like molecule suspension the line-scanned optical tweezer. The rod like moleculses are bacteriaphage fd with length (L) 880 nm and diameter(D) 6.5 nm and TMV with L 300 nm and D 200 nm. We probed different ratios of sphere radius R and rod length L and compared with theoretical models of Yaman^. The experimental data agrees with the model with slight discrepancy due to the flexibility of rod molecules. At high salt concentration, we also observed the steric repulsion due rod molecule stuck on the spheres. We gratefully acknowledge support from the NSF (DMR-9623441) and MRSEC (DMR-9632598). K. Yaman, C. Jeppesen and C.M. Marques, Europhys. Lett., 42, 221 (1998).

  7. Jourdain Principle of a Super-Thin Elastic Rod Dynamics

    Institute of Scientific and Technical Information of China (English)

    XUE Yun; SHANG Hui-Lin

    2009-01-01

    A super thin elastic rod is modeled with a background of DNA super coiling structure, and its dynamics is discussed based on the Jourdain variation. The cross section of the rod is taken as the object of this study and two velocity spaces about arc coordinate and the time are obtained respectively. Virtual displacements of the section on the two velocity spaces are defined and can be expressed in terms of Jourdaln variation. Jourdain principles of a super thin elastic rod dynamics on arc coordinate and the time velocity space are established,respectively, which show that there are two ways to realize the constraint conditions. If the constitutive relation of the rod is linear, the Jourdaln principle takes the Euler-Lagrange form with generalized coordinates. The Kirchhoff equation, Lagrange equation and Appell equation can be derived from the present Jourdaln principle.While the rod subjected to a surface constraint, Lagrange equation with undetermined multipliers may be derived.

  8. Systematic parametric design/calculation of the piston rod unit

    Science.gov (United States)

    Kacani, V.

    2015-08-01

    In this article a modern and economic method for the strength calculation of the piston rod unit and its components under different operating conditions will be presented. Herefore the commercial FEA - Software will be linked with the company-owned calculation tools. The parametric user input will be followed by an automatic Pre- and Postprocessing. Afterwards the strength calculation is processed on all critical points of the piston rod connection, assisted by an extra module, based on general standards and special codes for reciprocating compressors. In this process most arrangements of the piston rod unit as well as the special geometries of the single-components (piston, piston rod and piston nut) can be considered easily. In this article the modeling of the notches, especially on the piston rod, piston as well as the piston nut will be covered in detail.

  9. Characterization of PWR vessel steel tearing under severe accident condition temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Matheron, Philippe, E-mail: philippe.matheron@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Chapuliot, Stephane, E-mail: stephane.chapuliot@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Nicolas, Laetitia, E-mail: laetitia.nicolas@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Laboratoire de Mecanique des Structures Industrielles Durables, UMR CNRS-EDF 2832, 1 avenue du General de Gaulle, F-92141 Clamart (France); Koundy, Vincent, E-mail: vincent.koundy@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Caroli, Cataldo, E-mail: cataldo.caroli@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer We characterized French PWR vessel steel tearing resistance at high temperatures. Black-Right-Pointing-Pointer Tearing tests on Compact Tension (CT) specimens were carried out. Black-Right-Pointing-Pointer The variability of tearing properties with PWR vessels specifications was studied. Black-Right-Pointing-Pointer We propose a tearing criterion (energy parameter Gfr) at high temperatures. - Abstract: In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead

  10. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C{sub 1}X+C{sub 2}X{sub G}, where X and X{sub G} stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C{sub 1} and C{sub 2} are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  11. NRL-Regulated Transcriptome Dynamics of Developing Rod Photoreceptors.

    Science.gov (United States)

    Kim, Jung-Woong; Yang, Hyun-Jin; Brooks, Matthew John; Zelinger, Lina; Karakülah, Gökhan; Gotoh, Norimoto; Boleda, Alexis; Gieser, Linn; Giuste, Felipe; Whitaker, Dustin Thad; Walton, Ashley; Villasmil, Rafael; Barb, Jennifer Joanna; Munson, Peter Jonathan; Kaya, Koray Dogan; Chaitankar, Vijender; Cogliati, Tiziana; Swaroop, Anand

    2016-11-22

    Gene regulatory networks (GRNs) guiding differentiation of cell types and cell assemblies in the nervous system are poorly understood because of inherent complexities and interdependence of signaling pathways. Here, we report transcriptome dynamics of differentiating rod photoreceptors in the mammalian retina. Given that the transcription factor NRL determines rod cell fate, we performed expression profiling of developing NRL-positive (rods) and NRL-negative (S-cone-like) mouse photoreceptors. We identified a large-scale, sharp transition in the transcriptome landscape between postnatal days 6 and 10 concordant with rod morphogenesis. Rod-specific temporal DNA methylation corroborated gene expression patterns. De novo assembly and alternative splicing analyses revealed previously unannotated rod-enriched transcripts and the role of NRL in transcript maturation. Furthermore, we defined the relationship of NRL with other transcriptional regulators and downstream cognate effectors. Our studies provide the framework for comprehensive system-level analysis of the GRN underlying the development of a single sensory neuron, the rod photoreceptor. Published by Elsevier Inc.

  12. Thermal post-bunkling analyses of functionally graded material rod

    Institute of Scientific and Technical Information of China (English)

    ZHAO Feng-qun; WANG Zhong-min; LIU Hong-zhao

    2007-01-01

    The non-linear governing differential equations of immovably simply supported functionally graded material (FGM) rod subjected to thermal loads were derived.The thermal post-buckling behaviors of FGM rod made of ZrO2 and Ti-6A1-4Vwere analyzed by shooting method. Firstly, the thermal post-buckling equilibrium paths of the FGM rod with different gradient index in the uniform temperature field were plotted,and compared with the behaviors of the homogeneous rods made of ZrO2 and Ti-6A1-4V materials, respectively. For given value of end rotation angles, the influence of gradient index on the thermal post-buckling behaviors of FGM rod was discussed. Secondly, the thermal post-buckling characteristics of the FGM rod were analyzed when the temperature difference parameter is changed while the bottom temperature parameter remains constant, and when the bottom temperature parameter is changed while the temperature difference parameter remains constant, and compared with the characteristics of the two homogeneous material rods.

  13. Anisotropy in CdSe quantum rods

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liang-shi

    2003-09-01

    The size-dependent optical and electronic properties of semiconductor nanocrystals have drawn much attention in the past decade, and have been very well understood for spherical ones. The advent of the synthetic methods to make rod-like CdSe nanocrystals with wurtzite structure has offered us a new opportunity to study their properties as functions of their shape. This dissertation includes three main parts: synthesis of CdSe nanorods with tightly controlled widths and lengths, their optical and dielectric properties, and their large-scale assembly, all of which are either directly or indirectly caused by the uniaxial crystallographic structure of wurtzite CdSe. The hexagonal wurtzite structure is believed to be the primary reason for the growth of CdSe nanorods. It represents itself in the kinetic stabilization of the rod-like particles over the spherical ones in the presence of phosphonic acids. By varying the composition of the surfactant mixture used for synthesis we have achieved tight control of the widths and lengths of the nanorods. The synthesis of monodisperse CdSe nanorods enables us to systematically study their size-dependent properties. For example, room temperature single particle fluorescence spectroscopy has shown that nanorods emit linearly polarized photoluminescence. Theoretical calculations have shown that it is due to the crossing between the two highest occupied electronic levels with increasing aspect ratio. We also measured the permanent electric dipole moment of the nanorods with transient electric birefringence technique. Experimental results on nanorods with different sizes show that the dipole moment is linear to the particle volume, indicating that it originates from the non-centrosymmetric hexagonal lattice. The elongation of the nanocrystals also results in the anisotropic inter-particle interaction. One of the consequences is the formation of liquid crystalline phases when the nanorods are dispersed in solvent to a high enough

  14. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1980-November 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1981-02-01

    Four tasks are reported: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  15. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

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

  17. VOF Calculations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    Directory of Open Access Journals (Sweden)

    M. Murase

    2012-01-01

    Full Text Available We improved the computational grid and schemes in the VOF (volume of fluid method with the standard − turbulent model in our previous study to evaluate CCFL (countercurrent flow limitation characteristics in a full-scale PWR hot leg (750 mm diameter, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa. In this paper, therefore, to evaluate applicability of the VOF method to different fluid properties and a different scale, we did numerical simulations for full-scale air-water conditions and the 1/15-scale air-water tests (50 mm diameter, respectively. The results calculated for full-scale conditions agreed well with CCFL data and showed that CCFL characteristics in the Wallis diagram were mitigated under 1.5 MPa steam-water conditions comparing with air-water flows. However, the results calculated for the 1/15-scale air-water tests greatly underestimated the falling water flow rates in calculations with the standard − turbulent model, but agreed well with the CCFL data in calculations with a laminar flow model. This indicated that suitable calculation models and conditions should be selected to get good agreement with data for each scale.

  18. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  19. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  20. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    Directory of Open Access Journals (Sweden)

    d’Auria Francesco

    2005-01-01

    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  1. Replacement of Co-base alloy for radiation exposure reduction in the primary system of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, Sun Jin [Hanyang Univ., Seoul (Korea, Republic of)

    1996-01-01

    Of numerous Co-free alloys developed to replace Co-base stellite used in valve hardfacing material, two iron-base alloys of Armacor M and Tristelle 5183 and one nickel-base alloy of Nucalloy 488 were selected as candidate Co-free alloys, and Stellite 6 was also selected as a standard hardfacing material. These four alloys were welded on 316SS substrate using TIG welding method. The first corrosion test loop of KAERI simulating the water chemistry and operation condition of the primary system of PWR was designed and fabricated. Corrosion behaviors of the above four kinds of alloys were evaluated using this test loop under the condition of 300 deg C, 1500 psi. Microstructures of weldment of these alloys were observed to identify both matrix and secondary phase in each weldment. Hardnesses of weld deposit layer including HAZ and substrate were measured using micro-Vickers hardness tester. The status on the technology of Co-base alloy replacement in valve components was reviewed with respect to the classification of valves to be replaced, the development of Co-free alloys, the application of Co-free alloys and its experiences in foreign NPPs, and the Co reduction program in domestic NPPs and industries. 18 tabs., 20 figs., 22 refs. (Author).

  2. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  3. Study of chemical additives in the cementation of radioactive waste of PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Vanessa Mota; Tello, Cledola Cassia Oliveira de, E-mail: vanessamotavieira@gmail.com, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2012-03-15

    In this research it has been studied the effects of chemical admixtures in the cementation process of radioactive wastes. These additives are used to improve the properties of waste cementation process, both of the paste and of the solidified product. However there are a large variety of these materials that are frequently changed or taken out of the market. Then it is essential to know the commercially available materials and their effects. The tests were carried out with a solution simulating the evaporator concentrate waste coming from PWR nuclear reactors. It was cemented using two formulations, A and B, incorporating higher or lower amount of waste, respectively. It was added chemical admixtures from two manufacturers (S and H), which were: accelerators, set retarders and superplasticizers. The experiments were organized by a factorial design 23. The measured parameters were: the viscosity, the setting time, the paste and product density and the compressive strength. The parameter evaluated in this study was the compressive strength at age of 28 days, is considered essential security issues relating to the handling, transport and storage of cemented waste product. The results showed that the addition of accelerators improved the compressive strength of the cemented products. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    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.

  5. Numerical Simulation of Size Effects on Countercurrent Flow Limitation in PWR Hot Leg Models

    Directory of Open Access Journals (Sweden)

    I. Kinoshita

    2012-01-01

    Full Text Available We have previously done numerical simulations using the two-fluid model implemented in the CFD software FLUENT6.3.26 to investigate effects of shape of a flow channel and its size on CCFL (countercurrent flow limitation characteristics in PWR hot leg models. We confirmed that CCFL characteristics in the hot leg could be well correlated with the Wallis parameters in the diameter range of 0.05 m≤D≤0.75 m. In the present study, we did numerical simulations using the two-fluid model for the air-water tests with D=0.0254 m to determine why CCFL characteristics for D=0.0254 m were severer compared with those in the range, 0.05 m≤D≤0.75 m. The predicted CCFL characteristics agreed with the data for D=0.0254 m and indicated that the CCFL difference between D=0.0254 m and 0.05 mm≤D≤0.75 mm was caused by the size effect and not by other factors.

  6. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2017-03-27

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  7. Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

    Energy Technology Data Exchange (ETDEWEB)

    Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steve [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spears, Bob [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kosbab, Ben [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-26

    This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA models for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.

  8. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  9. PWR composite materials use. A particular case of safety-related service water pipes

    Energy Technology Data Exchange (ETDEWEB)

    Pays, M.F.; Le Courtois, T

    1997-11-01

    This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

  10. Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

    Directory of Open Access Journals (Sweden)

    Ge Shao

    2013-01-01

    Full Text Available To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.

  11. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  12. Studies of residual stress measurement and analysis techniques for a PWR dissimilar weld joint

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Naoki, E-mail: naoki2_ogawa@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Muroya, Itaru; Iwamoto, Youichi; Ohta, Takahiro; Ochi, Mayumi; Hojo, Kiminobu [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Ogawa, Kazuo [Japan Nuclear Energy Safety Organization, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2012-02-15

    For evaluation of the PWSCC crack propagation behavior, a test model was produced using the same fabrication process of Japanese PWR plants and the stress distribution change was measured during a fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end and an operation condition test. For confirmation of validity of the numerical estimation method of the stress distribution, FE analysis was performed to calculate the stress distributions for each fabrication process. From the validation procedure, a standard residual stress evaluation method was established. Furthermore for consideration of characteristics of PWSCC's propagation behavior of the dissimilar welding joint of the safe end nozzles, the influence coefficients at the deepest point for the stress intensity factors of axial cracks with large aspect ratio a/c (crack depth/half of surface crack length) was prepared. The crack shape was assumed a rectangular shape and the stress intensity factors at the deepest point of the crack were calculated with change of crack depth using FE analysis. By using these stress distribution and influence coefficients, a behavior of a PWSCC crack propagation at the safe end nozzles can be estimated easily and rationally.

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

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

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

  14. Performance of monosphere new gel type ion exchange resins for condensate polisher at PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, S.; Nakamura, M.; Asou, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Izumi, T.; Deguchi, T.; Ino, T.; Hagiwara, M.

    1998-12-31

    There are two kinds of ion exchange resins of gel type and porous one which are used as condensate polisher in LWR nuclear power plants. In order to estimate the performance of these resins on the condensate polisher at the secondary cycle of Japanese PWR plants, a column test was performed setting the column test device in Ohi power station unit 1 of the Kansai Electric Power Co., Inc. and the variations of the resin properties and the samples at the end of column were analyzed. The column test showed that the cross-linking degree of the new gel resins used was lower than those of porous ones. The new resins captured larger amounts of Matrix-Diffused Crud than the conventional cation resins before regeneration but not after that. Whereas the surface adsorbed crud was less captured by the new resins than conventional anion resins. However, there were little differences among these resins in respects of rinsing characteristics, sphericity, water quality, break through capacity, etc. At the condensate polisher in the secondary system it was confirmed that new gel resins had almost the same performance as one of the conventional ones and could be applied to the actual plant. (M.N.)

  15. Fatigue Crack Growth Rate Behavior of Type 347 Stainless Steel in Simulated PWR Water Environment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The pressurizer surge line of a Korean standard nuclear power plane uses Nb stabilized type 347 stainless steel. The pressurizer surge line is the pipe connecting the pressurizer and the hot leg line, and the path controlling the pressure and temperature of the cooling system of the nuclear reactor, operated at 316 .deg. C and in a 150atm. The pressurizer surge line operated at high temperature and high pressure receives thermal stress by a temperature change and mechanical stress by a pressure change at the same time, and by being exposed to the high temperature and high pressure cooling water environment of a nuclear power plant, environmental fatigue by stress and corrosion is the main damage instrument. As the effect of environmental fatigue has been reported, through low cycle fatigue, fatigue life evaluations of austenite stainless steel have been conducted, but evaluations of fatigue crack growth rate to evaluate the soundness are very poor. In this study, evaluated characteristics of fatigue crack growth rate base on a change of dissolved oxygen in a PWR environment

  16. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

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

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

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

  18. Development of a parametric containment event tree model of a severe PWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  19. Treatment of defective fuel rods for interim storage

    Energy Technology Data Exchange (ETDEWEB)

    Muenchow, K.; Hummel, W. [AREVA NP GmbH, Erlangen (Germany)

    2013-07-01

    In this paper we look exclusively at the treatment of defective fuel rods for long-term dry interim storage at the nuclear power plant, in order to avoid off-site transports. AREVA has developed a technique that allows verifiably adequate drying of the defective fuel rods and reconstructs the barrier for retaining radioactive materials. This is done by individually encapsulating the defective fuel rods and achieving gas-tightness by seal welding. This guarantees the retention of radioactive materials during the storage period of at least 40 years in a transport and storage flask in an interim storage facility at site. (orig.)

  20. Gamma-ray spectroscopy on irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear], e-mail: laaterre@ipen.br

    2009-07-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)