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Sample records for burnup pressurized water

  1. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    International Nuclear Information System (INIS)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok

    2015-01-01

    The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235 U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233 U, 242 Pu, 150 Nd, and 133 Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code

  2. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Jayalal, M.L.; Ramachandran, Suja; Rathakrishnan, S.; Satya Murty, S.A.V.; Sai Baba, M.

    2015-01-01

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  3. Benefits of cycle stretchout in pressurized water reactor extended-burnup fuel cycles

    International Nuclear Information System (INIS)

    Matzie, R.A.; Leung, D.C.; Liu, Y.; Beekmann, R.W.

    1981-01-01

    Nuclear reactors are inherently capable of operating for a substantial period beyond their nominal end of cycle (EOC) as a result of negative moderator and fuel temperature coefficients and the decrease in xenon poisoning with lower core power levels. This inherent capability can be used to advantage to reduce annual uranium makeup requirements and cycle energy costs by the use of planned EOC stretchout. This paper discusses the fuel utilization efficiency and economics of both the five-batch, extended-burnup cycle and the three-batch, standard-burnup cycle, which can be improved by employing planned EOC (end of cycle) stretchout. 11 refs

  4. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  5. Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew; Kalinich, Donald A.; Powers, Dana A.

    2016-12-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  6. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  7. Investigation of the burn-up behavior of boron poison rods, placed in a fuel assembly of a pressurized water reactor

    International Nuclear Information System (INIS)

    Arnold, C.; Lutz, D.C.

    1979-09-01

    The excess reactivity of a pressurized water reactor is compensated by boron, disolved in the moderator. In addition during the first cycle boron poison rods are placed in fuel assemblies without control rods. The burn-up behavior of a poison rod in a Biblis B fuel assembly is analysed in the present paper. Multigroup spectrum calculations were performed. The influence of critical boron concentration depending from burn-up, the changes of fuel concentration and the concentration of burnable poison were taken into consideration. Furthermore the built-up of rapidly saturating fisson products 135 Xe and 149 Sm was considered. The interaction of these effects are discussed. Spatial influences are emphasized most. Finally two group cross sections were calculated. The results are compared with calculations for a fuel assembly of the same type without burnable poison rods. (orig.) [de

  8. Automatic determination of pressurized water reactor core loading patterns which maximize end-of-cycle reactivity within power peaking and burnup constraints

    International Nuclear Information System (INIS)

    Hobson, G.H.

    1985-01-01

    An automated procedure for determining the optimal core loading pattern for a pressurized water reactor which maximizes end-of-cycle k/sub eff/ while satisfying constraints on power peaking and discharge burnup has been developed. The optimization algorithm combines a two energy group, two-dimensional coarse-mesh finite difference diffusion theory neutronics model to simulate core conditions, a perturbation theory approach to determine reactivity, flux, power and burnup changes as a function of assembly shuffling, and Monte Carlo integer programming to select the optimal loading pattern solution. The core examined was a typical Cycle 2 reload with no burnable poisons. Results indicate that the core loading pattern that maximizes end-of-cycle k/sub eff/ results in a 5.4% decrease in fuel cycle costs compared with the core loading pattern that minimizes the maximum relative radial power peak

  9. Automatic determination of pressurized water reactor core loading patterns that maximize beginning-of-cycle reactivity within power-peaking and burnup constraints

    International Nuclear Information System (INIS)

    Hobson, G.H.; Turinsky, P.J.

    1986-01-01

    Computational capability has been developed to automatically determine a good estimate of the core loading pattern, which minimizes fuel cycle costs for a pressurized water reactor (PWR). Equating fuel cycle cost minimization with core reactivity maximization, the objective is to determine the loading pattern that maximizes core reactivity while satisfying power peaking, discharge burnup, and other constraints. The method utilizes a two-dimensional, coarse-mesh, finite difference scheme to evaluate core reactivity and fluxes for an initial reference loading pattern. First-order perturbation theory is applied to determine the effects of assembly shuffling on reactivity, power distribution, end-of-cycle burnup. Monte Carlo integer programming is then used to determine a near-optimal loading pattern within a range of loading patterns near the reference pattern. The process then repeats with the new loading pattern as the reference loading pattern and terminates when no better loading pattern can be determined. The process was applied with both reactivity maximization and radial power-peaking minimization as objectives. Results on a typical large PWR indicate that the cost of obtaining an 8% improvement in radial power-peaking margin is ≅2% in fuel cycle costs, for the reload core loaded without burnable poisons that was studied

  10. Fast reactor cycle calculation routine using a 3D-simulator and investigation of new burnup stategies for pressurized water reactors

    International Nuclear Information System (INIS)

    Li Yulun.

    1987-03-01

    Three-dimensional calculations of the longtime behaviour of PWR can be done in short computing times with satisfactory accuracy for power and burn-up distributions. This has been proved by comparison with operational data of Biblis-B. Various possibilities are investigated to increase the discharge burn-up and to improve the utilization of uranium. In view of the increase of discharge burn-up due to enhanced cycle number (decreased batch size) and decreased neutron leakage these new strategies are intensively studied in the conventional fuel management scheme (Out-in) and in the low leakage fuel management scheme (In-Out). By a conventional fuel management scheme with four cycle operation and a low leakage fuel management scheme with three cycle operation an attractive increase of discharge burn-up to about 40% can be achieved by an increase in the reload enrichment to 4%. (orig.) [de

  11. Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Okubo, Tsutomu

    2011-01-01

    Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

  12. Reactivity and isotopic composition of spent PWR [pressurized-water-reactor] fuel as a function of initial enrichment, burnup, and cooling time

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  13. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF 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 [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  14. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

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

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF 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 [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  15. Long-Term Dry Storage of High Burn-Up Spent Pressurized Water Reactor (PWR) Fuel in TAD (Transportation, Aging, and Disposal) Containers

    International Nuclear Information System (INIS)

    Hwang, Yong Soo

    2008-12-01

    A TAD canister, in conjunction with specially-designed over-packs can accomplish the functions of transportation, aging, and disposal (TAD) in the management of spent nuclear fuel (SNF). Industrial dry cask systems currently available for SNF are licensed for storage-only or for dual-purpose (i.e., storage and transportation). By extending the function to include the indefinite storage and perhaps, eventual geologic disposal, the TAD canister would have to be designed to enhance, among others, corrosion resistance, thermal stability, and criticality-safety control. This investigative paper introduces the use of these advanced iron-based, corrosion-resistant materials for SNF transportation, aging, and disposal.The objective of this investigative project is to explore the interest that KAERI would research and develop its specific SAM coating materials for the TAD canisters to satisfy the requirements of corrosion-resistance, thermal stability, and criticality-controls for long-term dry storage of high burn-up spent PWR fuel

  16. Conservative axial burnup distributions for actinide-only burnup credit

    International Nuclear Information System (INIS)

    Kang, C.; Lancaster, D.

    1997-11-01

    Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model each burnup mesh with the corresponding isotopic compositions. Although this approach represents a significant increase in modeling efforts compared to the uniform average burnup approach, it can adequately determine the reactivity effect of the axial burnup distribution. A major consideration is what axial burnup distributions are appropriate for use in light of many possible distributions depending on core operating conditions and histories. This paper summarizes criticality analyses performed to determine conservative axial burnup distributions. The conservative axial burnup distributions presented in this paper are included in the Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages, Revision 1 submitted in May 1997 by the US Department of Energy (DOE) to the US Nuclear Regulatory Commission (NRC). When approved by NRC, the conservative axial burnup distributions may be used to model PWR spent nuclear fuel for the purpose of gaining actinide only burnup credit

  17. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The purpose of this meeting was to review the ''state-of-the-art'' in the area of Fuel Performance at High Burnup for Water Reactors. Previous IAEA meetings on this topic were held in Mol in 1981 and 1984 and on related topics in Stockholm and Lyon in 1987. Fifty-five participants from 16 countries and two international organizations attended the meeting and 28 papers were presented and discussed. The papers were presented in five sub-sessions and during the meeting, working groups composed of the session chairmen and paper authors prepared the summary of each session with conclusions and recommendations for future work. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  18. 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)

  19. Phenomena and parameters important to burnup credit

    International Nuclear Information System (INIS)

    Parks, C.V.; Dehart, M.D.; Wagner, J.C.

    2001-01-01

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water- reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the United States and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given. (author)

  20. High burnup fuel onset conditions in dry storage. Prediction of EOL rod internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L.E.

    2015-07-01

    During dry storage, cladding resistance to failure can be affected by several degrading mechanisms like creep or hydrides radial reorientation. The driving force of these effects is the stress at which the cladding is submitted. The maximum stress in the cladding is determined by the end-of-reactor-life (EOL) rod internal pressure, PEOL, at the maximum temperature attained during dry storage. Thus, PEOL sets the initial conditions of storage for potential time-dependent changes in the cladding. Based on FRAPCON-3.5 calculations, the aim of this work is to analyse the PEOL of a PWR fuel rod irradiated to burnups greater than 60 GWd/tU, where limited information is available. In order to be conservative, demanding irradiation histories have been used with a peak linear power of 44 kW/m. FRAPCON-3.5 results show an increasing exponential trend of PEOL with burnup, from which a simple correlation has been derived. The comparison with experimental data found in the literature confirms the enveloping nature of the predicted curve. Based on that, a conservative prediction of cladding stress in dry storage has been obtained. The comparison with a critical stress threshold related to hydrides embrittlement seems to point out that this issue should not be a concern at burnups below 65 GWd/tU. (Author)

  1. The burnup dependence of light water reactor spent fuel oxidation

    International Nuclear Information System (INIS)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO 2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO 2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

  2. International Atomic Energy Agency (IAEA) Activity on Technical Influence of High Burnup UOX and MOX Water Reactor Fuel on Spent Fuel Management

    International Nuclear Information System (INIS)

    Lovasic, Z.; Einziger, R.

    2009-01-01

    This paper briefly reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on spent fuel management. These data will provide information on the impacts, regarding spent fuel management, for those countries operating light-water reactors (LWR)s and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. These properties are reviewed. The effects of these property changes on wet and dry storage, transportation, reprocessing, re-fabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. Data gaps were also identified during this investigation. The pros and cons of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed.. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, re-fabrication, and

  3. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

  4. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  5. Issues for effective implementation of burnup credit

    International Nuclear Information System (INIS)

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

    2001-01-01

    In the United States, burnup credit has been used in the criticality safety evaluation for storage pools at pressurized water reactors (PWRs) and considerable work has been performed to lay the foundation for use of burnup credit in dry storage and transport cask applications and permanent disposal applications. Many of the technical issues related to the basic physics phenomena and parameters of importance are similar in each of these applications. However, the nuclear fuel cycle in the United States has never been fully integrated and the implementation of burnup credit to each of these applications is dependent somewhat on the specific safety bases developed over the history of each operational area. This paper will briefly review the implementation status of burnup credit for each application area and explore some of the remaining issues associated with effective implementation of burnup credit. (author)

  6. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

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

  7. Determination of Fission Gas Inclusion Pressures in High Burnup Nuclear Fuel using Laser Ablation ICP-MS combined with SEM/EPMA and Optical Microscopy

    International Nuclear Information System (INIS)

    Horvath, Matthias I.; Guenther-Leopold, Ines; Kivel, Niko; Restani, Renato; Guillong, Marcel; Izmer, Andrei; Hellwig, Christian; Guenther, Detlef

    2008-01-01

    In approximately 20% of all fissions at least one of the fission products is gaseous. These are mainly xenon and krypton isotopes contributing up to 90% by the xenon isotopes. Upon reaching a burn-up of 60 - 75 GWd/tHM a so called High Burnup Structure (HBS) is formed in the cooler rim of the fuel. In this region a depletion of the noble fission gases (FG) in the matrix and an enrichment of FG in μm-sized pores can be observed. Recent calculations show that in these pores the pressure at room temperature can be as large as 30 MPa. The knowledge of the FG pressure in pores is important to understand the high burn-up fuel behavior under accident conditions (i.e. RIA or LOCA). With analytical methods routinely used for the characterization of solid samples, i.e. Electron Probe Micro Analysis (EPMA), Secondary Ion Mass Spectrometry (SIMS), the quantification of gaseous inclusions is very difficult to almost impossible. The combination of a laser ablation system (LA) with an inductively coupled plasma mass spectrometer (ICP-MS) offers a powerful tool for quantification of the gaseous pore inventory. This method offers the advantages of high spatial resolution with laser spot sizes down to 10 μm and low detection limits. By coupling with scanning electron microscopy (SEM) for the pore size distribution, EPMA for the FG inventory in the fuel matrix and optical microscopy for the LA-crater sizes, the pressures in the pores and porosity was calculated. As a first application of this calibration technique for gases, measurements were performed on pressurized water reactor (PWR) fuel with a rod average of 105 GWd/tHM to determine the local FG pressure distribution. (authors)

  8. Investigation of very high burnup UO{sub 2} fuels in Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cappia, Fabiola

    2017-03-27

    Historically, the average discharge burnup of Light Water Reactor (LWR) fuel has increased almost continuously. On one side, increase in the average discharge burnup is attractive because it contributes to decrease part of the fuel cycle costs. On the other side, it raises the practical problem of predicting the performance, longevity and properties of reactor fuel elements upon accumulation of irradiation damage and fission products both during in-reactor operation and after discharge. Performance of the fuel and structural components of the core is one of the critical areas on which the economic viability and public acceptance of nuclear energy production hinges. Along the pellet radius, the fuel matrix is subjected to extremely heterogeneous alteration and damage, as a result of temperature and burnup gradients. In particular, in the peripheral region of LWR UO{sub 2} fuel pellets, when the local burnup exceeds 50-70 GWd/tHM, a microstructural transformation starts to take place, as a consequence of enhanced accumulation of radiation damage, fission products and limited thermal recovery. The newly formed structure is commonly named High Burnup Structure (HBS). The HBS is characterised by three main features: (a) formation of submicrometric grains from the original grains, (b) depletion of fission gas from the fuel matrix, (c) steep increase in the porosity, which retains most of the gas depleted from the fuel matrix. The last two aspects rose significant attention because of the important impact of the fission gas behaviour on integral fuel performance. The porosity increase controls the gas-driven swelling, worsening the cladding loading once the fuel-cladding gap is closed. Another concern is that the large retention of fission gas within the HBS could lead to significant release at high burnups through the degradation of thermal conductivity or contribute to fuel pulverisation during accidental conditions. Need of more experimental investigations about the

  9. Spent fuel dissolution rates as a function of burnup and water chemistry

    International Nuclear Information System (INIS)

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of 129 I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and ∼ 65 MWd/kgM. (2) Oxidation of spent fuel up to the U 4 O 9+x stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of 129 I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and 129 I gap inventory for US LWR fuels

  10. Burnup verification tests with the FORK measurement system-implementation for burnup credit

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

    Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. It was designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program and is well suited to verify burnup and cooling time records at commercial Pressurized Water Reactor (PWR) sites. This report deals with the application of the FORK system to burnup credit operations

  11. Neutron physics of a high converting advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Berger, H.D.

    1985-01-01

    The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de

  12. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-09-01

    General Atomics (GA) has designed two legal weight truck (LWT) casks, the GA-4 and GA-9, to carry four pressurized-water-reactor (PWR) and nine boiling-water-reactor (BWR) fuel assemblies, respectively. GA plans to submit applications for certification to the US Nuclear Regulatory Commission (NRC) for the two casks in mid-1993. GA will include burnup credit analysis in the Safety Analysis Report for Packaging (SARP) for the GA-4 Cask. By including burnup credit in the criticality safety analysis for PWR fuels with initial enrichments above 3% U-235, public and occupation risks are reduced and cost savings are realized. The GA approach to burnup credit analysis incorporates the information produced in the US Department of Energy Burnup Credit Program. This paper describes the application of burnup credit to the criticality control design of the GA-4 Cask

  13. High Burnup Effects Program

    International Nuclear Information System (INIS)

    Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D.

    1990-04-01

    This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs

  14. Nuclear Energy Research Initiative. Development of a Stabilized Light Water Reactor Fuel Matrix for Extended Burnup

    International Nuclear Information System (INIS)

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-01-01

    The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble dopants in the UO 2 matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus

  15. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  16. Burnup credit in a dry storage module

    International Nuclear Information System (INIS)

    Thornton, J.R.

    1989-01-01

    Comparison of spent fuel storage expansion options available to Oconee Nuclear Station revealed that dry storage could be economically competitive with transshipment and rod consolidation. Economic competitiveness, however, mandated large unit capacity while existing cask handling facilities at Oconee severely limited size and weight. The dry storage concept determined to best satisfy these conflicting criteria is a 24 pressurized water reactor (PWR) fuel assembly capacity NUTECH Horizontal Modular Storage (NUHOMS) system. The Oconee version of the NUHOMS system takes advantage of burnup credit in demonstrating criticality safety. The burnup credit criticality analysis was performed by Duke Power Company's Design Engineering Department. This paper was prepared to summarize the criticality control design features employed in the Oconee NUHOMS-24P DSC basket and to describe the incentives for pursuing a burnup credit design. Principal criticality design parameters, criteria, and analysis methodology are also presented

  17. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  18. Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC)

    International Nuclear Information System (INIS)

    2011-10-01

    This report presents the results of the Coordinated Research Project (CRP) on Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plants (FUWAC, 2006-2009). It provides an overview of the results of the investigations into the current state of water chemistry practice and concerns in the primary circuit of water cooled power reactors including: corrosion of primary circuit materials; deposit composition and thickness on the fuel; crud induced power shift; fuel oxide growth and thickness; radioactivity buildup in the reactor coolant system (RCS). The FUWAC CRP is a follow-up to the DAWAC CRP (Data Processing Technologies and Diagnostics for Water Chemistry and Corrosion Control in Nuclear Power Plants 2001-2005). The DAWAC project improved the data processing technologies and diagnostics for water chemistry and corrosion control in nuclear power plants (NPPs). With the improved methods for controlling and monitoring water chemistry now available, it was felt that a review of the principles of water chemistry management should be undertaken in the light of new materials, more onerous operating conditions, emergent issues such as CIPS, also known as axial offset anomaly (AOA) and the ageing of operating power plant. In the framework of this CRP, water chemistry specialists from 16 nuclear utilities and research organizations, representing 15 countries, exchanged experimental and operational data, models and insights into water chemistry management. The CD-ROM attached to this IAEA-TECDOC includes the report itself, detailed progress reports of three Research Coordination Meetings (RCMs) (Annexes I-III) and the reports and presentations made during the project by the participants.

  19. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  20. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    International Nuclear Information System (INIS)

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  1. Properties of the high burnup structure in nuclear light water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wiss, Thierry; Rondinella, Vincenzo V.; Konings, Rudy J.M. [European Commission, Joint Research Centre, Karlsruhe (Germany). Directorate Nuclear Safety and Security; and others

    2017-07-01

    The formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2-3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

  2. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  3. 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.

  4. Studies on validation possibilities for computational codes for criticality and burnup calculations of boiling water reactor fuel; Untersuchungen zu Validierungsmoeglichkeiten von Rechencodes fuer Kritikalitaets- und Abbrandrechnungen von Siedewasserreaktor-Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthais; Hannstein, Volker; Kilger, Robert; Sommer, Fabian; Stuke, Maik

    2017-06-15

    The Application of the method of Burn-up Credit on Boiling Water Reactor fuel is much more complex than in the case of Pressurized Water Reactors due to the increased heterogeneity and complexity of the fuel assemblies. Strongly varying enrichments, complex fuel assembly geometries, partial length fuel rods, and strong axial variations of the moderator density make the verification of conservative irradiation conditions difficult. In this Report, it was investigated whether it is possible to take into account the burn-up in criticality analyses for systems with irradiated Boiling Water Reactor fuel on the basis of freely available experimental data and by additionally applying stochastic methods. In order to achieve this goal, existing methods for stochastic analysis were adapted and further developed in order to being applicable to the specific conditions needed in Boiling Water Reactor analysis. The aim was to gain first insight whether a workable scheme for using burn-up credit in Boiling Water Reactor applications can be derived. Due to the fact that the different relevant quantities, like e.g. moderator density and the axial power profile, are strongly correlated, the GRS-tool SUnCISTT for Monte-Carlo uncertainty quantification was used in the analysis. This tool was coupled to a simplified, consistent model for the irradiation conditions. In contrast to conventional methods, this approach allows to simultaneously analyze all involved effects.

  5. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  6. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately.

  7. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately

  8. Development of an extended-burnup Mark B design. Second semiannual progress report, January-June 1979

    International Nuclear Information System (INIS)

    1979-11-01

    The immediate goal of the DOE/AP and L/B and W project is to extend the burnup of light water reactor fuel assemblies beyond present limits to 50,000 MWd/mtU batch average burnup. Fuel management plans and fuel designs are being directed to attain the increased burnup limits. Lead-test assemblies of extended-burnup designs will be manufactured, irradiated in a commercial pressurized water reactor, and examined to support extended-burnup fuel cycles. This report, covering the period from January through June 1979, is the second semiannual progress report for the program. Efforts have included analyses of extended-burnup fuel cycles, developed of both annular fuel pellet and segmented rod designs, and design of a nondestructive post-irradiation examination system

  9. Improvement in fuel utilization in pressurized heavy water reactors due to increased heavy water purity

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1991-01-01

    This paper reports that in a pressurized heavy water reactor (PHWR), the reactivity of the reactor and, consequently, the discharge burnup of the fuel depend on the isotopic purity of the heavy water used in the reactor. The optimal purity of heavy water used in PHWRs, in turn, depends on the cost of fabricated uranium fuel and on the incremental cost incurred in improving the heavy water purity. The physics and economics aspects of the desirability of increasing the heavy water purity in PHWRs in India were first examined in 1978. With the cost data available at that time, it was found that improving the heavy water purity from 99.80% to 99.95% was economically attractive. The same problem is reinvestigated with current cost data. Even now, there is sufficient incentive to improve the isotopic purity of heavy water used in PHWRs. Admittedly, the economic advantage that can be derived depends on the cost of the fabricated fuel. Nevertheless, irrespective of the economics, there is also a fairly substantial saving in natural uranium. That the increase in the heavy water purity is to be maintained only in the low-pressure moderator system, and not in the high-pressure coolant system, makes the option of achieving higher fuel burnup with higher heavy water purity feasible

  10. Findings of an international study on burnup credit

    International Nuclear Information System (INIS)

    Brady, M.C.; Takano, M.; Okuno, H.; DeHart, M.D.; Nouri, A.

    1996-01-01

    Findings from a four year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses are presented in this paper. Approximately 20 participants from 11 countries have provided results for most problems. Four detailed benchmark problems for Pressurized Water Reactor (PWR) fuel have been completed and are summarized in this paper. Preliminary results from current work addressing burnup credit for Boiling Water Reactor (BWR) fuel will also be discussed as well as planned activities for additional benchmarks including Mixed-Oxide (MOX) fuels, subcritical benchmarks, international databases, and other activities

  11. Measurement of fission gas release, internal pressure and cladding creep rate in the fuel pins of PHWR bundle of normal discharge burnup

    Energy Technology Data Exchange (ETDEWEB)

    Viswanathan, U.K. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Sah, D.N., E-mail: dnsah@barc.gov.i [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Rath, B.N.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2009-08-01

    Fuel pins of a Pressurised Heavy Water Reactor (PHWR) fuel bundle discharged from Narora Atomic Power Station unit no. 1 after attaining a fuel burnup of 7528 MWd/tU have been subjected to two types of studies, namely (i) puncture test to estimate extent of fission gas release and internal pressure in the fuel pin and (ii) localized heating of the irradiated fuel pin to measure the creep rate of the cladding in temperature range 800 deg. C - 900 deg. C. The fission gas release in the fuel pins from the outer ring of the bundle was found to be about 8%. However, only marginal release was found in fuel pins from the middle ring and the central fuel pin. The internal gas pressure in the outer fuel pin was measured to be 0.55 +- 0.05 MPa at room temperature. In-cell isothermal heating of a small portion of the outer fuel pins was carried out at 800 deg. C, 850 deg. C and 900 deg. C for 10 min and the increase in diameter of the fuel pin was measured after heat treatment. Creep rates of the cladding obtained from the measurement of the diameter change of the cladding due to heating at 800 deg. C, 850 deg. C and 900 deg. C were found respectively to be 2.4 x 10{sup -5} s{sup -1}, 24.6 x 10{sup -5} s{sup -1} and 45.6 x 10{sup -5} s{sup -1}.

  12. UABUC - Single energy point model burnup computer code for water reactors

    International Nuclear Information System (INIS)

    El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

    1981-01-01

    UABUC is a single energy point reactor burnup computer program in FORTRAN language. The program calculates the change in the isotopic composition of the uranium fuel as a function of irradiation time with all its associated quantities such as the average point flux, the conversion ratio, macroscopic fuel cross sections, and the point reactivity profile. A step-wise time analytical solution was developed for the nonlinear first order burnup differential equations. The ''Westcott'' convention of the effective cross sections was used except for plutonium-240 and uranium-238. For plutonium-240, an effective microscopic cross section was derived from the direct physical arguments taking into account the selfshielding effect of plutonium-240 as well as the 1 ev. resonance absorption. For uranium-238, an effective cross section, reflecting the effect of fast fission and resonance absorption was used. The fission products were treated in the three groups with 50, 300, and 800 barns. The yields in the groups were treated as functions of the type of fissionable nuclides, the effective neutron temperature, and the epithermal index. Xenon-135 and Samarium-149 were treated separately as functions of irradiation time. (author)

  13. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    Lott, Randy G.

    2003-01-01

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

  14. High Pressure Industrial Water Facility

    Science.gov (United States)

    1992-01-01

    In conjunction with Space Shuttle Main Engine testing at Stennis, the Nordberg Water Pumps at the High Pressure Industrial Water Facility provide water for cooling the flame deflectors at the test stands during test firings.

  15. Optimum Discharge Burnup and Cycle Length for PWRs

    International Nuclear Information System (INIS)

    Secker, Jeffrey R.; Johansen, Baard J.; Stucker, David L.; Ozer, Odelli; Ivanov, Kostadin; Yilmaz, Serkan; Young, E.H.

    2005-01-01

    This paper discusses the results of a pressurized water reactor fuel management study determining the optimum discharge burnup and cycle length. A comprehensive study was performed considering 12-, 18-, and 24-month fuel cycles over a wide range of discharge burnups. A neutronic study was performed followed by an economic evaluation. The first phase of the study limited the fuel enrichments used in the study to 235 U consistent with constraints today. The second phase extended the range of discharge burnups for 18-month cycles by using fuel enriched in excess of 5 wt%. The neutronic study used state-of-the-art reactor physics methods to accurately determine enrichment requirements. Energy requirements were consistent with today's high capacity factors (>98%) and short (15-day) refueling outages. The economic evaluation method considers various component costs including uranium, conversion, enrichment, fabrication and spent-fuel storage costs as well as the effect of discounting of the revenue stream. The resulting fuel cycle costs as a function of cycle length and discharge burnup are presented and discussed. Fuel costs decline with increasing discharge burnup for all cycle lengths up to the maximum discharge burnup considered. The choice of optimum cycle length depends on assumptions for outage costs

  16. Cell verification of parallel burnup calculation program MCBMPI based on MPI

    International Nuclear Information System (INIS)

    Yang Wankui; Liu Yaoguang; Ma Jimin; Wang Guanbo; Yang Xin; She Ding

    2014-01-01

    The parallel burnup calculation program MCBMPI was developed. The program was modularized. The parallel MCNP5 program MCNP5MPI was employed as neutron transport calculation module. And a composite of three solution methods was used to solve burnup equation, i.e. matrix exponential technique, TTA analytical solution, and Gauss Seidel iteration. MPI parallel zone decomposition strategy was concluded in the program. The program system only consists of MCNP5MPI and burnup subroutine. The latter achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, and data exchanging with MCNP5MPI. Also, the program was verified with the pressurized water reactor (PWR) cell burnup benchmark. The results show that it,s capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

  17. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    Ahlf, J.

    1983-01-01

    The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.) [de

  18. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  19. Modelling of some high burnup phenomena in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, K; Lindstroem, F; Massih, A R [ABB Atom AB, Vaesteraas (Sweden)

    1997-08-01

    In this paper the results of some modelling efforts carried out by ABB Atom to describe certain light water reactor fuel high burnup effects are presented. In particular the degradation of fuel thermal conductivity with burnup and its impact on fuel temperature is briefly discussed. The formation of a porous rim and its effect on a thermal fission gas release has been modelled and the model has been used to predict the release of pressurized water reactor fuel rods that were operated at low power densities. Furthermore, a mathematical model which combines the diffusion and re-solution controlled thermal release with grain boundary movement has been briefly described. The model is used to compare release with diffusion only and release caused by diffusion and grain boundary sweeping (due to grain growth). Finally, analytical expressions are obtained for the calculation of fuel stoichiometry as a function of burnup. (author). 20 refs, 10 figs, 1 tab.

  20. High-Pressure Liquid Chromatography of Irradiated Nuclear Fue - Separation of Neodymium for Burn-up Determination

    DEFF Research Database (Denmark)

    Larsen, N. R.

    1979-01-01

    Neodymium is separated from solutions of spent nuclear fuel by high-pressure liquid chromatography in methanol-nitric acid-water media using an anion-exchange column. Chromatograms obtained by monitoring at 280 nm, illustrate the difficulties especially with the fission product ruthenium in nuclear...

  1. Fully coupled modeling of burnup dependent light water reactor fuel performance using COMSOL Multiphysics

    International Nuclear Information System (INIS)

    Liu Rong; Zhou Wenzhong; Prudil, Andrew

    2015-01-01

    This paper presents the development of a light water reactor fuel performance code, which considers almost all the related physical models, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the equations are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. Comparisons are made for the simulation results between COMSOL and another simulation tool of BISON. The comparisons show the capability of our simulation tool to predict light water UO 2 fuel performances. In our modeling and simulation work, the performance of enhanced thermal conductivity UO 2 -BeO fuel and newly-adopted corrosion resistant SiC cladding material was also studied. UO 2 -BeO high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor would be improved. However, for SiC cladding, although due to its high thermal expansion, the gap closure time is delayed, irradiation induced point defects and defect-clusters in the SiC crystal will dramatically decrease SiC thermal conductivity, and cause significant increase in the fuel temperature. (author)

  2. Approach for implementing burnup credit in high-capacity truck casks

    International Nuclear Information System (INIS)

    Boshoven, J.; Hopf, J.; Su, S.

    1991-01-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks in 1992. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for the higher enrichments. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA presents here an approach to burnup credit analysis to be included in the Safety Analysis Report for Packaging (SARP). 6 refs., 2 figs., 5 tabs

  3. Water Pressure. Water in Africa.

    Science.gov (United States)

    Garrett, Carly Sporer

    The Water in Africa Project was realized over a 2-year period by a team of Peace Corps volunteers. As part of an expanded, detailed design, resources were collected from over 90 volunteers serving in African countries, photos and stories were prepared, and standards-based learning units were created for K-12 students. This unit, "Water…

  4. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

    2008-01-01

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  5. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  6. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  7. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    International Nuclear Information System (INIS)

    Zwicky, Hans-Urs; Low, Jeanett; Ekeroth, Ella

    2011-03-01

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 μm. This structure forms in UO 2 fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238 U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  8. Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

    International Nuclear Information System (INIS)

    Wagner, John C.; Parks, Cecil V.; Mueller, Don; Gauld, Ian C.

    2010-01-01

    Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

  9. The MOX Fuel Behaviour Test IFA-597.4: Temperature And Pressure Data To A Burn-Up Of 5.4 MWd/kg MOX

    International Nuclear Information System (INIS)

    McGrath, M. A.; Teshima, H.

    1998-02-01

    Characterising the behaviour of MOX fuel is becoming increasingly important as many commercial reactors are or will be operating with this type of fuel. With this as a driving force, a new joint programme experiment, IFA-597.4, has been loaded into the reactor at Halden for the purpose of establishing the fission gas release behaviour of MOX fuel. Both annular and solid pellet fuel is being utilised and the irradiation is being conducted such that the fuel is initially operated below the onset of fission gas release. The fuel will later be subjected to small power up ratings which will be held for short periods of time. These are designed to bring the fuel to just above the temperature threshold for fission gas release thus allowing the FGR behaviour of both solid and annular MOX fuel to be established. The rig contains two fuel rods of active length 220 mm and diameter 8.05 mm. Both fuel rods contain MOX fuel with an initial Pu-fissile content of 6.07% and both are instrumented with a fuel centre thermocouple and a pressure transducer. The test is being performed under HBWR conditions and at the time of the reactor shutdown at the end of 1997 a mean burn-up of 5.4 MWd/kg MOX had been achieved with the rods at an average rating of 30 kW/m. The rod pressure data show that no fission gas had been released up to the shutdown. The fuel centre temperatures of both rods exhibit an initial increase concurrent with a fall in the monitored rod internal pressures as a result of fuel densification. It was estimated that about 1-1.4% fuel densification by volume had occurred in the two rods by a burn-up of about 3 MWd/kg MOX. (author)

  10. Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

    International Nuclear Information System (INIS)

    Penndorf, K.

    1976-04-01

    Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time

  11. High-pressure water facility

    Science.gov (United States)

    2006-01-01

    NASA Test Operations Group employees, from left, Todd Pearson, Tim Delcuze and Rodney Wilkinson maintain a water pump in Stennis Space Center's high-pressure water facility. The three were part of a group of employees who rode out Hurricane Katrina at the facility and helped protect NASA's rocket engine test complex.

  12. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  13. High pressure experimental water loop

    International Nuclear Information System (INIS)

    Grenon, M.

    1958-01-01

    A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm 2 and flow rates up to 5 m 3 /h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm 2 , and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author) [fr

  14. The MOX fuel behaviour test IFA-597.4/.5. Temperature and pressure data to a burn-up of 15 MWd/kg MOX

    International Nuclear Information System (INIS)

    Takano, K.

    1999-04-01

    The behaviour of MOX fuel should be investigated in detail for more effective use in the future, especially concerning its thermal performance and fission gas release. IFA-597.4 and IFA-597.5, containing two MOX fuel rods each with a fuel centre thermocouple and a pressure transducer, have been irradiated in the Halden Reactor to study the temperature threshold of fission gas release for MOX fuel and to explore potential differences in the thermal and fission gas release behaviour between solid and hollow pellets. The two rods of MOX fuel with an initial Pu-fissile content of 6.07 percent have solid pellets and hollow pellets respectively, and with an active length of about 220 mm. The diameter of the pellets is 8.05 mm with 180μm of diametral gap to the cladding. For the purpose of the test, power ramp operation, in which estimated peak temperature of the MOX pellets increases and decreases above and below the threshold for fission gas release in UO 2 fuel, is planned every 10 MWd/kgMOX of burn-up. The first ramp operation has been successfully performed at 10 MWd/kgMOX. When the estimated peak temperature of the fuel gets close to but below the threshold of UO 2 , fission gas release was observed at around 28 kW/m of power. Densification of the MOX pellets could be estimated to about 1.2 percent for the solid pellets and about 2,3 percent for the hollow pellets from normalised internal rod pressure. After 13.5 MWd/kgMOX the average assembly power has been operated low enough to observe swelling rate of MOX fuel pellets and behaviour after significant fission gas release. The burn-up had reached 15.5 MWd/kgMOX as of the end of 1998. The target burn-up of this MOX test is 60 MWd/kgMOX (author) (ml)

  15. Water cooled static pressure probe

    Science.gov (United States)

    Lagen, Nicholas T. (Inventor); Eves, John W. (Inventor); Reece, Garland D. (Inventor); Geissinger, Steve L. (Inventor)

    1991-01-01

    An improved static pressure probe containing a water cooling mechanism is disclosed. This probe has a hollow interior containing a central coolant tube and multiple individual pressure measurement tubes connected to holes placed on the exterior. Coolant from the central tube symmetrically immerses the interior of the probe, allowing it to sustain high temperature (in the region of 2500 F) supersonic jet flow indefinitely, while still recording accurate pressure data. The coolant exits the probe body by way of a reservoir attached to the aft of the probe. The pressure measurement tubes are joined to a single, larger manifold in the reservoir. This manifold is attached to a pressure transducer that records the average static pressure.

  16. Development and benchmark verification of a parallelized Monte Carlo burnup calculation program MCBMPI

    International Nuclear Information System (INIS)

    Yang Wankui; Liu Yaoguang; Ma Jimin; Yang Xin; Wang Guanbo

    2014-01-01

    MCBMPI, a parallelized burnup calculation program, was developed. The program is modularized. Neutron transport calculation module employs the parallelized MCNP5 program MCNP5MPI, and burnup calculation module employs ORIGEN2, with the MPI parallel zone decomposition strategy. The program system only consists of MCNP5MPI and an interface subroutine. The interface subroutine achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, data exchanging with MCNP5MPI. Also, the program was verified with the Pressurized Water Reactor (PWR) cell burnup benchmark, the results showed that it's capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

  17. Fuel analysis code FAIR and its high burnup modelling capabilities

    International Nuclear Information System (INIS)

    Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1995-01-01

    A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

  18. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

  19. Choosing the optimum burnup

    International Nuclear Information System (INIS)

    Geller, L.; Goldstein, L.; Franks, W.A.

    1986-01-01

    This paper reviews some of the considerations utilities must evaluate when going to higher discharge burnups. The advantages and disadvantages of higher discharge burnups are described, as well as a consistent approach for evaluating optimum discharge burnup and its comparison to current practice. When an analysis is performed over the life of the plant, the design of the terminal cycles has significant impact on the lifetime savings from higher burnups. Designs for high burnup cycles have a greater average inventory value in the core. As one goes to higher burnup, there is a greater likelihood of discarding a larger value in unused fuel unless the terminal cycles are designed carefully. This effect can be large enough in some cases to wipe out the lifetime cost savings relative to operating with a higher discharge burnup cycle

  20. Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC). Additional Information

    International Nuclear Information System (INIS)

    2011-10-01

    This report presents the results of the Coordinated Research Project (CRP) on Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plants (FUWAC, 2006-2009). It provides an overview of the results of the investigations into the current state of water chemistry practice and concerns in the primary circuit of water cooled power reactors including: corrosion of primary circuit materials; deposit composition and thickness on the fuel; crud induced power shift; fuel oxide growth and thickness; radioactivity buildup in the reactor coolant system (RCS). The FUWAC CRP is a follow-up to the DAWAC CRP (Data Processing Technologies and Diagnostics for Water Chemistry and Corrosion Control in Nuclear Power Plants 2001-2005). The DAWAC project improved the data processing technologies and diagnostics for water chemistry and corrosion control in nuclear power plants (NPPs). With the improved methods for controlling and monitoring water chemistry now available, it was felt that a review of the principles of water chemistry management should be undertaken in the light of new materials, more onerous operating conditions, emergent issues such as CIPS, also known as axial offset anomaly (AOA) and the ageing of operating power plant. In the framework of this CRP, water chemistry specialists from 16 nuclear utilities and research organizations, representing 15 countries, exchanged experimental and operational data, models and insights into water chemistry management. This CD-ROM attached to the printed IAEA-TECDOC includes the report itself, detailed progress reports of three Research Coordination Meetings (RCMs) (Annexes I-III) and the reports and presentations made during the project by the participants.

  1. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  2. Extended burnup with SEU fuel in Atucha-1 NPP

    International Nuclear Information System (INIS)

    Alvarez, L.; Casario, J.; Fink, J.; Perez, R.; Higa, M.

    2002-01-01

    Atucha-1 is a Pressurized Heavy Water Reactor originally fuelled with natural uranium. Fuel Assemblies consist of 36 fuel rods and the active length is 5300 mm. The total length of the fuel assembly is about 6 m. The average discharge burnup of natural UO 2 fuel is 5900 MWd/tU. After the deregulation of the Argentine electricity market there was an important incentive to reduce the impact of fuel cost on the cost of generation. To keep the competitiveness of the nuclear energy against another sources of electricity it was necessary to reduce the cost of the nuclear fuel. With this objective a program to introduce SEU (0.85 % 235 U) fuel in Atucha-1 was launched in 1993. As a result of this program the average SEU fuel discharge burnup increased to more than 11000 MWd/tU. The first SEU fuels were introduced in Atucha-1 in 1995 and, in the present stage of the program, 71% of core positions are loaded with this type of fuel. This paper describes key aspects of Atucha-1 fuel design and their relevance limiting the burnup extension and shows relevant data regarding the SEU in-reactor performance. At the present time 125 SEU Fuel Assemblies have been irradiated without failures associated with the extended burnup or unfavorable influences on the operation of the power station. (author)

  3. International studies on burnup credit criticality safety by an OECD/NEA working group

    International Nuclear Information System (INIS)

    Brady, M.C.; Okuno, H.; DeHart, M.D.; Nouri, A.; Sartori, E.

    1998-01-01

    The results and conclusions from a six-year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses is presented. Approximately 20 participants from 12 countries have provided results for most problems. Four detailed benchmark problems for pressurized-water-reactor fuel have been completed. Results from work being finalized, addressing burnup credit for boiling-water-reactor fuel, are discussed, as well as planned activities for additional benchmarks, including mixed-oxide fuels, and other activities

  4. Optimisation of water chemistry to ensure reliable water reactor fuel performance at high burnup and in ageing plant (FUWAC): an International Atomic Energy Agency coordinated research project

    Energy Technology Data Exchange (ETDEWEB)

    Killeen, J.C. [International Atomic Energy Agency, Vienna (Austria); Nordmann, F. [Advanced Nuclear Technology International Europe AB, Beauchamp (France); Schunk, J. [Paks NPP (Hungary); Vonkova, K. [Nuclear Research Inst., Rez (Czech Republic)

    2010-07-01

    The IAEA project 'Optimisation of Water Chemistry to ensure Reliable Water Reactor Fuel Performance at High Burnup and in Aging Plant' (FUWAC) was initiated with the objectives of monitoring, maintaining and optimising water chemistry regimes in primary circuits of water cooled power reactors, taking into account high burnup operation, mixed cores and plant aging, including following issues and remedies. This report provides some highlights of the work undertaken by the project participants. Clad oxidation studies have been undertaken and include operational data from the South Ukraine WWER where no corrosion problems have been seen on either Westinghouse ZIRLO™ or Russian alloy E110 fuel cladding. Work on the Russian alloy E110 showed that potassium in the coolant is preferable to lithium for mitigating fuel cladding oxidation. Studies on crud behaviour in PWR have shown a dependence on crud thickness and pHT. The nature and mechanisms for boron deposition in fuel cladding cruds have been investigated which is the root cause of crud induced power shifts (CIPS). Operational experience at French PWRs shows no difference in the CIPS behaviour between units with Alloy 600 or 690 steam generators, whilst Korean experience provides information on the Ni/Fe ratio on fuel cladding crud and the occurrence of CIPS. Coolant additions have been studied, for example in BWR units using zinc addition, crud is more tenacious. Zinc is also added to PWR units, mainly for dose rate control and in some cases for PWSCC mitigation of Alloy 600. At low levels there has been no clear evidence of any effect of zinc on CIPS, but there is a benefit on fuel oxidation. It is suggested that zinc addition should be considered where there is SG replacement or fuel core management modification. One possibility for the elimination of fuel crud is decontamination. Such an operation is time consuming, expensive, includes several risks of corrosion and induces a large quantity of

  5. Establishing a PWR burn-up library

    International Nuclear Information System (INIS)

    Lutz, D.C.

    1981-01-01

    Starting out from data file ENDF/B IV /1/, a cross-section library has been established for the calculation of operating conditions in pressurized water reactors of the type used in BIBLIS B. The library includes macroscopic, homogenized 2-group cross-sections for all types of fuel elements used in this reactor, including those equipped with boron glass rods. For their calculation the previous irradiation of the fuel has been taken into consideration by approximation. Information on fuel consumption from cell burn-up calculations has been stored in a separate data file. It was designed as a base for the determination of cross sections to be used in the calculation of the incident ''main-steam pipe fracture''. For this library the description of cross sections as a function of the moderator status chose the water densities at 300 0 C/155 bar, 190 0 C/140 bar and 100 0 C/100 bar as fixed values. The burn-up library has been tested by a three-dimensional calculation for the 1sup(st) cycle of the BIBLIS B-reactor using program QUABOX /2/. This showed variances with the anticipated course concerning critically, which can be explained almost quantitatively by known deficiencies of the ENDF/b-IV library. (orig.) [de

  6. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

    Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics

  7. The effect of fuel burnup and dispersed water intrusion on the criticality of spent high-level nuclear fuel in a geologic repository

    International Nuclear Information System (INIS)

    Culbreth, W.G.; Zielinski, P.R.

    1994-01-01

    Studies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies. In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of burnup that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.1 was used to calculate values of the effective multiplication factor, k eff . Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that k eff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container

  8. Parametric studies for the nuclear design of high-conversion pressurized water reactors

    International Nuclear Information System (INIS)

    Axmann, J.; Oldekop, W.

    1987-01-01

    Undermoderated high-conversion pressurized water reactors with steel canning tubes offer the possibility of high burnup together with a comparatively low consumption of fissionable material; however, they require a relatively large inventory of fissionable material. The effects of different fuel compositions upon the specific consumption of fissionable material are investigated for a fixed burnup and moderator-to-fuel volume ratios varying between 0.5 and 2.0. Moreover, the required inventory of fissionable material is determined and the influence on the costs of electric power generation is shown. Further investigations deal with the neutron-physical effects of decreasing fuel rod diameters and the influence of differing steel additives. It appears that the parasitic neutron absorption by alloying constituents depends on the moderation level in a non-uniform manner and that the contribution of the fissionable material to the electric power generation costs is rather independent of the moderator-to-fuel volume ratio. (orig.) [de

  9. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

  10. Simulation of low pressure water hammer

    Science.gov (United States)

    Himr, D.; Habán, V.

    2010-08-01

    Numerical solution of water hammer is presented in this paper. The contribution is focused on water hammer in the area of low pressure, which is completely different than high pressure case. Little volume of air and influence of the pipe are assumed in water, which cause sound speed change due to pressure alterations. Computation is compared with experimental measurement.

  11. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  12. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    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)

  13. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1985-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

  14. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1980-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

  15. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.; DeHart, M.D.

    1999-01-01

    Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel

  16. An evaluation of the deployment of AIROX-recycled fuel in pressurized water reactors

    International Nuclear Information System (INIS)

    Jahshan, S.N.; McGeehan, T.J.

    1994-01-01

    An analytical evaluation is made of the pressurized water reactor (PWR) in-core performance of recycled light water reactor fuel that has been Atomics International reduction oxidation (AIROX) reprocessed and reenriched with fissile materials. The neutronics performance is shown to lie within the neutronics performance of existing high-performance and high-burnup fuels. Three AIROX-recycled fuels are compared with a high-burnup virgin fuel and an equivalent mixed-oxide (MOX) fuel. The AIROX-recycled fuel neutronics performance lies consistently between the virgin and the MOX fuel for both the pin power peaking and the reactivity response characteristics in PWRs. Among the attractive features of AIROX-recycled fuel is that it can optimize fissile and fertile fuel use, minimize final fuel disposal impact on the environment, and provide energy in the process of denaturing weapons-grade fissile materials. The fuel material performance may be anticipated from high-burnup virgin fuel and from MOX fuel performance. Recommendations for lead rod testing and for optimization of the AIROX-processing and resintering techniques are made

  17. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-08-01

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA's activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ''FUMEX'', allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs

  18. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA`s activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ``FUMEX``, allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs.

  19. Neutronics performances study of silicon carbide as an inert matrix to achieve very high burn-up for light water reactor fuels

    International Nuclear Information System (INIS)

    Chabert, C.; Coulon-Picard, E.; Pelletier, M.

    2007-01-01

    In order to extend the actual limits of light water reactors, the Cea has put emphasis on the exploration of major fuel innovations that would allow us to increase the competitiveness, the safety and flexibility, while keeping the standard PWR environment. Different fuel concepts have been chosen and are actually studied to evaluate their advantages and drawbacks. The objectives of these new fuels are to increase the safety performances and to achieve a very high burn-up. One concept is a CERCER fuel with silicon carbide (SiC) as an inert matrix devoted to reduce the fuel temperature at nominal conditions. Besides the investigation of the neutronic performance, analyses on the thermomechanical performances, the fuel fabrication, the fuel reprocessing and economic aspects have been performed. This paper presents particularly neutronic results obtained for the CERCER fuel. The results show that a very high burn-up, a high safety performance and a better competitiveness cannot be achieved with this fuel concept. (authors)

  20. Regulatory status of burnup credit for dry storage and transport of spent nuclear fuel in the United States

    International Nuclear Information System (INIS)

    Carlson, D.E.

    2001-01-01

    During 1999, the Spent Fuel Project Office of the U.S. Nuclear Regulatory Commission (NRC) introduced technical guidance for allowing burnup credit in the criticality safety analysis of casks for transporting or storing spent fuel from pressurized water reactors. This paper presents the recommendations embodied by the current NRC guidance, discusses associated technical issues, and reviews information needs and industry priorities for expanding the scope and content of the guidance. Allowable analysis approaches for burnup credit must account for the fuel irradiation variables that affect spent fuel reactivity, including the axial and horizontal variation of burnup within fuel assemblies. Consistent with international transport regulations, the burnup of each fuel assembly must be verified by pre-loading measurements. The current guidance limits the credited burnup to no more than 40 GWd/MTU and the credited cooling time to five years, imposes a burnup offset for fuels with initial enrichments between 4 and 5 wt% 235U, does not include credit for fission products, and excludes burnup credit for damaged fuels and fuels that have used burnable absorbers. Burnup credit outside these limits may be considered when adequately supported by technical information beyond that reviewed to-date by the NRC staff. The guidance further recommends that residual subcritical margins from the neglect of fission products, and any other nuclides not credited in the licensing-basis analysis, be estimated for each cask design and compared against estimates of the maximum reactivity effects associated with remaining computational uncertainties and potentially nonconservative modeling assumptions. The NRC's Office of Nuclear Regulatory Research is conducting a research program to help develop the technical information needed for refining and expanding the evolving guidance. Cask vendors have announced plans to submit the first NRC license applications for burnup credit later this year

  1. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.

  2. Some implications of batch average burnup calculations on predicted spent fuel compositions

    International Nuclear Information System (INIS)

    Alexander, C.W.; Croff, A.G.

    1984-01-01

    The accuracy of using batch-averaged burnups to determine spent fuel characteristics (such as isotopic composition, activity, etc.) was examined for a typical pressurized-water reactor (PWR) fuel discharge batch by comparing characteristics computed by (a) performing a single depletion calculation using the average burnup of the spent fuel and (b) performing separate depletion calculations based on the relative amounts of spent fuel in each of twelve burnup ranges and summing the results. The computations were done using ORIGEN 2. Procedure (b) showed a significant shift toward a greater quantity of the heavier transuranics, which derive from multiple neutron captures, and a corresponding decrease in the amounts of lower transuranics. Those characteristics which derive primarily from fission products, such as total radioactivity and total thermal power, are essentially identical for the two procedures. Those characteristics that derive primarily from the heavier transuranics, such as spontaneous fission neutrons, are underestimated by procedure (a)

  3. On the rate determining step in fission gas release from high burn-up water reactor fuel during power transients

    International Nuclear Information System (INIS)

    Walker, C.T.; Mogensen, M.

    1987-01-01

    The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved. (orig.)

  4. Summary report on the international comparison of NEACRP burnup benchmark calculations for high conversion light water reactor lattices

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Ishiguro, Yukio; Takano, Hideki

    1988-10-01

    The results of the NEACRP HCLWR cell burnup benchmark calculations are summarized in this report. Fifteen organizations from eight countries participated in this benchmark and submitted twenty solutions. Large differences are still observed among the calculated values of void reactivities and conversion ratios. These differences are mainly caused from the discrepancies in the reaction rates of U-238, Pu-239 and fission products. The physics problems related to these results are briefly investigated in the report. In the specialists' meeting on this benchmark calculations held in April 1988, it was recommended to perform continuous energy Monte Carlo calculations in order to obtain reference solutions for design codes. The conclusions resulted from the specialists' meeting are also presented. (author)

  5. High temperature pressure water's blowdown into water. Experimental results

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Kasahara, Yoshiyuki; Iida, Hiromasa

    1994-01-01

    The purpose of the present experimental study is to clarify the phenomena in blowdown of high temperature and pressure water in pressure vessel into the containment water for evaluation of design of an advanced marine reactor(MRX). The water blown into the containment water flushed and formed steam jet plume. The steam jet condensed in the water, but some stream penetrated to gas phase of containment and contributed to increase of containment pressure. (author)

  6. Appropriate burnup measurements for transportation burnup credit

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.

    1997-01-01

    This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy requirements are established for dependent and independent systems. The requirements have been tested and are achievable, ensuring safe operation without extra cost. 6 refs

  7. High burnup models in computer code fair

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

    1997-08-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

  8. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

    1997-01-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

  9. Research on burnup physics

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1974-07-01

    the validity of the equivalent material buckling definition, measurements were performed on experimental heavy water zero power reactor RB in Vinca, for three lattice pitches and lattice mixed of two kinds of fuel elements. The application of quasilinear programming for solving the control rod distribution problem to achieve maximum average fuel burnup resulted in a flexible optimization procedure (author)

  10. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-01-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  11. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-01-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  12. Managing water pressure for water savings in developing countries

    African Journals Online (AJOL)

    2014-03-03

    Mar 3, 2014 ... effort into providing customers with a reliable level of service, often via poor water ... budgets. There are many factors contributing to water losses in water .... given relationship does not reflect the impact of pressure on.

  13. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

    International Nuclear Information System (INIS)

    Wagner, J.C.; DeHart, M.D.

    2000-01-01

    This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ''end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified

  14. Burnup credit in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    2001-01-01

    The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analytical activities in the country are described, as well as international collaborations in this field. Perspectives for future research and development of new applications are finally resumed. (author)

  15. OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN

    International Nuclear Information System (INIS)

    Hesse, Ulrich; Sieberer, Johann

    2006-01-01

    1 - Description of program or function: In OREST, the 1-dimensional lattice code HAMMER and the isotope generation and depletion code ORIGEN are directly coupled for burnup simulation in light-water reactor fuels (GRS recommended). Additionally heavy water and graphite moderated systems can be calculated. New version differs from the previous version in the following features: An 84-group-library LIB84 for up to 200 isotopes is used to update the 3-group -POISON-XS. LIB84 uses the same energy boundaries as THERMOS and HAMLET in . In this way, high flexibility is achieved in very different reactor models. The coupling factor between THERMOS and HAMLET is now directly transferred from HAMMER to THERES and omits the equation 4 (see page 6 of the manual). Sandwich-reactor fuel reactivity and burnup calculations can be started with NGEOM = 1. Thorium graphite reactivity and burnup calculations can be started with NLIBE = 1. High enriched U-235 heavy water moderated reactivity and burnup calculations can be started. HAMLET libraries in for U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-242, Am-241, Am-243 and Zirconium are updated using resonance parameters. NEA-1324/04: A new version of the module hamme97.f has replaced the old one. 2 - Method of solution: For the user-defined irradiation history, an input data processor generates program loops over small burnup steps for the main codes HAMMER and ORIGEN. The user defined assembly description is transformed to an equivalent HAMMER fuel cell. HAMMER solves the integral neutron transport equation in a four-region cylindrical or sandwiched model with reflecting boundaries and runs with fuel power calculated rod temperatures. ORIGEN runs with HAMMER-calculated cross sections and neutron spectra and calculates isotope concentrations during burnup by solving the buildup-, depletion- and decay-chain equations. An output data processor samples the outputs of the program modules and generates tabular works for the

  16. Triton burnup in JET

    International Nuclear Information System (INIS)

    Chipsham, E.; Jarvis, O.N.; Sadler, G.

    1989-01-01

    Triton burnup measurements have been made at JET using time-integrated copper activation and time-resolved silicon detector techniques. The results confirm the classical nature of both the confinement and the slowing down of the 1 MeV tritons in a plasma. (author) 8 refs., 3 figs

  17. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  18. Water-Based Pressure-Sensitive Paints

    Science.gov (United States)

    Jordan, Jeffrey D.; Watkins, A. Neal; Oglesby, Donald M.; Ingram, JoAnne L.

    2006-01-01

    Water-based pressure-sensitive paints (PSPs) have been invented as alternatives to conventional organic-solvent-based pressure-sensitive paints, which are used primarily for indicating distributions of air pressure on wind-tunnel models. Typically, PSPs are sprayed onto aerodynamic models after they have been mounted in wind tunnels. When conventional organic-solvent-based PSPs are used, this practice creates a problem of removing toxic fumes from inside the wind tunnels. The use of water-based PSPs eliminates this problem. The waterbased PSPs offer high performance as pressure indicators, plus all the advantages of common water-based paints (low toxicity, low concentrations of volatile organic compounds, and easy cleanup by use of water).

  19. Water Delivery--It's All about Pressure

    Science.gov (United States)

    Roman, Harry T.

    2005-01-01

    There is a great deal of wisdom in the old saying "water seeks its level." In fact, the concept has bearing on a very practical side of human life as well, since the public water delivery system is based on it. In this article, the author discusses the concept behind water pressure and describes how the water systems work based on this concept.…

  20. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  1. Low cost sonoluminescence experiment in pressurized water

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, L; Insabella, M [LADOP, University of Mar del Plata (Argentina); Bilbao, L [INFIP, University of Buenos Aires and CONICET (Argentina)

    2012-06-19

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  2. Low cost sonoluminescence experiment in pressurized water

    International Nuclear Information System (INIS)

    Bernal, L; Insabella, M; Bilbao, L

    2012-01-01

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  3. Water-Based Pressure Sensitive Paint

    Science.gov (United States)

    Oglesby, Donald M.; Ingram, JoAnne L.; Jordan, Jeffrey D.; Watkins, A. Neal; Leighty, Bradley D.

    2004-01-01

    Preparation and performance of a water-based pressure sensitive paint (PSP) is described. A water emulsion of an oxygen permeable polymer and a platinum porphyrin type luminescent compound were dispersed in a water matrix to produce a PSP that performs well without the use of volatile, toxic solvents. The primary advantages of this PSP are reduced contamination of wind tunnels in which it is used, lower health risk to its users, and easier cleanup and disposal. This also represents a cost reduction by eliminating the need for elaborate ventilation and user protection during application. The water-based PSP described has all the characteristics associated with water-based paints (low toxicity, very low volatile organic chemicals, and easy water cleanup) but also has high performance as a global pressure sensor for PSP measurements in wind tunnels. The use of a water-based PSP virtually eliminates the toxic fumes associated with the application of PSPs to a model in wind tunnels.

  4. Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

    International Nuclear Information System (INIS)

    Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

    2016-01-01

    At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

  5. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

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

  6. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  7. Safety aspects of using gadolinium as burnable poison in pressurized water reactors

    International Nuclear Information System (INIS)

    Vandenberg, C.; Bonet, H.; Charlier, A.; Motte, F.

    1979-01-01

    Within the framework of an experimental program on the behavior of gadolinium in light water reactors (LWRs), the BR3 power plant, a small 11-MW(electric) pressurized water reactor, was operated successfully with a core containing 5% Gd 2 O 3 -UO 2 rods. The core reached an average burnup increase of 22,000 MWd/tM, corresponding to 500 effective full-power days in a single cycle. These results were used to extrapolate the consequences on safety of extending such a control policy to large LWRs. In this context, the following factors were investigated: impact on the design, reactivity control and core behavior operated with lower and more constant boric acid concentration, environmental impact, fuel handling, etc

  8. Water-Pressure Distribution on Seaplane Float

    Science.gov (United States)

    Thompson, F L

    1929-01-01

    The investigation presented in this report was conducted for the purpose of determining the distribution and magnitude of water pressures likely to be experienced on seaplane hulls in service. It consisted of the development and construction of apparatus for recording water pressures lasting one one-hundredth second or longer and of flight tests to determine the water pressures on a UO-1 seaplane float under various conditions of taxiing, taking off, and landing. The apparatus developed was found to operate with satisfactory accuracy and is suitable for flight tests on other seaplanes. The tests on the UO-1 showed that maximum pressures of about 6.5 pounds per square inch occur at the step for the full width of the float bottom. Proceeding forward from the step the maximum pressures decrease in magnitude uniformly toward the bow, and the region of highest pressures narrows toward the keel. Immediately abaft the step the maximum pressures are very small, but increase in magnitude toward the stern and there once reached a value of about 5 pounds per square inch. (author)

  9. Simulation of the behaviour of nuclear fuel under high burnup conditions

    International Nuclear Information System (INIS)

    Soba, Alejandro; Lemes, Martin; González, Martin Emilio; Denis, Alicia; Romero, Luis

    2014-01-01

    Highlights: • Increasing the time of nuclear fuel into reactor generates high burnup structure. • We analyze model to simulate high burnup scenarios for UO 2 nuclear fuel. • We include these models in the DIONISIO 2.0 code. • Tests of our models are in very good agreement with experimental data. • We extend the range of predictability of our code up to 60 MWd/KgU average. - Abstract: In this paper we summarize all the models included in the latest version of the DIONISIO code related to the high burnup scenario. Due to the extension of nuclear fuels permanence under irradiation, physical and chemical modifications are developed in the fuel material, especially in the external corona of the pellet. The codes devoted to simulation of the rod behaviour under irradiation need to introduce modifications and new models in order to describe those phenomena and be capable to predict the behaviour in all the range of a general pressurized water reactor. A complex group of subroutines has been included in the code in order to predict the radial distribution of power density, burnup, concentration of diverse nuclides and porosity within the pellet. The behaviour of gadolinium as burnable poison also is modelled into the code. The results of some of the simulations performed with DIONISIO are presented to show the good agreement with the data selected for the FUMEX I/II/III exercises, compiled in the NEA data bank

  10. Method for adding additional isotopes to actinide-only burnup credit

    International Nuclear Information System (INIS)

    Lancaster, D.B.; Fuentes, E.; Kang, C.

    1998-01-01

    The Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages requires computer code validation to be performed against a benchmark set of chemical assays for isotopic concentration and against a benchmark set of critical experiments for package criticality. Both sets contain all the isotopes included in the methodology. The chemical assays used include the uranium and plutonium isotopes, while the critical experiments were composed of UO 2 or MOX rods, covering the isotopes in the actinide only approach. Since other isotopes are not included in the validation benchmark sets, it would be necessary to justify both the content and worth of any additional isotope for which burnup credit is to be taken (i.e., both the concentration and criticality effect of each particular isotope must be validated). A method is proposed here that can be used for any number of additional isotopes. As does the actinide-only burnup credit methodology, this method makes use of chemical assay data to establish the conservatism in the prediction of each isotope's concentration. Criticality validation is also performed using a benchmark set of UO 2 and MOX critical experiments, where the additional isotopes are validated using worth experiments to conservatively account for any uncertainty in their cross sections. The remaining requirements (analysis and modeling parameters, loading criteria generation, and physical implementation and controls) are performed exactly as described in the actinide-only burnup credit methodology. This report provides insight into each particular requirement in the new methodology

  11. 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.

  12. Evaluation of burnup credit for accommodating PWR spent nuclear fuel in high-capacity cask designs

    International Nuclear Information System (INIS)

    Wagner, John C.

    2003-01-01

    This paper presents an evaluation of the amount of burnup credit needed for high-density casks to transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic 32-assembly cask and the current regulatory guidance were used as bases for this evaluation. By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based loading curves, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of SNF assemblies in high-capacity storage and transportation casks. The impact of varying selected calculational assumptions is also investigated, and considerable improvement in effectiveness is shown with the inclusion of the principal fission products (FPs) and minor actinides and the use of a bounding best-estimate approach for isotopic validation. Given sufficient data for validation, the most significant component that would improve accuracy, and subsequently enhance the utilization of burnup credit, is the inclusion of FPs. (author)

  13. Pressure dependence of dynamical heterogeneity in water

    International Nuclear Information System (INIS)

    Teboul, Victor

    2008-01-01

    Using molecular dynamics simulations we investigate the effect of pressure on the dynamical heterogeneity in water. We show that the effect of a pressure variation in water is qualitatively different from the effect of a temperature variation on the dynamical heterogeneity in the liquid. We observe a strong decrease of the aggregation of molecules of low mobility together with a decrease of the characteristic time associated with this aggregation. However, the aggregation of the most mobile molecules and the characteristic time of this aggregation are only slightly affected. In accordance with this result, the non-Gaussian parameter shows an important decrease with pressure while the characteristic time t* of the non-Gaussian parameter is only slightly affected. These results highlight then the importance of pressure variation investigations in low temperature liquids on approach to the glass transition

  14. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  15. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  16. High pressure water jet cutting and stripping

    Science.gov (United States)

    Hoppe, David T.; Babai, Majid K.

    1991-01-01

    High pressure water cutting techniques have a wide range of applications to the American space effort. Hydroblasting techniques are commonly used during the refurbishment of the reusable solid rocket motors. The process can be controlled to strip a thermal protective ablator without incurring any damage to the painted surface underneath by using a variation of possible parameters. Hydroblasting is a technique which is easily automated. Automation removes personnel from the hostile environment of the high pressure water. Computer controlled robots can perform the same task in a fraction of the time that would be required by manual operation.

  17. High pressure water jet mining machine

    Science.gov (United States)

    Barker, Clark R.

    1981-05-05

    A high pressure water jet mining machine for the longwall mining of coal is described. The machine is generally in the shape of a plowshare and is advanced in the direction in which the coal is cut. The machine has mounted thereon a plurality of nozzle modules each containing a high pressure water jet nozzle disposed to oscillate in a particular plane. The nozzle modules are oriented to cut in vertical and horizontal planes on the leading edge of the machine and the coal so cut is cleaved off by the wedge-shaped body.

  18. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  19. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  20. Low cost sonoluminescence experiment in pressurized water

    Science.gov (United States)

    Bernal, L.; Insabella, M.; Bilbao, L.

    2012-06-01

    We present a low cost design for demostration and mesurements of light emmision from a sonoluminescence experiment. Using presurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  1. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  2. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  3. ABB high burnup fuel

    International Nuclear Information System (INIS)

    Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

  4. Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

    International Nuclear Information System (INIS)

    Roudier, S.; Badel, D.; Beraha, R.; Champ, M.; Tricot, N.; Tran Dai, P.

    1994-01-01

    The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: 1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; 2) guidelines for nuclear design and manufacturing requirements related to safety and 3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs

  5. Iterative solution to the optimal poison management problem in pressurized water reactors

    International Nuclear Information System (INIS)

    Colletti, J.P.; Levine, S.H.; Lewis, J.B.

    1983-01-01

    A new method for solving the optimal poison management problem for a multiregion pressurized water reactor has been developed. The optimization objective is to maximize the end-of-cycle core excess reactivity for any given beginning-of-cycle fuel loading. The problem is treated as an optimal control problem with the region burnup and control absorber concentrations acting as the state and control variables, respectively. Constraints are placed on the power peaking, soluble boron concentration, and control absorber concentrations. The solution method consists of successive relinearizations of the system equations resulting in a sequence of nonlinear programming problems whose solutions converge to the desired optimal control solution. Application of the method to several test problems based on a simplified three-region reactor suggests a bang-bang optimal control strategy with the peak power location switching between the inner and outer regions of the core and the critical soluble boron concentration as low as possible throughout the cycle

  6. Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

    Energy Technology Data Exchange (ETDEWEB)

    Roudier, S [Direction de la Surete des Installations Nucleaires, Fontenay-aux-Roses (France); Badel, D; Beraha, R [Direction Regionale de l` Industrie, de la Recherche et de l` Environnement Rhone-Alpes, Lyon (France); Champ, M; Tricot, N; Tran Dai, P [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

    1994-12-31

    The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: (1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; (2) guidelines for nuclear design and manufacturing requirements related to safety and (3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs.

  7. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

    2001-01-01

    Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

  8. Benefits of the delta K of depletion benchmarks for burnup credit validation

    International Nuclear Information System (INIS)

    Lancaster, D.; Machiels, A.

    2012-01-01

    Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO 2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, k eff . The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

  9. Status of burnup credit for transport of SNF in the United States

    International Nuclear Information System (INIS)

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

    2004-01-01

    Allowing credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transportation, and disposal of spent nuclear fuel (SNF) while maintaining a subcritical margin sufficient to establish an adequate safety basis. This paper reviews the current status of burnup credit applied to the design and transport of SNF casks in the United States. The existing U.S. regulatory guidance on burnup credit is limited to pressurized-water-reactor (PWR) fuel and to allowing credit only for actinides in the SNF. By comparing loading curves against actual SNF discharge data for U.S. reactors, the potential benefits that can be realized using the current regulatory guidance with actinide-only burnup credit are illustrated in terms of the inventory allowed in high-capacity casks and the concurrent reduction in SNF shipments. The additional benefits that might be realized by extending burnup credit to credit for select fission products are also illustrated. The curves show that, although fission products in SNF provide a small decrease in reactivity compared with actinides, the additional negative reactivity causes the SNF inventory acceptable for transportation to increase from roughly 30% to approximately 90% when fission products are considered. A savings of approximately $150M in transport costs can potentially be realized for the planned inventory of the repository. Given appropriate experimental data to support code validation, a realistic best-estimate analysis of burnup credit that includes validated credit for fission products is the enhancement that will yield the most significant impact on future transportation plans

  10. The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

  11. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

  12. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  13. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  14. Chloride Ingress into Concrete under Water Pressure

    DEFF Research Database (Denmark)

    Lund, Mia Schou; Sander, Lotte Braad; Grelk, Bent

    2011-01-01

    The chloride ingress into concrete under water pressures of 100 kPa and 800 kPa have been investigated by experiments. The specimens were exposed to a 10% NaCl solution and water mixture. For the concrete having w/c = 0.35 the experimental results show the chloride diffusion coefficient at 800 k......Pa (~8 atm.) is 12 times greater than at 100 kPa (~1 atm.). For w/c = 0.45 and w/c = 0.55 the chloride diffusion coefficients are 7 and 3 times greater. This means that a change in pressure highly influences the chloride ingress into the concrete and thereby the life length models for concrete structures....

  15. Validating analysis methodologies used in burnup credit criticality calculations

    International Nuclear Information System (INIS)

    Brady, M.C.; Napolitano, D.G.

    1992-01-01

    The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

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

    International Nuclear Information System (INIS)

    Wagner, J.C.

    2001-01-01

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

  17. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  18. Impact of extended burnup on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-04-01

    The Advisory Group Meeting was held in Vienna from 2 to 5 December 1991, to review, analyse, and discuss the effects of burnup extension in both light and heavy water reactors on all aspects of the fuel cycle. Twenty experts from thirteen countries participated in this meeting. There was consensus that both economic and environmental benefits are driving forces toward the achievement of higher burnups and that the present trend of burnup extension may be expected to continue. The extended burnup has been considered for the three main stages of the fuel cycle: the front end, in-reactor issues and the back end. Thirteen papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  19. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  20. Evaluation of burnup credit for fuel storage analysis -- Experience in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    1995-01-01

    Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (recracking) and the implementation of new analysis approaches with reduced conservation (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); an other BWR storage analysis is being developed at this moment. The elimination of the ''fresh fuel assumption'' increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduced new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis. The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible

  1. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  2. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  3. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  4. Recent developments in high pressure water technology

    International Nuclear Information System (INIS)

    Johnson, N.A.; Johnson, T.

    1992-01-01

    High Pressure Water Jetting has advanced rapidly in the last decade to a point where the field is splitting into specialised areas. This has left the end user or client in the dark as to whether water jetting will work and if so what equipment is best suited to their particular application. The aim of this paper is to give an overview of:-1. The way water is delivered to the surface and the parameters which control the concentration of energy available on impact. 2. The factors governing application driven selection of equipment. 3. The effects to technical advances in pumps and delivery systems on equipment selection with reference to their to their application to concrete removal and nuclear decontamination. (Author)

  5. Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

    International Nuclear Information System (INIS)

    Mobasheran, A.S.; Boshoven, J.; Lake, B.

    1995-01-01

    The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit

  6. Modified fuel assembly design for pressurized water reactors with improved fuel utilization

    International Nuclear Information System (INIS)

    Galperin, A.; Ronen, Y.

    1983-01-01

    A method for reactivity control through variation of the moderator content in the reactor core was proposed. The main idea is to adjust the amount of water in the core from a low value at beginning of cycle to a high value at end of cycle, so as to compensate for fissile material burnup and buildup of fission products. The possible implementation of this idea may be carried out by introducing a number of hollow tubes into the fuel assembly between the fuel rods. Then variation of the moderator content in the core may be managed through a change of the water level in these tubes. cated a potential savings in the fuel cycle requirements and costs. Preliminary steady-state thermal-hydraulic calculations indicate the possibility of implementing the proposed method in the existing pressurized water reactor plants. Feasibility of the proposed design may be finally established after rigorous thermal hydraulics as well as safety analysis calculations. Furthermore, there is need to elaborate the mechanical design of the pressure vessel internals together with cost benefit analysis

  7. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  8. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  9. Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz

    2000-01-01

    This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

  10. 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).

  11. Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

    Directory of Open Access Journals (Sweden)

    Gajda Paweł

    2016-01-01

    Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

  12. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  13. Improvements for Monte Carlo burnup calculation

    Energy Technology Data Exchange (ETDEWEB)

    Shenglong, Q.; Dong, Y.; Danrong, S.; Wei, L., E-mail: qiangshenglong@tsinghua.org.cn, E-mail: d.yao@npic.ac.cn, E-mail: songdr@npic.ac.cn, E-mail: luwei@npic.ac.cn [Nuclear Power Inst. of China, Cheng Du, Si Chuan (China)

    2015-07-01

    Monte Carlo burnup calculation is development trend of reactor physics, there would be a lot of work to be done for engineering applications. Based on Monte Carlo burnup code MOI, non-fuel burnup calculation methods and critical search suggestions will be mentioned in this paper. For non-fuel burnup, mixed burnup mode will improve the accuracy of burnup calculation and efficiency. For critical search of control rod position, a new method called ABN based on ABA which used by MC21 will be proposed for the first time in this paper. (author)

  14. Pressure pressure-balanced pH sensing system for high temperature and high pressure water

    International Nuclear Information System (INIS)

    Tachibana, Koji

    1995-01-01

    As for the pH measurement system for high temperature, high pressure water, there have been the circumstances that first the reference electrodes for monitoring corrosion potential were developed, and subsequently, it was developed for the purpose of maintaining the soundness of metallic materials in high temperature, high pressure water in nuclear power generation. In the process of developing the reference electrodes for high temperature water, it was clarified that the occurrence of stress corrosion cracking in BWRs is closely related to the corrosion potential determined by dissolved oxygen concentration. As the types of pH electrodes, there are metal-hydrogen electrodes, glass electrodes, ZrO 2 diaphragm electrodes and TiO 2 semiconductor electrodes. The principle of pH measurement using ZrO 2 diaphragms is explained. The pH measuring system is composed of YSZ element, pressure-balanced type external reference electrode, pressure balancer and compressed air vessel. The stability and pH response of YSZ elements are reported. (K.I.)

  15. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  16. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

    Melin, P.; Lavoine, O.; Houdaille, B.

    1986-04-01

    The examinations carried out on the FRAGEMA fuel of EDF reactors show its good behavior in service. The results of research and development programs developed by EDF, FGA and the CEA show that this fuel can be irradiated up to a high burnup, and allow to point out the axies of research to improve still the performance of the product in a more and more soliciting environment (increase of power and burnup coupled with load following). Among the solutions considered, there are the design and fabrication adjustments (geometry, initial pressurization), more fundamental changes concerning fuel cans and fuel pellets, which need still research and development programs [fr

  17. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  18. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

    EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

  19. Status of advanced small pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Peipei; Zhou Yun

    2012-01-01

    In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

  20. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

  1. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  2. Pressure equalization systems in pressurized water reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.; Wunderlich, F.

    1979-01-01

    For the development of a pressure reduction system in PWR fuel rods the capability of charcoal to adsorb Helium, Xenon and Krypton at temperatures of about 300 0 C was investigated. The influence of the adsorption on fuel rod internal pressure and in creep strain on the tube was evaluated in a design study. (orig.) [de

  3. Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2016-01-01

    Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

  4. Water Pressure Distribution on a Flying Boat Hull

    Science.gov (United States)

    Thompson, F L

    1931-01-01

    This is the third in a series of investigations of the water pressures on seaplane floats and hulls, and completes the present program. It consisted of determining the water pressures and accelerations on a Curtiss H-16 flying boat during landing and taxiing maneuvers in smooth and rough water.

  5. Analysis of the behavior under irradiation of high burnup nuclear fuels with the computer programs FRAPCON and FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Regis; Silva, Antonio Teixeira e, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The objective of this paper is to verify the validity and accuracy of the results provided by computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods, in steady-state and transient operational conditions at high burnup. To achieve this goal, the results provided by these computer simulations are compared with experimental data available in the database FUMEX III. Through the results, it was found that the computer programs used have a good ability to predict the operational behavior of PWR fuel rods in high burnup steady-state conditions and under the influence of Reactivity Initiated Accident (RIA). (author)

  6. Disposal criticality analysis methodology's principal isotope burnup credit

    International Nuclear Information System (INIS)

    Doering, T.W.; Thomas, D.A.

    2001-01-01

    This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)

  7. High burnup MOX fuel assembly

    International Nuclear Information System (INIS)

    Blanpain, P.; Brunel, L.

    1999-01-01

    From the outset, the MOX product was required to have the same performance as UO 2 in terms of burnup and operational flexibility. In fact during the first years the UO 2 managements could not be applied to MOX. The changeover to an AFA 2G type fuel allowed an improvement in NPP operational flexibility. The move to the AFA 3G design fuel will enable an increase in the burnup of the MOX assemblies to the level of the UO 2 ones ('MOX Parity' project). But the FRAMATOME fuel development objective does not stop at the obtaining of parity between the current MOX and UO 2 products: this parity must remain guaranteed and the MOX managements must evolve in the same way as the UO 2 managements. The goal of the MOX product development programmes underway with COGEMA and the CEA is the demonstration over the next 10 years of a fuel capable of reaching burnups of 70 GWD/T. The research programmes focus on the fission gas release aspect, with three issues explored: optimization of pellet microstructures and validation in experimental reactor ; build-up of experience feedback from fission gas release at elevated burnups in commercial reactors, both for current and experimental products; adaptation and qualification of the design models and tools, over the ranges and for the products concerned. The product arising from these development programmes should be offered on the market around 2010. While meeting safety requirements, it will cater for the needs of the utilities in terms of product reliability, personnel dosimetry and kWh output costs (increase in burnup, NPP maneuverability and availability, minimization of process waste). (authors)

  8. Fuel cycle cost considerations of increased discharge burnups

    International Nuclear Information System (INIS)

    Scherpereel, L.R.; Frank, F.J.

    1982-01-01

    Evaluations are presented that indicate the attainment of increased discharge burnups in light water reactors will depend on economic factors particular to individual operators. In addition to pure resource conserving effects and assuming continued reliable fuel performance, a substantial economic incentive must exist to justify the longer operating times necessary to achieve higher burnups. Whether such incentive will exist or not will depend on relative price levels of all fuel cycle cost components, utility operating practices, and resolution of uncertainties associated with the back-end of the fuel cycle. It is concluded that implementation of increased burnups will continue at a graduated pace similar to past experience, rather than finding universal acceptance of particular increased levels at any particular time

  9. Determining Atmospheric Pressure Using a Water Barometer

    Science.gov (United States)

    Lohrengel, C. Frederick, II; Larson, Paul R.

    2012-01-01

    The atmosphere is an envelope of compressible gases that surrounds Earth. Because of its compressibility and nonuniform heating by the Sun, it is in constant motion. The atmosphere exerts pressure on Earth's surface, but that pressure is in constant flux. This experiment allows students to directly measure atmospheric pressure by measuring the…

  10. Computerized cost model for pressurized water reactors

    International Nuclear Information System (INIS)

    Meneely, T.K.; Tabata, Hiroaki; Labourey, P.

    1999-01-01

    A computerized cost model has been developed in order to allow utility users to improve their familiarity with pressurized water reactor overnight capital costs and the various factors which influence them. This model organizes its cost data in the standard format of the Energy Economic Data Base (EEDB), and encapsulates simplified relationships between physical plant design information and capital cost information in a computer code. Model calculations are initiated from a base case, which was established using traditional cost calculation techniques. The user enters a set of plant design parameters, selected to allow consideration of plant models throughout the typical three- and four-loop PWR power range, and for plant sites in Japan, Europe, and the United States. Calculation of the new capital cost is then performed in a very brief time. The presentation of the program's output allows comparison of various cases with each other or with separately calculated baseline data. The user can start at a high level summary, and by selecting values of interest on a display grid show progressively more and more detailed information, including links to background information such as individual cost driver accounts and physical plant variables for each case. Graphical presentation of the comparison summaries is provided, and the numerical results may be exported to a spreadsheet for further processing. (author)

  11. Evaluation of pressure transducers under turbid natural waters

    Digital Repository Service at National Institute of Oceanography (India)

    Joseph, A.; Desa, E.; Desa, E.; Smith, D.; Peshwe, V.B.; VijayKumar, K.; Desa, J.A.E.

    , the use of rho sub(eff) in contrast to the bulk density, significantly improves the measurement accuracy. For celar waters, precision density measurements made on discrete water samples agreed with rho sub(eff) values derived from pressure measurements...

  12. Analysis of strategies for improving uranium utilization in pressurized water reactors. Annual technical progress report for FY 1980

    International Nuclear Information System (INIS)

    Sefcik, J.A.; Driscoll, M.J.; Lanning, D.D.

    1981-01-01

    Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used to verify and supplement these techniques

  13. Effects of pellet-to-cladding gap design parameters on the reliability of high burnup PWR fuel rods under steady state and transient conditions

    International Nuclear Information System (INIS)

    Tas, Fatma Burcu; Ergun, Sule

    2013-01-01

    Highlights: • Fuel performance of a typical Pressurized Water Reactor rod is analyzed. • Steady state fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • Transient fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • The optimum pellet to cladding gap thickness and gap gas pressure values of the simulated fuel are determined. • The effects of pellet to cladding gap design parameters on nuclear fuel reliability are examined. - Abstract: As an important improvement in the light water nuclear reactor operations, the nuclear fuel burnup rate is increased in recent decades and this increase causes heavier duty for the nuclear fuel. Since the high burnup fuel is exposed to very high thermal and mechanical stresses and since it operates in an environment with high radiation for about 18 month cycles, it carries the risk of losing its integrity. In this study; it is aimed to determine the effects of pellet–cladding gap thickness and gap pressure on reliability of high burnup nuclear fuel in Pressurized Water Reactors (PWRs) under steady state operation conditions and suggest optimum values for the examined parameters only and validate these suggestions for a transient condition. In the presented study, fuel performance was analyzed by examining the effects of pellet–cladding gap thickness and gap pressure on the integrity of high burnup fuels. This work is carried out for a typical Westinghouse type PWR fuel. The steady state conditions were modeled and simulated with FRAPCON-3.4a steady state fuel performance code and the FRAPTRAN-1.4 fuel transient code was used to calculate transient fuel behavior. The analysis included the changes in the important nuclear fuel design limitations such as the centerline temperature, cladding stress, strain and oxidation with the change in pellet–cladding gap thickness and initial pellet–cladding gap gas

  14. Cycle studies: material balance estimation in the domain of pressurized water and boiling water reactors. Experimental qualification

    International Nuclear Information System (INIS)

    Chabert, Christine

    1994-01-01

    This study is concerned with the physics of the fuel cycle the aim being to develop and make recommendations concerning schemes for calculating the neutronics of light water reactor fuel cycles. A preliminary study carried out using the old fuel cycle calculation scheme APOLLO1- KAFKA and the library SERMA79 has shown that for the compositions of totally dissolved assemblies from Pressurized Water Reactors (type 17*17) and also for the first time, for Boiling Water Reactor assemblies (type 8*8), the differences between calculation and measurement are large and must be reduced. The integration of the APOLLO2 neutronics code into the fuel cycle calculation scheme improves the results because it can model the situation more precisely. A comparison between APOLLO1 and APOLLO2 using the same options, demonstrated the consistency of the two methods for PWR and BWR geometries. Following this comparison, we developed an optimised scheme for PWR applications using the library CEA86 and the code APOLLO2. Depending on whether the information required is the detailed distribution of the composition of the irradiated fuel or the average composition (estimation of the total material balance of the fuel assembly), the physics options recommended are different. We show that the use of APOLLO2 and the library CEA86 improves the results and especially the estimation of the Pu 239 content. Concerning the Boiling Water Reactor, we have highlighted the need to treat several axial sections of the fuel assembly (variation of the void-fraction, heterogeneity of composition). A scheme using Sn transport theory, permits one to obtain a better coherence between the consumption of U 235 , the production of plutonium and burnup, and a better estimation of the material balance. (author) [fr

  15. Pressurized water reactor with a reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1979-01-01

    The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

  16. Adaptive Reference Control for Pressure Management in Water Networks

    DEFF Research Database (Denmark)

    Kallesøe, Carsten; Jensen, Tom Nørgaard; Wisniewski, Rafal

    2015-01-01

    Water scarcity is an increasing problem worldwide and at the same time a huge amount of water is lost through leakages in the distribution network. It is well known that improved pressure control can lower the leakage problems. In this work water networks with a single pressure actuator and several....... Subsequently, these relations are exploited in an adaptive reference control scheme for the actuator pressure that ensures constant pressure at the critical points. Numerical experiments underpin the results. © Copyright IEEE - All rights reserved....

  17. Capillarity Induced Negative Pressure of Water Plugs in Nanochannels

    NARCIS (Netherlands)

    Tas, Niels Roelof; Mela, P.; Kramer, Tobias; Berenschot, Johan W.; van den Berg, Albert

    2003-01-01

    We have found evidence that water plugs in hydrophilic nanochannels can be at significant negative pressure due to tensile capillary forces. The negative pressure of water plugs in nanochannels induces bending of the thin channel capping layer, which results in a visible curvature of the liquid

  18. Device for measuring a burnup degree

    International Nuclear Information System (INIS)

    Ito, Toshiaki; Goto, Seiichiro

    1979-01-01

    Purpose: To measure the burnup degree at high efficiency and accuracy. Constitution: The outer metal wall of fuel assemblies is heated under gamma radiation with long half life gamma rays in inverse proportion to the burnup degree and issues infrared radiation in proportion to the intensity of the gamma rays. An image pick-up tube is opposed to one surface of the fuel assemblies to detect the radiated infrared rays. Since the output signal from the pick-up tube is subjected to the absorptive damping by the distance between the pick-up tube and the fuel assembly, as well as water filled in the gap therebetween, it is corrected through a main amplifier comprising a signal correction circuit composed of a characteristic section inverse to the absorption property and a characteristic section inverse to the square of the distance. The corrected output signal is displayed on a display unit such as CRT or recorded in a film or a magnetic tape. (Furukawa, Y.)

  19. Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

    Directory of Open Access Journals (Sweden)

    Geoff Recktenwald

    2013-04-01

    Full Text Available Zirconium dioxide has received particular attention as a fuel matrix because of its ability to form a solid solution with transuranic elements, natural radiation stability and desirable mechanical properties. However, zirconium dioxide has a lower coefficient of thermal conductivity than uranium dioxide and this presents an obstacle to the deployment of these fuels in commercial reactors. Here we show that axial doping of a zirconium dioxide based fuel with erbium reduces power peaking and fuel temperature. Full core simulations of a modified AP1000 core were done using MCNPX 2.7.0. The inert matrix fuel contained 15 w/o transuranics at its beginning of life and constituted 28% of the assemblies in the core. Axial doping reduced power peaking at startup by more than ~23% in the axial direction and reduced the peak to average power within the core from 1.80 to 1.44. The core was able to remain critical between refueling while running at a simulated 2000 MWth on an 18 month refueling cycle. The results show that the reactor would maintain negative core average reactivity and void coefficients during operation. This type of fuel cycle would reduce the overall production of transuranics in a pressurized water reactor by 86%.

  20. Improvements on burnup chain model and group cross section library in the SRAC system

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.

    1992-01-01

    Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)

  1. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  2. The commercial impact of burnup increase

    International Nuclear Information System (INIS)

    Fenzlein, C.; Schricker, W.

    2002-01-01

    Deregulation has a dramatic effect on competition in the electricity markets. This will lead to a continued pressure on the prices in virtually all areas of the nuclear fuel cycle and will encourage further optimization, technical and technological progress and innovations with respect to further cost reductions of power production. The permission of direct disposal, in Germany legally granted in 1994 as an alternative to the reprocessing path, made possible cost savings and has consequently resulted in a decline of reprocessing prices. In addition, suppliers as well as operators are making considerable efforts to reduce the disposal costs fraction by optimizing disposal technologies and concepts. The increase of discharge has essentially contributed to the reduction the disposal cost fraction. Compared to former scenarios, the economic potential of burn-up increase is decreasing

  3. Simulation of High Burnup Structure in UO2 Using Potts Model

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho

    2009-01-01

    The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels

  4. An automated optimization of core fuel loading pattern for pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Renji

    1988-11-01

    An optimum method was adopted to search for an optimum fuel loading pattern in pressurized water reactors. A radial power peak factor was chosen as the objective function of the optimum loading. The direct search method with shuffling rules is used to find optimum solution. The search for an optimum loading pattern with the smallest radial power peak by exchanging fuel assemblies was made. The search process is divided into two steps. In the first step fresh fuels or high reactivity fuels are arranged which are placed in core interior to have a reasonable fuel loading pattern. To further reduce the radial power peak factor, the second step will be necessary to rearrange the exposed lower reactivity fuel around the assemblies which has the radial power peak. In optimum process 1.5 group coarse mesh diffusion theory or two group nodal Green function diffusion theory is utilized to calculate the two dimensional power distribution after each shuffle. Also, above two methods are combinatively utilized to calculate the state quantity. It is not only true to save CPU time, but also can obtian exact results. Besides above function, the code MSOFEL is used to search critical boron concentration and to predict burn-up. The code has been written with FORTRAN-4. The optimum loading pattern was chosen for OCONEE and QINSHAN nuclear power plants as reference examples. The validity and feasibility of MSOFEL was demonstrated

  5. AUTOLOAD, an automatic optimal pressurized water reactor reload design system with an expert module

    International Nuclear Information System (INIS)

    Li, Z.; Levine, S.H.

    1994-01-01

    An automatic optimal pressurized water reactor (PWR) reload design expert system AUTOLOAD has been developed. It employs two important new techniques. The first is a new loading priority scheme that defines the optimal placement of the fuel in the core that has the maximum end-of-cycle state k eff . The second is a new power-shape-driven progressive iteration method for automatically determining the burnable poison (BP) loading in the fresh fuel assemblies. The Haling power distribution is used in converting the theoretically optimal solution into the practical design, which meets the design constraints for the given fuel assemblies. AUTOLOAD is a combination of C and FORTRAN languages. It requires only the required cycle length, the maximum peak normalized power, the BP type, the number of fresh fuel assemblies, the assembly burnup, and BP histories of the available fuel assemblies as its input. Knowledge-based modules have been built into the expert system computer code to perform all of the tasks involved in reloading a PWR. AUTOLOAD takes only ∼ 30 CPU min on an IBM 3090 600s mainframe to accomplish a practical reload design. A maximum of 12.5% fresh fuel enrichment saving is observed compared with the core used by the utility

  6. Iterative solution to the optimal control of depletion problem in pressurized water reactors

    International Nuclear Information System (INIS)

    Colletti, J.P.

    1981-01-01

    A method is described for determining the optimal time and spatial dependence of control absorbers in the core of a pressurized water reactor over a single refueling cycle. The reactor is modeled in two dimensions with many regions using two-group diffusion theory. The problem is formulated as an optimal control problem with the cycle length fixed and the initial reactor state known. Constraints are placed on the regionwise normalized powers, control absorber concentrations, and the critical soluble boron concentration of the core. The cost functional contains two terms which may be used individually or together. One term maximizes the end-of-cycle (EOC) critical soluble boron concentration, and the other minimizes the norm of the distance between the actual and a target EOC burnup distribution. Results are given for several test problems which are based on a three-region model of the Three Mile Island Unit 1 reactor. The resulting optimal control strategies are bang-bang and lead to EOC states with the power peaking at its maximum and no control absorbers remaining in the core. Throughout the cycle the core soluble boron concentration is zero

  7. Automatic optimized reload and depletion method for a pressurized water reactor

    International Nuclear Information System (INIS)

    Ahn, D.H.; Levene, S.H.

    1985-01-01

    A new method has been developed to automatically reload and deplete a pressurized water reactor (PWR) so that both the enriched inventory requirements during the reactor cycle and the cost of reloading the core are minimized. This is achieved through four stepwise optimization calculations: (a) determination of the minimum fuel requirement for an equivalent three-region core model, (b) optimal selection and allocation of fuel assemblies for each of the three regions to minimize the reload cost, (c) optimal placement of fuel assemblies to conserve regionwise optimal conditions, and (d) optimal control through poison management to deplete individual fuel assemblies to maximize end-of-cycle k /SUB eff/ . The new method differs from previous methods in that the optimization process automatically performs all tasks required to reload and deplete a PWR. In addition, the previous work that developed optimization methods principally for the initial reactor cycle was modified to handle subsequent cycles with fuel assemblies having burnup at beginning of cycle. Application of the method to the fourth reactor cycle at Three Mile Island Unit 1 has shown that both the enrichment and the number of fresh reload fuel assemblies can be decreased and fully amortized fuel assemblies can be reused to minimize the fuel cost of the reactor

  8. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  9. In Situ Raman Study of Liquid Water at High Pressure.

    Science.gov (United States)

    Romanenko, Alexandr V; Rashchenko, Sergey V; Goryainov, Sergey V; Likhacheva, Anna Yu; Korsakov, Andrey V

    2018-06-01

    A pressure shift of Raman band of liquid water (H 2 O) may be an important tool for measuring residual pressures in mineral inclusions, in situ barometry in high-pressure cells, and as an indicator of pressure-induced structural transitions in H 2 O. However, there was no consensus as to how the broad and asymmetric water Raman band should be quantitatively described, which has led to fundamental inconsistencies between reported data. In order to overcome this issue, we measured Raman spectra of H 2 O in situ up to 1.2 GPa using a diamond anvil cell, and use them to test different approaches proposed for the description of the water Raman band. We found that the most physically meaningful description of water Raman band is the decomposition into a linear background and three Gaussian components, associated with differently H-bonded H 2 O molecules. Two of these components demonstrate a pronounced anomaly in pressure shift near 0.4 GPa, supporting ideas of structural transition in H 2 O at this pressure. The most convenient approach for pressure calibration is the use of "a linear background + one Gaussian" decomposition (the pressure can be measured using the formula P (GPa) = -0.0317(3)·Δν G (cm -1 ), where Δν G represents the difference between the position of water Raman band, fitted as a single Gaussian, in measured spectrum and spectrum at ambient pressure).

  10. SIMULATION OF NEGATIVE PRESSURE WAVE PROPAGATION IN WATER PIPE NETWORK

    Directory of Open Access Journals (Sweden)

    Tang Van Lam

    2017-11-01

    Full Text Available Subject: factors such as pipe wall roughness, mechanical properties of pipe materials, physical properties of water affect the pressure surge in the water supply pipes. These factors make it difficult to analyze the transient problem of pressure evolution using simple programming language, especially in the studies that consider only the magnitude of the positive pressure surge with the negative pressure phase being neglected. Research objectives: determine the magnitude of the negative pressure in the pipes on the experimental model. The propagation distance of the negative pressure wave will be simulated by the valve closure scenarios with the help of the HAMMER software and it is compared with an experimental model to verify the quality the results. Materials and methods: academic version of the Bentley HAMMER software is used to simulate the pressure surge wave propagation due to closure of the valve in water supply pipe network. The method of characteristics is used to solve the governing equations of transient process of pressure change in the pipeline. This method is implemented in the HAMMER software to calculate the pressure surge value in the pipes. Results: the method has been applied for water pipe networks of experimental model, the results show the affected area of negative pressure wave from valve closure and thereby we assess the largest negative pressure that may appear in water supply pipes. Conclusions: the experiment simulates the water pipe network with a consumption node for various valve closure scenarios to determine possibility of appearance of maximum negative pressure value in the pipes. Determination of these values in real-life network is relatively costly and time-consuming but nevertheless necessary for identification of the risk of pipe failure, and therefore, this paper proposes using the simulation model by the HAMMER software. Initial calibration of the model combined with the software simulation results and

  11. High burnup issues and modelling strategies

    International Nuclear Information System (INIS)

    Dutta, B.K.

    2005-01-01

    The performance of high burnup fuel is affected by a number of phenomena, such as, conductivity degradation, modified radial flux profile, fission gas release from high burnup structures, PCMI, burnup dependent thermo-mechanical properties, etc. The modelling strategies of some of these phenomena are available in literature. These can be readily incorporated in a fuel modelling performance code. The computer code FAIR has been developed in BARC over the years to evaluate the fuel performance at extended burnup and modelling of the fuel rods for advanced fuel cycles. The present paper deals with the high burnup issues in the fuel pins, their modelling strategies and results of the case studies specifically involving high burnup fuel. (author)

  12. Burnup credit activities in the United States

    International Nuclear Information System (INIS)

    Lake, W.H.; Thomas, D.A.; Doering, T.W.

    2001-01-01

    This report covers progress in burnup credit activities that have occurred in the United States of America (USA) since the International Atomic Energy Agency's (IAEA's) Advisory Group Meeting (AGM) on Burnup Credit was convened in October 1997. The Proceeding of the AGM were issued in April 1998 (IAEA-TECDOC-1013, April 1998). The three applications of the use of burnup credit that are discussed in this report are spent fuel storage, spent fuel transportation, and spent fuel disposal. (author)

  13. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  14. Calculation study of the WWER-440 fuel performance for extended burnup

    International Nuclear Information System (INIS)

    Kujal, J.; Pazdera, F.; Barta, O.

    1984-01-01

    The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

  15. CARA design criteria for HWR fuel burnup extension

    International Nuclear Information System (INIS)

    Florido, P.C.; Cirimello, R.O.; Bergallo, J.E.; Marino, A.C.; Delmastro, D.F.; Brasnarof, D.O.; Gonzalez, J.H.; Juanico, L.A.

    2002-01-01

    A new concept for HWR fuel bundles, namely CARA, is presented. The CARA design allows to improve all the major performances in the PHWR fuel technology. Among others, it reaches higher burnup and thermohydraulic safety margins, together with lower fuel pellet temperatures and Zry/HM mass ratio. Moreover, it keeps the fuel mass content per unit length and the channel pressure drop by using a single diameter of fuel rods. (author)

  16. Application of burnup credit concept to transport

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Nakagome, Yoshihiro.

    1994-01-01

    For the design and safety assessment of the casks for transporting spent fuel, the fuel contained in them has been assumed to be new fuel. The reason is, it was difficult to evaluate the variation of the reactivity of fuel, and the research on the affecting factors and the method of measuring burnup were not much advanced. Recently, high burnup fuel has been adopted, and initial degree of enrichment rose. The research has been advanced for pursuing the economy of the casks for spent fuel, and burnup credit has become applicable to their design and safety assessment. As the result, the containing capacity increases by about 20%. When burnup credit is considered, it is necessary to confirm accurately the burnup of spent fuel. The burnup dependence of the concentration of fissile substances and neutron emissivity, the coolant void dependence of the concentration of fissile substances, and the relation of neutron multiplication rate with initial degree of enrichment or burnup are discussed. The conceptual design of casks considering burnup credit and its assessment, the merit, problem and the countermeasures to it when burnup credit is introduced are described. (K.I.)

  17. Burnup credit for storage and transportation casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1988-01-01

    The application of burnup credit to storage and transportation cask licensing results in a significant improvement in cask capacity and an associated reduction of the cost per kilogram of uranium in the cask contents. The issues for licensing with burnup credit deal primarily with the treatment of fission product poisons and methods of verification of burnup during cask operations. Other issues include benchmarking of cross-section sets and codes and the effect of spatial variation of burnup within an assembly. The licensing of burnup credit for casks will be complex, although the criticality calculations are not themselves difficult. Attention should be directed to the use of fission product poisons and the uncertainties that they introduce. Verification of burnup by measurements will remove some of the concerns for criticality safety. Calculations for burnup credit casks should consider rod-to-rod and axial variations of burnup, as well as variability of burnable poisons it they are used in the assembly. In spite of the complexity of cask burnup credit licensing issues, these issues appear to be resolvable within the current state of the art of criticality safety

  18. Re-evaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya, E-mail: suyama.kenya@jaea.go.jp [Office of International Relations, Nuclear Safety Division, Ministry of Education, Culture, Sports, Science and Technology - Japan, 3-2-2 Kasumigaseki, Chiyoda-ku, Tokyo 100-8959 (Japan); Murazaki, Minoru; Ohkubo, Kiyoshi [Fuel Cycle Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai-mura, Ibaraki 319-1195 (Japan); Nakahara, Yoshinori [Research Group for Analytical Science, Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai-mura, Ibaraki 319-1195 (Japan); Uchiyama, Gunzo [Fuel Cycle Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai-mura, Ibaraki 319-1195 (Japan)

    2011-05-15

    Highlights: > The specifications required for the analyses of the destructive assay data taken from irradiated fuel in Ohi-1 and Ohi-2 PWRs were documented in this paper. > These data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. > These destructive assay data are suitable for the benchmarking of the burnup calculation code systems. - Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.

  19. High-pressure-induced water penetration into 3-isopropylmalate dehydrogenase

    International Nuclear Information System (INIS)

    Nagae, Takayuki; Kawamura, Takashi; Chavas, Leonard M. G.; Niwa, Ken; Hasegawa, Masashi; Kato, Chiaki; Watanabe, Nobuhisa

    2012-01-01

    Structures of 3-isopropylmalate dehydrogenase were determined at pressures ranging from 0.1 to 650 MPa. Comparison of these structures gives a detailed picture of the swelling of a cavity at the dimer interface and the generation of a new cleft on the molecular surface, which are accompanied by water penetration. Hydrostatic pressure induces structural changes in proteins, including denaturation, the mechanism of which has been attributed to water penetration into the protein interior. In this study, structures of 3-isopropylmalate dehydrogenase (IPMDH) from Shewanella oneidensis MR-1 were determined at about 2 Å resolution under pressures ranging from 0.1 to 650 MPa using a diamond anvil cell (DAC). Although most of the protein cavities are monotonically compressed as the pressure increases, the volume of one particular cavity at the dimer interface increases at pressures over 340 MPa. In parallel with this volume increase, water penetration into the cavity could be observed at pressures over 410 MPa. In addition, the generation of a new cleft on the molecular surface accompanied by water penetration could also be observed at pressures over 580 MPa. These water-penetration phenomena are considered to be initial steps in the pressure-denaturation process of IPMDH

  20. Solubility and physical properties of sugars in pressurized water

    International Nuclear Information System (INIS)

    Saldaña, Marleny D.A.; Alvarez, Víctor H.; Haldar, Anupam

    2012-01-01

    Highlights: ► Sugar solubility in pressurized water and density at high pressures were measured. ► Glucose solubility was higher than that of lactose as predicted by their σ-profiles. ► Sugar aqueous solubility decreased with an increase in pressure from 15 to 120 bar. ► Aqueous glucose molecular packing shows high sensitivity to pressure. ► The COSMO-SAC model qualitatively predicted the sugar solubility data. - Abstract: In this study, the solubility, density, and refractive index of glucose and lactose in water as a function of temperature were measured. For solubility of sugars in pressurized water, experimental data were obtained at pressures of (15 to 120) bar and temperatures of (373 to 433) K using a dynamic flow high pressure system. Density data for aqueous sugar solutions were obtained at pressures of (1 to 300) bar and temperatures of (298 to 343) K. The refractive index of aqueous sugar solutions was obtained at 293 K and atmospheric pressure. Activity coefficient models, Van Laar and the Conductor-like Screening Model-Segment Activity Coefficient (COSMO-SAC), were used to fit and predict the experimental solubility data, respectively. The results obtained showed that the solubility of both sugars in pressurized water increase with an increase in temperature. However, with the increase of pressure from 15 bar to 120 bar, the solubility of both sugars in pressurized water decreased. The Van Laar model fit the experimental aqueous solubility data with deviations lower than 13 and 53% for glucose and lactose, respectively. The COSMO-SAC model predicted qualitatively the aqueous solubility of these sugars.

  1. Importance of pressure reducing valves (PRVs) in water supply networks.

    Science.gov (United States)

    Signoreti, R. O. S.; Camargo, R. Z.; Canno, L. M.; Pires, M. S. G.; Ribeiro, L. C. L. J.

    2016-08-01

    Challenged with the high rate of leakage from water supply systems, these managers are committed to identify control mechanisms. In order to standardize and control the pressure Pressure Reducing Valves (VRP) are installed in the supply network, shown to be more effective and provide a faster return for the actual loss control measures. It is known that the control pressure is while controlling the occurrence of leakage. Usually the network is sectored in areas defined by pressure levels according to its topography, once inserted the VRP in the same system will limit the downstream pressure. This work aims to show the importance of VRP as loss reduction for tool.

  2. Assessment of water pipes durability under pressure surge

    Science.gov (United States)

    Pham Ha, Hai; Minh, Lanh Pham Thi; Tang Van, Lam; Bulgakov, Boris; Bazhenova, Soafia

    2017-10-01

    Surge phenomenon occurs on the pipeline by the closing valve or pump suddenly lost power. Due to the complexity of the water hammer simulation, previous researches have only considered water hammer on the single pipe or calculation of some positions on water pipe network, it have not been analysis for all of pipe on the water distribution systems. Simulation of water hammer due to closing valve on water distribution system and the influence level of pressure surge is evaluated at the defects on pipe. Water hammer on water supply pipe network are simulated by Water HAMMER software academic version and the capacity of defects are calculated by SINTAP. SINTAP developed from Brite-Euram projects in Brussels-Belgium with the aim to develop a process for assessing the integrity of the structure for the European industry. Based on the principle of mechanical fault, indicating the size of defects in materials affect the load capacity of the product in the course of work, the process has proposed setting up the diagram to fatigue assessment defect (FAD). The methods are applied for water pipe networks of Lien Chieu district, Da Nang city, Viet Nam, the results show the affected area of wave pressure by closing the valve and thereby assess the greatest pressure surge effect to corroded pipe. The SINTAP standard and finite element mesh analysis at the defect during the occurrence of pressure surge which will accurately assess the bearing capacity of the old pipes. This is one of the bases to predict the leakage locations on the water distribution systems. Amount of water hammer when identified on the water supply networks are decreasing due to local losses at the nodes as well as the friction with pipe wall, so this paper adequately simulate water hammer phenomena applying for actual water distribution systems. The research verified that pipe wall with defect is damaged under the pressure surge value.

  3. Experimental programmes related to high burnup fuel

    International Nuclear Information System (INIS)

    Vasudeva Rao, P.R.; Vidhya, R.; Ananthasivan, K.; Srinivasan, T.G.; Nagarajan, K.

    2002-01-01

    The experimental programmes undertaken at IGCAR with regard to high burn-up fuels fall under the following categories: a) studies on fuel behaviour, b) development of extractants for aqueous reprocessing and c) development of non-aqueous reprocessing techniques. An experimental programme to measure the carbon potential in U/Pu-FP-C systems by methane-hydrogen gas equilibration technique has been initiated at IGCAR in order to understand the evolution of fuel and fission product phases in carbide fuel at high burn-up. The carbon potentials in U-Mo-C system have been measured by this technique. The free energies and enthalpies of formation of LaC 2 , NdC 2 and SmC 2 have been measured by measuring the vapor pressures of CO over the region Ln 2 O 3 -LnC 2 -C during the carbothermic reduction of Ln 2 O 3 by C. The decontamination from fission products achieved in fuel reprocessing depends strongly on the actinide loading of the extractant phase. Tri-n-butyl phosphate (TBP), presently used as the extractant, does not allow high loadings due to its propensity for third phase formation in the extraction of Pu(IV). A detailed study of the allowable Pu loadings in TBP and other extractants has been undertaken in IGCAR, the results of which are presented in this paper. The paper also describes the status of our programme to develop a non-aqueous route for the reprocessing of fast reactor fuels. (author)

  4. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  5. Performance Evaluation of Pressure Transducers for Water Impacts

    Science.gov (United States)

    Vassilakos, Gregory J.; Stegall, David E.; Treadway, Sean

    2012-01-01

    The Orion Multi-Purpose Crew Vehicle is being designed for water landings. In order to benchmark the ability of engineering tools to predict water landing loads, test programs are underway for scale model and full-scale water impacts. These test programs are predicated on the reliable measurement of impact pressure histories. Tests have been performed with a variety of pressure transducers from various manufacturers. Both piezoelectric and piezoresistive devices have been tested. Effects such as thermal shock, pinching of the transducer head, and flushness of the transducer mounting have been studied. Data acquisition issues such as sampling rate and anti-aliasing filtering also have been studied. The response of pressure transducers have been compared side-by-side on an impulse test rig and on a 20-inch diameter hemisphere dropped into a pool of water. The results have identified a range of viable configurations for pressure measurement dependent on the objectives of the test program.

  6. Bridge pressure flow scour for clear water conditions

    Science.gov (United States)

    2009-10-01

    The equilibrium scour at a bridge caused by pressure flow with critical approach velocity in clear-water simulation conditions was studied both analytically and experimentally. The flume experiments revealed that (1) the measured equilibrium scour pr...

  7. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  8. Electrochemical noise measurements under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2000-01-01

    Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

  9. Design of a pressurized water loop heated by electric resistances

    International Nuclear Information System (INIS)

    Ribeiro, S.V.G.

    1981-01-01

    A pressurized water loop design is presented. Its operating pressure is 420 psi and we seek to simulate qualitatively some thermo-hydraulic phenomena of PWR reactors. The primary circuit simulator consists basically of two elements: 1)the test section housing 16 electric resistences dissipating a total power of 100 Kw; 2)the loop built of SCH40S 304L steel piping, consisting of the pump, a heat exchanger and the pressurizer. (Author) [pt

  10. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  11. On OH production in water containing atmospheric pressure plasmas

    NARCIS (Netherlands)

    Bruggeman, P.J.; Schram, D.C.

    2010-01-01

    In this paper radical production in atmospheric pressure water containing plasmas is discussed. As OH is often an important radical in these discharges the paper focuses on OH production. Besides nanosecond pulsed coronas and diffusive glow discharges, several other atmospheric pressure plasmas

  12. Complex cooling water systems optimization with pressure drop consideration

    CSIR Research Space (South Africa)

    Gololo, KV

    2012-12-01

    Full Text Available Pressure drop consideration has shown to be an essential requirement for the synthesis of a cooling water network where reuse/recycle philosophy is employed. This is due to an increased network pressure drop associated with additional reuse...

  13. Where Did the Water Go?: Boyle's Law and Pressurized Diaphragm Water Tanks

    Science.gov (United States)

    Brimhall, James; Naga, Sundar

    2007-01-01

    Many homes use pressurized diaphragm tanks for storage of water pumped from an underground well. These tanks are very carefully constructed to have separate internal chambers for the storage of water and for the air that provides the pressure. One might expect that the amount of water available for use from, for example, a 50-gallon tank would be…

  14. Experimental Study and Engineering Practice of Pressured Water Coupling Blasting

    Directory of Open Access Journals (Sweden)

    J. X. Yang

    2017-01-01

    Full Text Available Overburden strata movement in large space stope is the major reason that induces the appearance of strong mining pressure. Presplitting blasting for hard coal rocks is crucial for the prevention and control of strong pressure in stope. In this study, pressured water coupling blasting technique was proposed. The process and effect of blasting were analyzed by orthogonal test and field practice. Results showed that the presence of pressure-bearing water and explosive cartridges in the drill are the main influence factors of the blasting effect of cement test block. The high load-transmitting performance of pore water and energy accumulation in explosive cartridges were analyzed. Noxious substances produced during the blasting process were properly controlled because of the moistening, cooling, and diluting effect of pore water. Not only the goal of safe and static rock fragmentation by high-explosive detonation but also a combination of superdynamic blast loading and static loading effect of the pressured water was achieved. Then the practice of blasting control of hard coal rocks in Datong coal mine was analyzed to determine reasonable parameters of pressured water coupling blasting. A good presplitting blasting control effect was achieved for the hard coal rocks.

  15. Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

    2009-01-01

    An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

  16. Burnup determination of mass spectrometry for nuclear fuels

    International Nuclear Information System (INIS)

    Zhang Chunhua.

    1987-01-01

    The various methods currently being used in burnup determination of nuclear fuels are studied and reviewed. The mass spectrometry method of destructive testing is discussed emphatically. The burnup determination of mass spectrometry includes heavy isotopic abundance ratio method and isotope dilution mass spectrometry used as burnup indicator for the fission products. The former is applied to high burnup level, but the later to various burnup level. According to experiences, some problems which should be noticed in burnup determination of mass spectrometry are presented

  17. Pressure control for minimizing leakage in water distribution systems

    OpenAIRE

    Nourhan Samir; Rawya Kansoh; Walid Elbarki; Amr Fleifle

    2017-01-01

    In the last decades water resources availability has been a major issue on the international agenda. In a situation of worsening scarcity of water resources and the rapidly increasing of water demands, the state of water losses management is part of manâs survival on earth. Leakage in water supply networks makes up a significant amount, sometimes more than 70% of the total water losses. The best practices suggest that pressure management is one of the most effective way to reduce the amount o...

  18. Steam generator for pressurized-water reactors

    International Nuclear Information System (INIS)

    Michel, E.

    1971-01-01

    In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

  19. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Todosow, M.; Raitses, G.; Galperin, A.

    2009-01-01

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  20. The phase diagram of water at negative pressures: virtual ices.

    Science.gov (United States)

    Conde, M M; Vega, C; Tribello, G A; Slater, B

    2009-07-21

    The phase diagram of water at negative pressures as obtained from computer simulations for two models of water, TIP4P/2005 and TIP5P is presented. Several solid structures with lower densities than ice Ih, so-called virtual ices, were considered as possible candidates to occupy the negative pressure region of the phase diagram of water. In particular the empty hydrate structures sI, sII, and sH and another, recently proposed, low-density ice structure. The relative stabilities of these structures at 0 K was determined using empirical water potentials and density functional theory calculations. By performing free energy calculations and Gibbs-Duhem integration the phase diagram of TIP4P/2005 was determined at negative pressures. The empty hydrates sII and sH appear to be the stable solid phases of water at negative pressures. The phase boundary between ice Ih and sII clathrate occurs at moderate negative pressures, while at large negative pressures sH becomes the most stable phase. This behavior is in reasonable agreement with what is observed in density functional theory calculations.

  1. Pu-rich MOX agglomerate-by-agglomerate model for fuel pellet burnup analysis

    International Nuclear Information System (INIS)

    Chang, G.S.

    2004-01-01

    In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO 2 and 95% depleted UO 2 . Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO 2 concentration of 30.0 = (5 / 16.67 x 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of Agglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life, and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling. (author)

  2. Pressure: the politechnics of water supply in Mumbai.

    Science.gov (United States)

    Anand, Nikhil

    2011-01-01

    In Mumbai, most all residents are delivered their daily supply of water for a few hours every day, on a water supply schedule. Subject to a more precarious supply than the city's upper-class residents, the city's settlers have to consistently demand that their water come on “time” and with “pressure.” Taking pressure seriously as both a social and natural force, in this article I focus on the ways in which settlers mobilize the pressures of politics, pumps, and pipes to get water. I show how these practices not only allow settlers to live in the city, but also produce what I call hydraulic citizenship—a form of belonging to the city made by effective political and technical connections to the city's infrastructure. Yet, not all settlers are able to get water from the city water department. The outcomes of settlers' efforts to access water depend on a complex matrix of socionatural relations that settlers make with city engineers and their hydraulic infrastructure. I show how these arrangements describe and produce the cultural politics of water in Mumbai. By focusing on the ways in which residents in a predominantly Muslim settlement draw water despite the state's neglect, I conclude by pointing to the indeterminacy of water, and the ways in which its seepage and leakage make different kinds of politics and publics possible in the city.

  3. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development

  4. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  5. Steam plant for pressurized water reactors

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

  6. Tensile Strength of Water Exposed to Pressure Pulses

    DEFF Research Database (Denmark)

    Andersen, Anders Peter; Mørch, Knud Aage

    2012-01-01

    at an extended water-solid interface by imposing a tensile stress pulse which easily causes cavitation. Next, a compressive pulse of duration ~1 ms and a peak intensity of a few bar is imposed prior to the tensile stress pulse. A dramatic increase of the tensile strength is observed immediately after......It is well known that pressurization for an extended period of time increases the tensile strength of water, but little information is available on the effect of pressure pulses of short duration. This is addressed in the present paper where we first measure the tensile strength of water...

  7. Is high-pressure water the cradle of life?

    International Nuclear Information System (INIS)

    Bassez, Marie-Paule

    2003-01-01

    Several theories have been proposed for the synthesis of prebiotic molecules. This letter shows that the structure of supercritical water, or high-pressure water, could trigger prebiotic synthesis and the origin of life deep in the oceans, in hydrothermal vent systems. Dimer geometries of high-pressure water may have a point of symmetry and a zero dipole moment. Consequently, simple apolar molecules found in submarine hydrothermal vent systems will dissolve in the apolar environment provided by the apolar form of the water dimer. Apolar water could be the medium which helps precursor molecules to concentrate and react more efficiently. The formation of prebiotic molecules could thus be linked to the structure of the water inside chimney nanochannels and cavities where hydrothermal piezochemistry and shock wave chemistry could occur. (letter to the editor)

  8. Preferential removal of Sm by evaporation from Nd-Sm mixture and its application in direct burn-up determination of spent nuclear fuel

    International Nuclear Information System (INIS)

    Sajimol, R.; Bera, S.; Nalini, S.; Sivaraman, N.; Joseph, M.; Kumar, T.

    2016-01-01

    Rate of evaporation of Sm and Nd from their mixture was studied based on their ion intensities using thermal ionization mass spectrometry. Because of the comparatively larger evaporation rate of Sm, it was found possible to get the isotopic composition of Nd (fission product monitor) free from isobaric interference of Sm isotopes. The decrease in ion intensity of Sm was studied as a function of time and filament temperature. Based on this study, an easy and time effective method for the determination of burn-up of spent nuclear fuel was examined and the results are compared with that obtained by the conventional method. Typical burn-up value obtained for a pressurized heavy water reactor fuel dissolver solution using the direct method by preferential evaporation of Sm is: 0.84 at.%, whereas the one obtained by the use of conventional method is 0.82 at.%. In both the cases, Nd was employed as the fission product monitor. (author)

  9. Contribution of water vapor pressure to pressurization of plutonium dioxide storage containers

    Science.gov (United States)

    Veirs, D. Kirk; Morris, John S.; Spearing, Dane R.

    2000-07-01

    Pressurization of long-term storage containers filled with materials meeting the US DOE storage standard is of concern.1,2 For example, temperatures within storage containers packaged according to the standard and contained in 9975 shipping packages that are stored in full view of the sun can reach internal temperatures of 250 °C.3 Twenty five grams of water (0.5 wt.%) at 250 °C in the storage container with no other material present would result in a pressure of 412 psia, which is limited by the amount of water. The pressure due to the water can be substantially reduced due to interactions with the stored material. Studies of the adsorption of water by PuO2 and surface interactions of water with PuO2 show that adsorption of 0.5 wt.% of water is feasible under many conditions and probable under high humidity conditions.4,5,6 However, no data are available on the vapor pressure of water over plutonium dioxide containing materials that have been exposed to water.

  10. Response of steam-water mixtures to pressure transients

    International Nuclear Information System (INIS)

    Hull, L.M.

    1985-01-01

    During the transition phase of a hypothetical core-disruptive accident in a liquid-metal fast breeder reactor, melting fuel-steel mixtures may begin to boil, resulting in a two-phase mixture of molten reactor fuel and steel vapor. Dispersal of this mixture by pressure transients may prevent recriticality of the fuel material. This paper describes the results of a series of experiments that investigated the response of two-phase mixtures to pressure transients. Simulant fluids (steam/water) were used in a transparent 10.2-cm-dia, 63.5-cm-long acrylic tube. The pressure transient was provided by releasing pressurized nitrogen from a supply tank. The data obtained are in the form of pressure-time records and high-speed movies. The varied parameters are initial void fraction (10% and 40%) and transient pressure magnitude (3.45 and 310 kPa)

  11. Decay ratio estimation in pressurized water reactor

    International Nuclear Information System (INIS)

    Por, G.; Runkel, J.

    1990-11-01

    The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

  12. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  13. Water Pressure Distribution on a Twin-Float Seaplane

    Science.gov (United States)

    Thompson, F L

    1930-01-01

    This is the second of a series of investigations to determine water pressure distribution on various types of seaplane floats and hulls, and was conducted on a twin-float seaplane. It consisted of measuring water pressures and accelerations on a TS-1 seaplane during numerous landing and taxiing maneuvers at various speeds and angles. The results show that water pressures as great as 10 lbs. per sq. in.may occur at the step in various maneuvers and that pressures of approximately the same magnitude occur at the stern and near the bow in hard pancake landings with the stern way down. At the other parts of the float the pressures are less and are usually zero or slightly negative for some distance abaft the step. A maximum negative pressure of 0.87 lb. Per square inch was measured immediately abaft the step. The maximum positive pressures have a duration of approximately one-twentieth to one-hundredth second at any given location and are distributed over a very limited area at any particular instant.

  14. Water cycles in closed ecological systems: effects of atmospheric pressure.

    Science.gov (United States)

    Rygalov, Vadim Y; Fowler, Philip A; Metz, Joannah M; Wheeler, Raymond M; Bucklin, Ray A

    2002-01-01

    In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from ~1 to 10 L m-2 d-1 (~1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

  15. Water cycles in closed ecological systems: effects of atmospheric pressure

    Science.gov (United States)

    Rygalov, Vadim Y.; Fowler, Philip A.; Metz, Joannah M.; Wheeler, Raymond M.; Bucklin, Ray A.; Sager, J. C. (Principal Investigator)

    2002-01-01

    In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from 1 to 10 L m-2 d-1 (1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

  16. Value of burnup credit beyond actinides

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.; Kang, Chi.

    1997-01-01

    DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs

  17. Evaluating the Laplace pressure of water nanodroplets from simulations

    Science.gov (United States)

    Malek, Shahrazad M. A.; Sciortino, Francesco; Poole, Peter H.; Saika-Voivod, Ivan

    2018-04-01

    We calculate the components of the microscopic pressure tensor as a function of radial distance r from the centre of a spherical water droplet, modelled using the TIP4P/2005 potential. To do so, we modify a coarse-graining method for calculating the microscopic pressure (Ikeshoji et al 2003 Mol. Simul. 29 101) in order to apply it to a rigid molecular model of water. As test cases, we study nanodroplets ranging in size from 776 to 2880 molecules at 220 K. Beneath a surface region comprising approximately two molecular layers, the pressure tensor becomes approximately isotropic and constant with r. We find that the dependence of the pressure on droplet radius is that expected from the Young-Laplace equation, despite the small size of the droplets.

  18. Noise investigations in boiling water and pressurized water reactors

    International Nuclear Information System (INIS)

    Kosaly, G.

    1979-08-01

    The paper begins with the general discussion of the neutron noise measured by in-core and ex-core detectors. The theory of BWR noise is surveyed in detail. After the theoretical survey the physical background of the measuring methods applied in the fluctuation analysis of BWRs is discussed. The section on PWR noise discusses the information content of the neutron, pressure and displacement signals about the vibration of the vessel and its internals. (author)

  19. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  20. Fission gas release and fuel rod chemistry related to extended burnup

    International Nuclear Information System (INIS)

    1993-04-01

    The purpose of the meeting was to review the state of the art in fission gas release and fuel rod chemistry related to extended burnup. The meeting was held in a time when national and international programmes on water reactor fuel irradiated in experimental reactors were still ongoing or had reached their conclusion, and when lead test assemblies had reached high burnup in power reactors and been examined. At the same time, several out-of-pile experiments on high burnup fuel or with simulated fuel were being carried out. As a result, significant progress has been registered since the last meeting, particularly in the evaluation of fuel temperature, the degradation of the global thermal conductivity with burnup and in the understanding of the impact on fission gas release. Fifty five participants from 16 countries and one international organization attended the meeting. 28 papers were presented. A separate abstract was prepared for each of the papers. Refs, figs, tabs and photos

  1. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    Leroy, C.; Bitsch, D.; Millot, J.P.

    1985-10-01

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  2. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-01-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

  3. Cavitation nuclei in water exposed to transient pressures

    DEFF Research Database (Denmark)

    Andersen, Anders Peter; Mørch, Knud Aage

    2015-01-01

    A model of skin-stabilized interfacial cavitation nuclei and their response to tensile and compressive stressing is presented. The model is evaluated in relation to experimental tensile strength results for water at rest at the bottom of an open water-filled container at atmospheric pressure...... and room temperature. These results are obtained by recording the initial growth of cavities generated by a short tensile pulse applied to the bottom of the container. It is found that the cavitation nuclei shift their tensile strength depending on their pressure history. Static pressurization...... for an extended period of time prior to testing is known to increase the tensile strength of water, but little information is available on how it is affected by compression pulses of short duration. This is addressed by imposing compression pulses of approximately 1 ms duration and a peak intensity of a few bar...

  4. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  5. Microwave measurements of water vapor partial pressure at high temperatures

    International Nuclear Information System (INIS)

    Latorre, V.R.

    1991-01-01

    One of the desired parameters in the Yucca Mountain Project is the capillary pressure of the rock comprising the repository. This parameter is related to the partial pressure of water vapor in the air when in equilibrium with the rock mass. Although there are a number of devices that will measure the relative humidity (directly related to the water vapor partial pressure), they generally will fail at temperatures on the order of 150C. Since thee author has observed borehole temperatures considerably in excess of this value in G-Tunnel at the Nevada Test Site (NTS), a different scheme is required to obtain the desired partial pressure data at higher temperatures. This chapter presents a microwave technique that has been developed to measure water vapor partial pressure in boreholes at temperatures up to 250C. The heart of the system is a microwave coaxial resonator whose resonant frequency is inversely proportional to the square root of the real part of the complex dielectric constant of the medium (air) filling the resonator. The real part of the dielectric constant of air is approximately equal to the square of the refractive index which, in turn, is proportional to the partial pressure of the water vapor in the air. Thus, a microwave resonant cavity can be used to measure changes in the relative humidity or partial pressure of water vapor in the air. Since this type of device is constructed of metal, it is able to withstand very high temperatures. The actual limitation is the temperature limit of the dielectric material in the cable connecting the resonator to its driving and monitoring equipment-an automatic network analyzer in our case. In the following sections, the theory of operation, design, construction, calibration and installation of the microwave diagnostics system is presented. The results and conclusions are also presented, along with suggestions for future work

  6. Burnup verification using the FORK measurement system

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

    Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK measurement system, designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program, has been used to verify reactor site records for burnup and cooling time for many years. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. This report deals with the application of the FORK system to burnup credit operations based on measurements performed on spent fuel assemblies at the Oconee Nuclear Station of Duke Power Company

  7. Configuration of LWR fuel enrichment or burnup yielding maximum power

    International Nuclear Information System (INIS)

    Bartosek, V.; Zalesky, K.

    1976-01-01

    An analysis is given of the spatial distribution of fuel burnup and enrichment in a light-water lattice of given dimensions with slightly enriched uranium, at which the maximum output is achieved. It is based on the spatial solution of neutron flux using a one-group diffusion model in which linear dependence may be expected of the fission cross section and the material buckling parameter on the fuel burnup and enrichment. Two problem constraints are considered, i.e., the neutron flux value and the specific output value. For the former the optimum core configuration remains qualitatively unchanged for any reflector thickness, for the latter the cases of a reactor with and without reflector must be distinguished. (Z.M.)

  8. A Neural-Network-Based Nonlinear Adaptive State-Observer for Pressurized Water Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2013-10-01

    Full Text Available Although there have been some severe nuclear accidents such as Three Mile Island (USA, Chernobyl (Ukraine and Fukushima (Japan, nuclear fission energy is still a source of clean energy that can substitute for fossil fuels in a centralized way and in a great amount with commercial availability and economic competitiveness. Since the pressurized water reactor (PWR is the most widely used nuclear fission reactor, its safe, stable and efficient operation is meaningful to the current rebirth of the nuclear fission energy industry. Power-level regulation is an important technique which can deeply affect the operation stability and efficiency of PWRs. Compared with the classical power-level controllers, the advanced power-level regulators could strengthen both the closed-loop stability and control performance by feeding back the internal state-variables. However, not all of the internal state variables of a PWR can be obtained directly by measurements. To implement advanced PWR power-level control law, it is necessary to develop a state-observer to reconstruct the unmeasurable state-variables. Since a PWR is naturally a complex nonlinear system with parameters varying with power-level, fuel burnup, xenon isotope production, control rod worth and etc., it is meaningful to design a nonlinear observer for the PWR with adaptability to system uncertainties. Due to this and the strong learning capability of the multi-layer perceptron (MLP neural network, an MLP-based nonlinear adaptive observer is given for PWRs. Based upon Lyapunov stability theory, it is proved theoretically that this newly-built observer can provide bounded and convergent state-observation. This observer is then applied to the state-observation of a special PWR, i.e., the nuclear heating reactor (NHR, and numerical simulation results not only verify its feasibility but also give the relationship between the observation performance and observer parameters.

  9. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    International Nuclear Information System (INIS)

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting

  10. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  11. Experiments on aerosol removal by high-pressure water spray

    Energy Technology Data Exchange (ETDEWEB)

    Corno, Ada del, E-mail: delcorno@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Morandi, Sonia, E-mail: morandi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Parozzi, Flavio, E-mail: parozzi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Araneo, Lucio, E-mail: lucio.araneo@polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy); CNR-IENI, via Cozzi 53, I-20125 Milano (Italy); Casella, Francesco, E-mail: francesco2.casella@mail.polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy)

    2017-01-15

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m{sup 3}. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m{sup 3}. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was

  12. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  13. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  14. Supercritical water gasification with decoupled pressure and heat transfer modules

    KAUST Repository

    Dibble, Robert

    2017-09-14

    The present invention discloses a system and method for supercritical water gasification (SCWG) of biomass materials wherein the system includes a SCWG reactor and a plurality of heat exchangers located within a shared pressurized vessel, which decouples the function of containing high pressure from the high temperature function. The present invention allows the heat transfer function to be conducted independently from the pressure transfer function such that the system equipment can be designed and fabricated in manner that would support commercial scaled-up SCWG operations. By using heat exchangers coupled to the reactor in a series configuration, significant efficiencies are achieved by the present invention SCWG system over prior known SCWG systems.

  15. Experiments on aerosol removal by high-pressure water spray

    International Nuclear Information System (INIS)

    Corno, Ada del; Morandi, Sonia; Parozzi, Flavio; Araneo, Lucio; Casella, Francesco

    2017-01-01

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m"3. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m"3. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was detected with 1

  16. Methodology for surge pressure evaluation in a water injection system

    Energy Technology Data Exchange (ETDEWEB)

    Meliande, Patricia; Nascimento, Elson A. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Civil; Mascarenhas, Flavio C.B. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Lab. de Hidraulica Computacional; Dandoulakis, Joao P. [SHELL of Brazil, Rio de Janeiro, RJ (Brazil)

    2009-07-01

    Predicting transient effects, known as surge pressures, is of high importance for offshore industry. It involves detailed computer modeling that attempts to simulate the complex interaction between flow line and fluid in order to ensure efficient system integrity. Platform process operators normally raise concerns whether the water injection system is adequately designed or not to be protected against possible surge pressures during sudden valve closure. This report aims to evaluate the surge pressures in Bijupira and Salema water injection systems due to valve closure, through a computer model simulation. Comparisons among the results from empirical formulations are discussed and supplementary analysis for Salema system were performed in order to define the maximum volumetric flow rate for which the design pressure was able to withstand. Maximum surge pressure values of 287.76 bar and 318.58 bar, obtained in Salema and Bijupira respectively, using empirical formulations have surpassed the operating pressure design, while the computer model results have pointed the greatest surge pressure value of 282 bar in Salema system. (author)

  17. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  18. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  19. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  20. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  1. Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

    International Nuclear Information System (INIS)

    Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

    1994-01-01

    This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

  2. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-01-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  3. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  4. Nonlinear vibration of a hemispherical dome under external water pressure

    International Nuclear Information System (INIS)

    Ross, C T F; McLennan, A; Little, A P F

    2011-01-01

    The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

  5. Nonlinear vibration of a hemispherical dome under external water pressure

    Science.gov (United States)

    Ross, C. T. F.; McLennan, A.; Little, A. P. F.

    2011-07-01

    The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

  6. Improved water density feedback model for pressurized water reactors

    International Nuclear Information System (INIS)

    Casadei, A.L.

    1976-01-01

    An improved water density feedback model has been developed for neutron diffusion calculations of PWR cores. This work addresses spectral effects on few-group cross sections due to water density changes, and water density predictions considering open channel and subcooled boiling effects. An homogenized spectral model was also derived using the unit assembly diffusion method for employment in a coarse mesh 3D diffusion computer program. The spectral and water density evaluation models described were incorporated in a 3D diffusion code, and neutronic calculations for a typical PWR were completed for both nominal and accident conditions. Comparison of neutronic calculations employing the open versus the closed channel model for accident conditions indicates that significant safety margin increases can be obtained if subcooled boiling and open channel effects are considered in accident calculations. This is attributed to effects on both core reactivity and power distribution, which result in increased margin to fuel degradation limits. For nominal operating conditions, negligible differences in core reactivity and power distribution exist since flow redistribution and subcooled voids are not significant at such conditions. The results serve to confirm the conservatism of currently employed closed channel feedback methods in accident analysis, and indicate that the model developed in this work can contribute to show increased safety margins for certain accidents

  7. An evaluation of the influence of fuel design parameters and burnup on pellet/cladding interaction for boiling water reactor fuel rod through in-core diameter measurement

    International Nuclear Information System (INIS)

    Yanagisawa, K.

    1986-01-01

    The influence of design parameters and burning on pellet/cladding interaction (PCI) of current boiling water reactor fuel rods was studied through in-core diameter measurement. Thinner cladding and a smaller diametral gap enhanced the PCI during startup. At constant power, fuel with SiO 2 added greatly reduced PCI due to relaxation. The fuel with a small grain size greatly reduced PCI due to densification. Preirradiation of rods up to 23 MWd/kgU caused a large PCI not only in a small gap but also in a large gap rod. Relaxation and permanent deformation was small. In the power increase experiment, one rod experienced PCI failure. The spurt times of coolant radioactivity coincided well with the sudden drop of cladding axial strain and marked crack opening at the rod surface. The estimated hoop stress predicted by FEMAXI-III was 350 MPa at the failure

  8. Pressure wave propagation in the discharge piping with water pool

    International Nuclear Information System (INIS)

    Bang, Young S.; Seul, Kwang W.; Kim, In Goo

    2004-01-01

    Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined

  9. Flashing of high-pressure saturated water into the pool water

    International Nuclear Information System (INIS)

    Takamasa, Tomoji; Kondo, Koichi; Aya, Izuo.

    1997-01-01

    This paper presents an experimental study on a saturated high-pressure water discharging into a water pool. The purpose of the experiment is to clarify the phenomena that occur by a blow-down of the water from the pressure vessel into the water-filled containment in the case of a wall-crack accident or a LOCA in a passive safety reactor. The results show that a flashing oscillation (FO) occurs when the water discharges into the pool, under specified experimental conditions. The range of the flashing location oscillates between a point very close to and some distance away from the vent hole. The pressures in the vent tube and water pool constantly fluctuate due to the flashing oscillation. The pressure oscillation and alternating flashing location might be caused by the balancing action between the supply of saturated water, flashing at the control volume and steam condensation on the steam-water interface. The frequencies of FO, or frequencies of pressure oscillation and alternating flashing location, increased as water subcooling increased, and as discharging pressure and vent hole diameter decreased. A linear analysis was conducted using a spherical flashing bubble model in which the motion of bubble is controlled by steam condensation. The effects of these parameters on the period of FO in the experiments can be predicted well by the analysis. (author)

  10. Threshold burnup for recrystallization and model for rim porosity in the high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong

    1998-01-01

    Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75μm and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup

  11. Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2009-07-01

    The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

  12. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  13. Physical models for high burnup fuel

    International Nuclear Information System (INIS)

    Kanyukova, V.; Khoruzhii, O.; Likhanskii, V.; Solodovnikov, G.; Sorokin, A.

    2003-01-01

    In this paper some models of processes in high burnup fuel developed in Src of Russia Troitsk Institute for Innovation and Fusion Research are presented. The emphasis is on the description of the degradation of the fuel heat conductivity, radial profiles of the burnup and the plutonium accumulation, restructuring of the pellet rim, mechanical pellet-cladding interaction. The results demonstrate the possibility of rather accurate description of the behaviour of the fuel of high burnup on the base of simplified models in frame of the fuel performance code if the models are physically ground. The development of such models requires the performance of the detailed physical analysis to serve as a test for a correct choice of allowable simplifications. This approach was applied in the SRC of Russia TRINITI to develop a set of models for the WWER fuel resulting in high reliability of predictions in simulation of the high burnup fuel

  14. Burnup analysis of the power reactor, 2

    International Nuclear Information System (INIS)

    Ezure, Hideo

    1975-09-01

    In burnup analysis of JPDR-1 with FLARE, it was found to have problems. The program FLORA was developed for solution of the problems. By their bench mark tests FLORA was found to be useful for three-dimensional thermal-hydro-dynamic analysis of BWRs. It was applied to analysis of the burnup of JPDR-1. The input data and option of FLORA were corrected on referring to the results of gammer probe tests for JPDR-1. The void, source and burnup distributions were calculated each month during the operation. The burnup distribution in three assemblies revealed by a destructive test agrees better with that by FLORA than by FLARE. It was shown that the distortion of power distribution around the control rods by FLORA was smaller and closer to that by the gammer probe tests than by FLARE, and the connector of fuel assemblies and the plugs in the reflector had much influence on the power distribution. (auth.)

  15. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

  16. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)

  17. Ultra-high pressure water jet: Baseline report

    International Nuclear Information System (INIS)

    1997-01-01

    The ultra-high pressure waterjet technology was being evaluated at Florida International University (FIU) as a baseline technology. In conjunction with FIU's evaluation of efficiency and cost, this report covers the evaluation conducted for safety and health issues. It is a commercially available technology and has been used for various projects at locations throughout the country. The ultra-high pressure waterjet technology acts as a cutting tool for the removal of surface substrates. The Husky trademark pump feeds water to a lance that directs the high pressure water at the surface to be removed. The safety and health evaluation during the testing demonstration focused on two main areas of exposure. These were dust and noise. The dust exposure was found to be minimal, which would be expected due to the wet environment inherent in the technology, but noise exposure was at a significant level. Further testing for noise is recommended because of the outdoor environment where the testing demonstration took place. In addition, other areas of concern found were arm-hand vibration, ergonomics, heat stress, tripping hazards, electrical hazards, lockout/tagout, fall hazards, slipping hazards, hazards associated with the high pressure water, and hazards associated with air pressure systems

  18. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Science.gov (United States)

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  19. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2008-01-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

  20. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  1. Variations of free gas content in water during pressure fluctuations

    International Nuclear Information System (INIS)

    Keller, A.; Zielke, W.

    1977-01-01

    In this paper an experimental programme is described in order to determine the influence of the cavitation nuclei distribution on cavitation inception. This programme has been used to measure air bubbles dimensions and number and particularly to determine the influence of quick pressure variations on the size on the number of bubbles in a pipe. An optical device counting scattered light is used as a measuring technique. Gas bubbles go through an optical control volume where they receive a high intensity light beam and scatter the light, then led to a photomultiplier; the signals are sorted and counted according to their size. If the number of nuclei, the dimensions of the control volume and the velocity of the water are known, it is possible to determine bubbles concentrations and the bulk modulus of the water. This measuring technique has been applied to a flow in a 140 mm diameter pipe with quick pressure variations from 2 bar to 0-10 bar. During the variations, the void fraction depends on the Reynolds number of the flow and on the gas content of the water. The bulk modulus has been computed with different conditions. Most results concern pressures slightly over the vapor pressure. Air content has a strong influence on cavitation and on water compressibility after a vapor cavity collapse

  2. 'CANDLE' burnup regime after LWR regime

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nagata, Akito

    2008-01-01

    CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

  3. Absorber rod bundle actuator in a pressurized water nuclear reactor

    International Nuclear Information System (INIS)

    Martin, J.; Peletan, R.

    1984-01-01

    The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

  4. Development test procedure High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    Development testing will be performed on the water jet cleaning fixture to determine the most effective arrangement of water jet nozzles to remove contamination from the surfaces of canisters and other debris. The following debris may be stained with dye to simulate surface contaminates: Mark O, Mark I, and Mark II Fuel Storage Canisters (both stainless steel and aluminum), pipe of various size, (steel, stainless, carbon steel and aluminum). Carbon steel and stainless steel plate, channel, angle, I-beam and other surfaces, specifically based on the Scientific Ecology Group (SEG) inventory and observations of debris within the basin. Test procedure for developmental testing of High Pressure Water Jet System

  5. Calculation of pellet radial power distributions with a Monte Carlo burnup code

    International Nuclear Information System (INIS)

    Suzuki, Motomu; Yamamoto, Toru; Nakata, Tetsuo

    2010-01-01

    The Japan Nuclear Energy Safety Organization (JNES) has been working on an irradiation test program of high-burnup MOX fuel at Halden Boiling Water Reactor (HBWR). MOX and UO 2 fuel rods had been irradiated up to about 64 GWd/t (rod avg.) as a Japanese utilities research program (1st phase), and using those fuel rods, in-situ measurement of fuel pellet centerline temperature was done during the 2nd phase of irradiation as the JNES test program. As part of analysis of the temperature data, power distributions in a pellet radial direction were analyzed by using a Monte Carlo burnup code MVP-BURN. In addition, the calculated results of deterministic burnup codes SRAC and PLUTON for the same problem were compared with those of MVP-BURN to evaluate their accuracy. Burnup calculations with an assembly model were performed by using MVP-BURN and those with a pin cell model by using SRAC and PLUTON. The cell pitch and, therefore, fuel to moderator ratio in the pin cell calculation was determined from the comparison of neutron energy spectra with those of MVP-BURN. The fuel pellet radial distributions of burnup and fission reaction rates at the end of the 1st phase irradiation were compared between the three codes. The MVP-BURN calculation results show a large peaking in the burnup and fission rates in the pellet outer region for the UO 2 and MOX pellets. The SRAC calculations give very close results to those of the MVP-BURN. On the other hand, the PLUTON calculations show larger burnup for the UO 2 and lower burnup for the MOX pellets in the pellet outer region than those of MVP-BURN, which lead to larger fission rates for the UO 2 and lower fission rates for the MOX pellets, respectively. (author)

  6. Comparative study on plutonium and MA recycling in equilibrium burnup and standard burnup of PWR

    International Nuclear Information System (INIS)

    Waris, Abdul; Kurniadi, Rizal; Su'ud, Zaki; Permana, Sidik

    2005-01-01

    The equilibrium burnup model is a powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor. However, this method needs to be verified since the method is not a standard tool. The present study aimed to compare the characteristics of plutonium recycling and plutonium and minor actinides (MA) recycling in PWR with the equilibrium burnup and the standard burnup. In order to become more comprehensive study, an influence of moderator-to-fuel volume ratio (MFR) changes by changing the pin-pitch of fuel cell has been evaluated. The MFR ranges from 0.5 to 4.0. For the equilibrium burnup we used equilibrium cell-burnup code. We have employed 1368 nuclides in the equilibrium calculation with 129 of them are heavy metals (HMs). For standard burnup, SRAC2002 code has been utilized with 26 HMs and 66 fission products (FPs). The JENDL 3.2 library has been employed for both burnup schemes. The uranium, plutonium and MA vector, which resulted from the equilibrium burnup are directly used as fuel input composition for the standard burnup calculation. Both burnup results demonstrate that plutonium recycling and plutonium and MA recycling can be conducted safer in tight lattice core. They are also show the similar trend in neutron spectrum, which become harder with the increasing number of recycled heavy nuclides as well as the decreasing of the MFR values. However, there are some discrepancy on the effective multiplication factor and the conversion ratio, especially for the reactor core for MFR ≥ 2.0. (author)

  7. Validation of SCALE-4 for burnup credit applications

    International Nuclear Information System (INIS)

    Bowman, S.M.; DeHart, M.D.; Parks, C.V.

    1995-01-01

    In the past, a criticality analysis of PWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at ORNL in support of the US DOE efforts to demonstrate a validation approach for criticality safety methods to be used in burnup credit cask design. To date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The ANSI/ANS-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors by correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

  8. Structural analysis of fuel rod applied to pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Danilo P.; Pinheiro, Andre Ricardo M.; Lotto, André A., E-mail: danilo.pinheiro@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The design of fuel assemblies applied to Pressurized Water Reactors (PWR) has several requirements and acceptance criteria that must be attended for licensing. In the case of PWR fuel rods, an important mechanical structural requirement is to keep the radial stability when submitted to the coolant external pressure. In the framework of the Accident Tolerant Fuel (ATF) program new materials have been studied to replace zirconium based alloys as cladding, including iron-based alloys. In this sense, efforts have been made to evaluate the behavior of these materials under PWR conditions. The present work aims to evaluate the collapse cold pressure of a stainless steel thin-walled tube similar to that used as cladding material of fuel rods by means of the comparison of numeric data, and experimental results. As a result of the simulations, it was observed that the collapse pressure has a value intermediate value between those found by regulatory requirements and analytical calculations. The experiment was carried out for the validation of the computational model using test specimens of thin-walled tubes considering empty tube. The test specimens were sealed at both ends by means of welding. They were subjected to a high pressure device until the collapse of the tubes. Preliminary results obtained from experiments with the empty test specimens indicate that the computational model can be validated for stainless steel cladding, considering the difference between collapse pressure indicated in the regulatory document and the actual limit pressure concerning to radial instability of tubes with the studied characteristics. (author)

  9. Technical Issues in the development of high burnup and long cycle fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  10. Technical Issues in the development of high burnup and long cycle fuel pellets

    International Nuclear Information System (INIS)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui

    2012-01-01

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  11. Reuse of waste water from high pressure water jet decontamination for reactor decommissioning scrap metal

    International Nuclear Information System (INIS)

    Deng Junxian; Li Xin; Hou Huijuan

    2011-01-01

    For recycle and reuse of reactor decommissioning scrap metal by high pressure water jet decontamination, large quantity of radioactive waste water will be generated. To save the cost of radioactive waste water treatment and to reduce the cost of the scrap decontamination, this part of radioactive waste water should be reused. Most of the radioactivities in the decontamination waste water come from the solid particle in the water. Thus to reuse the waste water, the solid particle in the waster should be removed. Different possible treatment technologies have been investigated. By cost benefit analysis the centrifugal separation technology is selected. (authors)

  12. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

  13. Water pressure control device for control rod drive

    International Nuclear Information System (INIS)

    Sato, Hideyuki.

    1981-01-01

    Purpose: To minimize the fluctuations in the reactor water level upon occurrence of abnormality by inputting the level signal of the reactor to an arithmetic unit for controlling the pressure of control rod drive water to thereby enable effective reactor level control. Constitution: Signal from a flow rate transmitter is inputted into an arithmetic unit to perform constant flow rate control upon normal operation. While on the other hand, if abnormality occurs such as feedwater pump trips, the arithmetic unit is switched from the constant flow rate control to the reactor water level control. Reactor water level signal is inputted into the arithmetic unit and the control valve is most suitably controlled, whereby water is fed from CST to the reactor by way of control rod drive water system to secure the reactor water level if feedwater to the reactor is interrupted by loss of coolants on the feedwater system. Since this enables to minimize the fluctuations in the reactor water level upon abnormality, the reactor water level can be controlled most suitably by the reactor water level signal. (Moriyama, K.)

  14. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    International Nuclear Information System (INIS)

    Leal, Luiz C; Derrien, Herve; Dunn, Michael E; Mueller, Don

    2007-01-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  15. Automated corrosion fatigue crack growth testing in pressurized water environments

    International Nuclear Information System (INIS)

    Ceschini, L.J.; Liaw, P.K.; Rudd, G.E.; Logsdon, W.A.

    1984-01-01

    This paper describes in detail a novel approach to construct a test facility for developing corrosion fatigue crack growth rate (FCGR) properties in aggressive environments. The environment studied is that of a pressurized water reactor (PWR) at 288 0 C (550 0 F) and 13.8 MPa (200 psig). To expedite data generation, each chamber was designed to accommodate two test specimens. A common water recirculation and pressurization system was employed to service two test chambers. Thus, four fatigue crack propagation rate tests could be conducted simultaneously in the pressurized water environment. The data analysis was automated to minimize the typically high labor costs associated with corrosion fatigue crack propagation testing. Verification FCGR tests conducted on an ASTM A469 rotor steel in a room temperature air environment as well as actual PWR environment FCGR tests performed on an ASTM A533 Grade B Class 2 pressure vessel steel demonstrated that the dual specimen test facility is an excellent system for developing the FCGR properties of materials in adverse environments

  16. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  17. Orange oil/water nanoemulsions prepared by high pressure homogenizer

    International Nuclear Information System (INIS)

    Kourniatis, Loretta R.; Spinelli, Luciana S.; Mansur, Claudia R.E.

    2010-01-01

    The objective of this work was to use the high-pressure homogenizer (HPH) to prepare stable oil/water nanoemulsions presenting narrow particle size distribution. The dispersions were prepared using nonionic surfactants based on ethoxylated ether. The size and distribution of the droplets formed, along with their stability, were determined in a Zetasizer Nano ZS particle size analyzer. The stability and the droplet size distribution in these systems do not present the significant differences with the increase of the processing pressure in the HPH). The processing time can promote the biggest dispersion in the size of particles, thus reducing its stability. (author)

  18. Swelling pressure and water absorption property of compacted granular bentonite during water absorption

    International Nuclear Information System (INIS)

    Oyamada, T.; Komine, H.; Murakami, S.; Sekiguchi, T.; Sekine, I.

    2012-01-01

    Document available in extended abstract form only. Bentonite is currently planned to be used as buffer materials in engineered barrier of radioactive waste disposal. Granular bentonites are expected as the materials used in constructions as buffer materials by in-situ compaction methods. After applying these buffer materials, it is expected that the condition of the buffer area changes in long-term by the seepage of groundwater into buffer area. Therefore, it is important to understand water movement and swelling behavior of the buffer materials for evaluating the performance of engineered barrier. In this study, we investigated water absorption property and swelling pressure of compacted granular bentonite. Specifically, the process of swelling pressure and amount of water absorption of granular bentonite-GX (Kunigel-GX, produced at the Tsukinuno mine in Japan) were observed by laboratory tests. To discuss the influence of maximum grain size of bentonite particle on swelling pressure and water absorption property, two types of samples were used. One is granular sample which is Bentonite-GX controlled under 2 mm the maximum grain size, the other is milled sample which is Bentonite-GX with the maximum grain size under 0.18 mm by milling with the agate mortar. In addition, the mechanism on the swelling pressure of compacted granular bentonite was considered and discussed. In the cases of granular sample, swelling pressure increases rapidly, then gradually continues to increase up to maximum value. In the cases of milled sample, swelling pressure also increases rapidly at first. However, then its value decreases before progressing of gradual increase continues. Especially, this trend was clearly observed at a relatively low dry density. At the peaks of these curves, the swelling pressure of granular samples is lower than that of milled samples. In addition, the increasing of swelling pressure by the time the peak observed during the process of swelling pressure from

  19. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  20. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    International Nuclear Information System (INIS)

    Barkauskas, V.; Plukiene, R.; Plukis, A.

    2016-01-01

    Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

  1. The effect of pressurizer-water-level on the low frequency component of the pressure spectrum in a PWR

    International Nuclear Information System (INIS)

    Por, G.; Izsak, E.; Valko, J.

    1984-09-01

    The pressure fluctuations were measured in the cooling system of the Paks-1 reactor. A shift of the peak was detected in low frequency component of the pressure fluctuation spectrum which is due to the fluctuations of water level in the pressurizer. Using the model of Katona and Nagy (1983), the eigenfrequencies, the magnitude of the shift and the sensitivity to the pressurizer water level were reproduced in good accordance with the experimental data. (D.Gy.)

  2. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    International Nuclear Information System (INIS)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  3. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  4. Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

    2016-08-15

    Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

  5. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  6. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

  7. The study and improvement of water level control of pressurizer

    International Nuclear Information System (INIS)

    Gao Peng; Zhang Qinshun

    2006-01-01

    The PI controller which is used widely in water level control of pressurizer in reactor control system usually leads dynamic overshoot and long setting time. The improvement project for intelligent fuzzy controller to take the place of PI controller is advanced. This paper researches the water level control of pressurizer in reactor control system of Daya Bay Phase I, and describes the method of intelligent fuzzy control in practice. Simulation indicates that the fuzzy control has advantages of small overshoot and short settling time. It can also improve control system's real time property and anti-interference ability. Especially for non-linear and time-varying complicated control systems, it can obtain good control results. (authors)

  8. EDF's nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1987-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

  9. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

    Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

    1984-01-01

    The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

  10. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  11. EDF'S nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1988-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

  12. Water-vapor pressure control in a volume

    Science.gov (United States)

    Scialdone, J. J.

    1978-01-01

    The variation with time of the partial pressure of water in a volume that has openings to the outside environment and includes vapor sources was evaluated as a function of the purging flow and its vapor content. Experimental tests to estimate the diffusion of ambient humidity through openings and to validate calculated results were included. The purging flows required to produce and maintain a certain humidity in shipping containers, storage rooms, and clean rooms can be estimated with the relationship developed here. These purging flows are necessary to prevent the contamination, degradation, and other effects of water vapor on the systems inside these volumes.

  13. Pressure-induced transformations in computer simulations of glassy water

    Science.gov (United States)

    Chiu, Janet; Starr, Francis W.; Giovambattista, Nicolas

    2013-11-01

    Glassy water occurs in at least two broad categories: low-density amorphous (LDA) and high-density amorphous (HDA) solid water. We perform out-of-equilibrium molecular dynamics simulations to study the transformations of glassy water using the ST2 model. Specifically, we study the known (i) compression-induced LDA-to-HDA, (ii) decompression-induced HDA-to-LDA, and (iii) compression-induced hexagonal ice-to-HDA transformations. We study each transformation for a broad range of compression/decompression temperatures, enabling us to construct a "P-T phase diagram" for glassy water. The resulting phase diagram shows the same qualitative features reported from experiments. While many simulations have probed the liquid-state phase behavior, comparatively little work has examined the transitions of glassy water. We examine how the glass transformations relate to the (first-order) liquid-liquid phase transition previously reported for this model. Specifically, our results support the hypothesis that the liquid-liquid spinodal lines, between a low-density and high-density liquid, are extensions of the LDA-HDA transformation lines in the limit of slow compression. Extending decompression runs to negative pressures, we locate the sublimation lines for both LDA and hyperquenched glassy water (HGW), and find that HGW is relatively more stable to the vapor. Additionally, we observe spontaneous crystallization of HDA at high pressure to ice VII. Experiments have also seen crystallization of HDA, but to ice XII. Finally, we contrast the structure of LDA and HDA for the ST2 model with experiments. We find that while the radial distribution functions (RDFs) of LDA are similar to those observed in experiments, considerable differences exist between the HDA RDFs of ST2 water and experiment. The differences in HDA structure, as well as the formation of ice VII (a tetrahedral crystal), are a consequence of ST2 overemphasizing the tetrahedral character of water.

  14. Assessment of a small pressurized water reactor for industrial energy

    International Nuclear Information System (INIS)

    Klepper, O.H.; Fuller, L.C.; Myers, M.L.

    1977-01-01

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton

  15. Solubility of solid ferrocene in pressurized hot water

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Hohnová, Barbora; Planeta, Josef; Roth, Michal

    2010-01-01

    Roč. 55, č. 8 (2010), s. 2866-2869 ISSN 0021-9568 R&D Projects: GA ČR GA203/07/0886; GA ČR GA203/08/1465; GA ČR GA203/08/1536 Institutional research plan: CEZ:AV0Z40310501 Keywords : pressurized hot water * ferrocene * solubility Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 2.089, year: 2010

  16. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  17. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  18. Solubilities of oxygenated aromatic solids in pressurized hot water

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Planeta, Josef; Roth, Michal

    2009-01-01

    Roč. 54, č. 5 (2009), s. 1457-1461 ISSN 0021-9568 R&D Projects: GA ČR GA203/07/0886; GA ČR GA203/08/1536 Institutional research plan: CEZ:AV0Z40310501 Keywords : oxygenated aromatics * solubility * pressurized hot water Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 1.695, year: 2009

  19. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  20. 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

  1. Drinking water fluoride and blood pressure? An environmental study.

    Science.gov (United States)

    Amini, Hassan; Taghavi Shahri, Seyed Mahmood; Amini, Mohamad; Ramezani Mehrian, Majid; Mokhayeri, Yaser; Yunesian, Masud

    2011-12-01

    The relationship between intakes of fluoride (F) from drinking water and blood pressure has not yet been reported. We examined the relationship of F in ground water resources (GWRs) of Iran with the blood pressure of Iranian population in an ecologic study. The mean F data of the GWRs (as a surrogate for F levels in drinking water) were derived from a previously conducted study. The hypertension prevalence and the mean of systolic and diastolic blood pressures (SBP & DBP) of Iranian population by different provinces and genders were also derived from the provincial report of non-communicable disease risk factor surveillance of Iran. Statistically significant positive correlations were found between the mean concentrations of F in the GWRs and the hypertension prevalence of males (r = 0.48, p = 0.007), females (r = 0.36, p = 0.048), and overall (r = 0.495, p = 0.005). Also, statistically significant positive correlations between the mean concentrations of F in the GWRs and the mean SBP of males (r = 0.431, p = 0.018), and a borderline correlation with females (r = 0.352, p = 0.057) were found. In conclusion, we found the increase of hypertension prevalence and the SBP mean with the increase of F level in the GWRs of Iranian population.

  2. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  3. Pressure Transient Model of Water-Hydraulic Pipelines with Cavitation

    Directory of Open Access Journals (Sweden)

    Dan Jiang

    2018-03-01

    Full Text Available Transient pressure investigation of water-hydraulic pipelines is a challenge in the fluid transmission field, since the flow continuity equation and momentum equation are partial differential, and the vaporous cavitation has high dynamics; the frictional force caused by fluid viscosity is especially uncertain. In this study, due to the different transient pressure dynamics in upstream and downstream pipelines, the finite difference method (FDM is adopted to handle pressure transients with and without cavitation, as well as steady friction and frequency-dependent unsteady friction. Different from the traditional method of characteristics (MOC, the FDM is advantageous in terms of the simple and convenient computation. Furthermore, the mechanism of cavitation growth and collapse are captured both upstream and downstream of the water-hydraulic pipeline, i.e., the cavitation start time, the end time, the duration, the maximum volume, and the corresponding time points. By referring to the experimental results of two previous works, the comparative simulation results of two computation methods are verified in experimental water-hydraulic pipelines, which indicates that the finite difference method shows better data consistency than the MOC.

  4. The role of Monte Carlo burnup calculations in quantifying plutonium mass in spent fuel assemblies with non-destructive assay

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, Jack D.; Tobin, Stephen J.; Trellue, Holly R.; Fensin, Michael L. [Los Alamos National Laboratory, Los Alamos, (United States)

    2011-12-15

    The Next Generation Safeguards Initiate (NGSI) of the United States Department of Energy has funded a multi-laboratory/university collaboration to quantify plutonium content in spent fuel (SF) with non-destructive assay (NDA) techniques and quantify the capability of these NDA techniques to detect pin diversions from SF assemblies. The first Monte Carlo based spent fuel library (SFL) developed for the NGSI program contained information for 64 different types of SF assemblies (four initial enrichments, burnups, and cooling times). The maximum amount of fission products allowed to still model a 17x17 Westinghouse pressurized water reactor (PWR) fuel assembly with four regions per fuel pin was modelled. The number of fission products tracked was limited by the available memory. Studies have since indicated that additional fission product inclusion and asymmetric burning of the assembly is desired. Thus, an updated SFL has been developed using an enhanced version of MCNPX, more powerful computing resources, and the Monte Carlo-based burnup code Monteburns, which links MCNPX to a depletion code and models a representative 1 Division-Slash 8 core geometry containing one region per fuel pin in the assemblies of interest, including a majority of the fission products with available cross sections. Often in safeguards, the limiting factor in the accuracy of NDA instruments is the quality of the working standard used in calibration. In the case of SF this is anticipated to also be true, particularly for several of the neutron techniques. The fissile isotopes of interest are co-mingled with neutron absorbers that alter the measured count rate. This paper will quantify how well working standards can be generated for PWR spent fuel assemblies and also describe the spatial plutonium distribution across an assembly. More specifically we will demonstrate how Monte Carlo gamma measurement simulations and a Monte Carlo burnup code can be used to characterize the emitted gamma

  5. Fission gas release from UO2 pellet fuel at high burn-up

    International Nuclear Information System (INIS)

    Vitanza, C.; Kolstad, E.; Graziani, U.

    1979-01-01

    Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

  6. Nuclear fuels with high burnup: safety requirements

    International Nuclear Information System (INIS)

    Phuc Tran Dai

    2016-01-01

    Vietnam authorities foresees to build 3 reactors from Russian design (VVER AES 2006) by 2030. In order to prepare the preliminary report on safety analysis the Vietnamese Agency for Radioprotection and Safety has launched an investigation on the behaviour of nuclear fuels at high burnups (up to 60 GWj/tU) that will be those of the new plants. This study deals mainly with the behaviour of the fuel assemblies in case of loss of coolant (LOCA). It appears that for an average burnup of 50 GWj/tU and for the advanced design of the fuel assembly (cladding and materials) safety requirements are fulfilled. For an average burnup of 60 GWj/tU, a list of issues remains to be assessed, among which the impact of clad bursting or the hydrogen embrittlement of the advanced zirconium alloys. (A.C.)

  7. Status of burnup credit implementation in Switzerland

    International Nuclear Information System (INIS)

    Grimm, P.

    1998-01-01

    Burnup credit is currently not used for the storage of spent fuel in the reactor pools in Switzerland, but credit is taken for integral burnable absorbers. Interest exists to take credit of burnup in future for the storage in a central away-from-reactor facility presently under construction. For spent fuel transports to foreign reprocessing plants the regulations of the receiving countries must be applied in addition to the Swiss licensing criteria. Burnup credit has been applied by one Swiss PWR utility for such transports in a consistent manner with the licensing practice in the receiving countries. Measurements of reactivity worths of small spent fuel samples in a Swiss zero-power research reactor are at an early stage of planning. (author)

  8. Burnup dependence of coolant void reactivity for ACR-1000 cell

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Hebert, A.; Roubstov, D.; Altiparmakov, D.; Irish, D.

    2008-01-01

    The Advanced Candu Reactor (ACR-1000) is light water cooled, fueled with enriched uranium and has a smaller lattice pitch than the Candu-6. As a result, the neutronic behavior of the ACR-1000 cell is expected to be somewhat different from that of the Candu-6 leading to a negative coolant void reactivity (CVR). Here we evaluate the CVR for the ACR-1000 cell using the lattice code DRAGON and compare our results with those obtained using the code WIMS-AECL. The differences observed between these two codes for the burnup dependence of the CVR is mainly explained in terms of the specific cell leakage model used by both codes. (authors)

  9. COGEMA/TRANSNUCLEAIRE's experience with burnup credit

    International Nuclear Information System (INIS)

    Chanzy, Y.; Guillou, E.

    1998-01-01

    Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)

  10. Possibilities of tritium removal from waste waters of pressurized water reactors and fuel reprocessing plants

    International Nuclear Information System (INIS)

    Ribnikar, S.V.; Pupezin, J.D.

    1975-01-01

    Starting from parameters known for heavy water production processes, a parallel was made with separation of tritium from water. The quantity in common is the total cascade flow. The most efficient processes appear to be hydrogen sulfide, water exchange, hydrogen- and water distillation. Prospects of application of new processes are discussed briefly. Problems concerning detritiation of pressurized water reactors and large fuel reprocessing plants are analyzed. Detritiation of the former should not present problems. With the latter, economical detritiation can be achieved only after some plant flow patterns are changed. (U.S.)

  11. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    Matausek, M.

    1984-01-01

    In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

  12. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  13. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  14. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  15. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

  16. Triton burnup in JET - profile effects

    International Nuclear Information System (INIS)

    Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van

    1991-01-01

    Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small ( 2 /s). (author) 4 refs., 3 figs

  17. Graph based hydraulic modelling of pressure in water distribution networks

    CSIR Research Space (South Africa)

    Mpiana, LA

    2017-09-01

    Full Text Available given period of time and can be formulated as Where 𝐵 is the amount of incoming water at the node 𝑖 and 𝑡 is the delivery time period of time required. The amount of incoming water required at the node for a period of time is compared... to the water demand 𝐷𝑖 in order to decide on the operation of the pressure reduction valves. 𝐵𝑖𝑗 is the coefficient to evaluation in the required amount 𝐵 over a specific period of time. For a node 𝑖, 𝐵𝑖𝑗 is equal to 0...

  18. Effect of fissile isotope burnup on criticality safety for stored disintegrated fuel rods

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Selby, G.P.

    1978-09-01

    If the fuel rods were to disintegrate and water added, a criticality could occur in a 13-in. PWR canister with fresh fuel enriched to 3.5 wt % 235 U. The question is, ''If credit could be taken for burnup, could this indicate a subcritical condition.'' In attempting to answer this question, a series of calculations were performed. A set of isotopic concentrations were generated for 5,000, 10,000, 15,000, and 20,000 MWD/MTU burnup levels. Four reflector materials, water, concrete and two types of soil, were considered. Results indicate that allowing credit for fissile isotope burnup does not completely remove the concern for criticality safety in the event of rod disintegration. Reactivities which are ''subcritical'' (k/sub eff/ = 0.95) would not occur for three of the four reflector materials at even the 20,000 MWD/MTU burnup level in the 13-in. canister. The water reflected canister would achieve the k/sub eff/ = 0.95 level near 18,000 MWD/MTU. A smaller canister could be postulated. If a quarter inch gap is allowed, a Westinghouse 17 x 17 PWR assembly requires a 12 1 / 4 inch diameter canister. For such a canister with water reflection the ''subcritical'' (k/sub eff/ = 0.95) level would be reached near 15,000 MWD/MTU. The soil reflected canisters would reach this level between 18,000 and 19,000 MWD/MTU. Considering the difficulties in taking credit for burnup, such modest gains in apparent safety are not encouraging. This situation might be improved, however, if credit were also taken for neutron absorption by fission product poisons produced during burnup. It is strongly recommended that other approaches to a solution of the criticality safety problem be considered

  19. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

  20. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  1. Enlarged Halden programme group meeting on high burn-up fuel performance, safety and reliability and degradation of in-core materials and water chemistry effects and man-machine systems research. Volume II

    International Nuclear Information System (INIS)

    1999-01-01

    An Enlarged Halden Programme Group Meeting on Fuel Performance and Materials Testing and Man-Machine Research was held at Loen, Norway 24 to 29 May, 1999. The papers presented covered work performed at the OECD Halden Reactor Project during the years up to April 1999 as well as studies conducted within organisations participating in the Halden Project. The Halden Boiling Water Reactor, HBWR, was built in the years 1955 - 1958 by the Norwegian Institutt for Atomenergi, and has since 1958 been operated internationally as one of the joint undertakings of the OECD Nuclear Energy Agency. The experimental operation of the Halden Boiling Water Reactor and associated programmes are sponsored through an international agreement by Institute for energy technology, Norway, the Belgian Nuclear Research Centre SCK/CEN, acting also on behalf of other public or private organisations in Belgium, the Risoe National Laboratory in Denmark, the Finnish Ministry of Trade and Industry, the Electricite de France, the Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, representing a German group of companies working in agreement with the German Federal Ministry for Education, Science, Research and Technology, the Italian Ente per le Nouve Tecnologie, l'Energia e l'Ambiente, the Japan Atomic Energy Research Institute, the Korean Atomic Energy Research Institute, the Spanish Centre de Investigaciones Energeticas, Medioambientales y Tecnologias, representing a group of national and industry organisations in Spain, the Swedish Nuclear Power Inspectorate, the Swiss Federal Nuclear Safety Inspectorate, representing also the Swiss nuclear utilities and the Paul Scherrer Institute, the Nuclear Electric pic, representing a group of nuclear research and industry organisations in the United Kingdom, and the United States Nuclear Regulatory Commission, and as associated parties: the Brazilian Comissao Nacional de Energia Nuclear (CNEN), the Czech Nuclear Research Institute, the Hungarian

  2. CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback

    International Nuclear Information System (INIS)

    Ahnert, Carol; Aragones, Jose M.

    1983-01-01

    1 - Description of problem or function: CARMEN is a system of programs developed for the neutronic calculation of PWR cycles. It includes the whole chain of analysis from cell calculations to core calculations with burnup. The core calculations are based on diffusion theory with cross sections depending on the relevant space-dependent feedback effects which are present at each moment along the cycles. The diffusion calculations are in one, two or three dimensions and in two energy groups. The feedback effects which are treated locally are: burnup, water density, power density and fission products. In order to study in detail these parameters the core should be divided into as many zones as different cross section sets are expected to be required in order to reproduce reality correctly. A relevant difference in any feedback parameter between zones produces different cross section sets for the corresponding zones. CARMEN is also capable to perform the following calculations: - Multiplication factor by burnup step with fixed boron concentration - Buckling and control rod insertion - Buckling search by burnup step - Boron search by burnup step - Control rod insertion search by burnup step. 2 - Method of solution: The cell code (LEOPARD-TRACA) generates the fuel assembly cross sections versus burnup. This is the basic library to be used in the CARMEN code proper. With a planar distribution guess for power density, water density and fluxes, the macroscopic cross sections by zone are calculated by CARMEN, and then a diffusion calculation is done in the whole geometry. With the distribution of power density, heat accumulated in the coolant and the thermal and fast fluxes determined in the diffusion calculation, CARMEN calculates the values of the most relevant parameters that influence the macroscopic cross sections by zone: burnup, water density, effective fuel temperature and fission product concentrations. If these parameters by zone are different from the reference

  3. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  4. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  5. Can the water content of highly compacted bentonite be increased by applying a high water pressure?

    International Nuclear Information System (INIS)

    Pusch, R.; Kasbohm, J.

    2001-10-01

    A great many laboratory investigations have shown that the water uptake in highly compacted MX-80 clay takes place by diffusion at low external pressure. It means that wetting of the clay buffer in the deposition holes of a KBS-3 repository is very slow if the water pressure is low and that complete water saturation can take several tens of years if the initial degree of water saturation of the buffer clay and the ability of the rock to give off water are low. It has therefore been asked whether injection of water can raise the degree of water saturation and if a high water pressure in the nearfield can have the same effect. The present report describes attempts to moisten highly compacted blocks of MX-80 clay with a dry density of 1510 kg/m 3 by injecting water under a pressure of 650 kPa through a perforated injection pipe for 3 and 20 minutes, respectively. The interpretation was made by determining the water content of a number of samples located at different distances from the pipe. An attempt to interpret the pattern of distribution of injected uranium acetate solution showed that the channels into which the solution went became closed in a few minutes and that dispersion in the homogenized clay gave low U-concentrations. The result was that the water content increased from about 9 to about 11-12 % within a distance of about 1 centimeter from the injection pipe and to slightly more than 9 % at a distance of about 4-5 cm almost independently of the injection time. Complete water saturation corresponds to a water content of about 30 % and the wetting effect was hence small from a practical point of view. By use of microstructural models it can be shown that injected water enters only the widest channels that remain after the compaction and that these channels are quickly closed by expansion of the hydrating surrounding clay. Part of the particles that are thereby released become transported by the flowing water and cause clogging of the channels, which is

  6. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

  7. Liquid radioactive waste processing system for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

  8. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  9. Culinary and pressure irrigation water system hydroelectric generation

    Energy Technology Data Exchange (ETDEWEB)

    Christiansen, Cory [Water Works Engineers, Pleasant Grove City, UT (United States)

    2016-01-29

    Pleasant Grove City owns and operates a drinking water system that included pressure reducing stations (PRVs) in various locations and flow conditions. Several of these station are suitable for power generation. The City evaluated their system to identify opportunities for power generation that can be implemented based on the analysis of costs and prediction of power generation and associated revenue. The evaluation led to the selection of the Battle Creek site for development of a hydro-electric power generating system. The Battle Creek site includes a pipeline that carries spring water to storage tanks. The system utilizes a PRV to reduce pressure before the water is introduced into the tanks. The evaluation recommended that the PRV at this location be replaced with a turbine for the generation of electricity. The system will be connected to the utility power grid for use in the community. A pelton turbine was selected for the site, and a turbine building and piping system were constructed to complete a fully functional power generation system. It is anticipated that the system will generate approximately 440,000 kW-hr per year resulting in $40,000 of annual revenue.

  10. Burnup calculation for a tokamak commercial hybrid reactor

    International Nuclear Information System (INIS)

    Feng Kaiming; Xie Zhongyou

    1990-08-01

    A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

  11. Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

    International Nuclear Information System (INIS)

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  12. A review on water fault diagnosis of PEMFC associated with the pressure drop

    International Nuclear Information System (INIS)

    Pei, Pucheng; Li, Yuehua; Xu, Huachi; Wu, Ziyao

    2016-01-01

    Highlights: • Reviewed the effect factors and estimations of pressure drop associated with water fault diagnosis. • Reviewed pressure drop-based water fault diagnosis using different indicators. • Deviation of pressure drop is used frequently to diagnose water fault. • Reviewed recovery strategies based on pressure drop used in commercial PEMFC. • Merits, demerits and application prospects of pressure drop-based water fault diagnosis are discussed. - Abstract: The pressure difference between the inlet and outlet of the reactant in fuel cells is called the pressure drop, which is related to the water amount inside the fuel cells. In recent years there have been many studies that used the pressure drop to detect the water content and diagnose water fault of proton exchange membrane fuel cells (PEMFCs). To our knowledge, there has not been a systematic review of these studies. In this paper, the effect variables of pressure drop are reviewed firstly. Then estimations of the theoretical pressure drop are reviewed mainly based on the following four aspects: Bernoulli’s equation, two-phase flow multiplier, Darcy’s law and artificial intelligence. Afterward, the water fault diagnosis based on the pressure drop using the following six indicators are reviewed: indicator of direct pressure drop, its deviation, frequency, multiplier, the ratio of pressure drop to flow rate and the flooding degree. In addition, the strategies of water fault recovery are also summarized. Finally the merits, demerits and application prospects of pressure drop-based water fault diagnosis are presented.

  13. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

    Lyric, Zoairia Idris; Mahmood, Mohammad Sayem; Motalab, Mohammad Abdul; Khan, Jahirul Haque

    2013-01-01

    Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor

  14. 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.

  15. Burnup credit effect on proposed cask payloads

    International Nuclear Information System (INIS)

    Hall, I.K.

    1989-01-01

    The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems which will allow safe and economical movement of commercial spent nuclear fuel and high-level waste from the generator to the Federal repository or Monitored Retrievable Storage (MRS) facility. Program schedule objectives for the initial phase of the CSDP include the development of certified spent fuel cask systems by 1995 to support Office of Civilian Radioactive Waste Management shipments from the utilities beginning in the late 1990s. Forty-nine proposals for developing a family of spent fuel casks were received and comparisons made. General conclusions that can be drawn from the comparisons are that (1) the new generation of casks will have substantially increased payloads in comparison to current casks, and (2) an even greater payload increase may be achievable with burnup credit. The ranges in the payload estimates do not allow a precise separation of the payload increase attributable to the proposed allowance of fuel burnup credit, as compared wilt the no-burnup-credit case. The beneficial effects of cask payload increases on overall costs and risks of transporting spent fuel are significant; therefore further work aimed toward taking advantage of burnup credit is warranted

  16. WWER-1000 Burnup Credit Benchmark (CB5)

    International Nuclear Information System (INIS)

    Manolova, M.A.

    2002-01-01

    In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)

  17. Whole core burnup calculations using 'MCNP'

    International Nuclear Information System (INIS)

    Haran, O.; Shaham, Y.

    1996-01-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

  18. Whole core burnup calculations using `MCNP`

    Energy Technology Data Exchange (ETDEWEB)

    Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

  19. Steam condensation behavior of high pressure water's blow down directly into water in containment under LOCA

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Ishida, Toshihisa; Yoritsune, Tsutomu; Kasahara, Y.

    1995-01-01

    JAERI has been conducting a design study of an advanced type Marine Reactor X (MRX) for merchant ships. By employing 'Integral type PWR', In-vessel type control rod drive systems', 'Water filled containment system' and 'Decay heat removal system by natural convection', MRX achieved a compact, light weight and highly safe plant. Experiments on steam condensation behavior of high pressure water's blow down into water have been conducted in order to investigate a major safety issue related to the design decision of 'Water filled containment system'. (author)

  20. Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1991-04-01

    To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs

  1. Burnup calculations using Monte Carlo method

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Degweker, S.B.

    2009-01-01

    In the recent years, interest in burnup calculations using Monte Carlo methods has gained momentum. Previous burn up codes have used multigroup transport theory based calculations followed by diffusion theory based core calculations for the neutronic portion of codes. The transport theory methods invariably make approximations with regard to treatment of the energy and angle variables involved in scattering, besides approximations related to geometry simplification. Cell homogenisation to produce diffusion, theory parameters adds to these approximations. Moreover, while diffusion theory works for most reactors, it does not produce accurate results in systems that have strong gradients, strong absorbers or large voids. Also, diffusion theory codes are geometry limited (rectangular, hexagonal, cylindrical, and spherical coordinates). Monte Carlo methods are ideal to solve very heterogeneous reactors and/or lattices/assemblies in which considerable burnable poisons are used. The key feature of this approach is that Monte Carlo methods permit essentially 'exact' modeling of all geometrical detail, without resort to ene and spatial homogenization of neutron cross sections. Monte Carlo method would also be better for in Accelerator Driven Systems (ADS) which could have strong gradients due to the external source and a sub-critical assembly. To meet the demand for an accurate burnup code, we have developed a Monte Carlo burnup calculation code system in which Monte Carlo neutron transport code is coupled with a versatile code (McBurn) for calculating the buildup and decay of nuclides in nuclear materials. McBurn is developed from scratch by the authors. In this article we will discuss our effort in developing the continuous energy Monte Carlo burn-up code, McBurn. McBurn is intended for entire reactor core as well as for unit cells and assemblies. Generally, McBurn can do burnup of any geometrical system which can be handled by the underlying Monte Carlo transport code

  2. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  3. Contribution to a neutronic calculation scheme for pressurized water reactors

    International Nuclear Information System (INIS)

    Martin Del Campo, C.

    1987-01-01

    This research thesis aims at developing and validating the set of data and codes which build up the neutron computation scheme of pressurized water reactors. More precisely, it focuses on the improvement of the precision of calculation of command clusters (absorbing components which can be inserted into the core to control the reactivity), and on the modelling of reflector representation (material placed around the core and reflecting back the escaping neutrons). For the first case, a precise calculation is performed, based on the transport theory. For the second case, diffusion constants obtained in the previous case and simplified equations are used to reduce the calculation cost

  4. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Vasile, A.; Dufour, Ph.; Golfier, H.; Grouiller, J.P.; Guillet, J.L.; Poinot, Ch.; Youinou, G.; Zaetta, A.

    2003-01-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1 . More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  5. Validation of the dynamic model for a pressurized water reactor

    International Nuclear Information System (INIS)

    Zwingelstein, Gilles.

    1979-01-01

    Dynamic model validation is a necessary procedure to assure that the developed empirical or physical models are satisfactorily representing the dynamic behavior of the actual plant during normal or abnormal transients. For small transients, physical models which represent isolated core, isolated steam generator and the overall pressurized water reactor are described. Using data collected during the step power changes that occured during the startup procedures, comparisons of experimental and actual transients are given at 30% and 100% of full power. The agreement between the transients derived from the model and those recorded on the plant indicates that the developed models are well suited for use for functional or control studies

  6. Environment sensitive cracking in light water reactor pressure boundary materials

    International Nuclear Information System (INIS)

    Haenninen, H.; Aho-Mantila, I.

    1985-01-01

    The purpose of the paper is to review the available methods and the most promising future possibilities of preventive maintenance to counteract the various forms of environment sensitive cracking of pressure boundary materials in light water reactors. Environment sensitive cracking is considered from the metallurgical, mechanical and environmental point of view. The main emphasis is on intergranular stress corrosion cracking of austenitic stainless steels and high strenght Ni-base alloys, as well as on corrosion fatigue of low alloy and stainless steels. Finally, some general ideas how to predict, reduce or eliminate environment sensitive cracking in service are presented

  7. Twin header bore welded steam generator for pressurized water reactors

    International Nuclear Information System (INIS)

    Davies, R.J.; Hirst, B.

    1979-01-01

    A description is given of a pressurized water reactor (PWR) steam generator concept, several examples of which have been in service for up to fourteen years. Details are given of the highly successful service record of this equipment and the features which have been incorporated to minimize corrosion and deposition pockets. The design employs a vertical U tube bundle carried off two horizontal headers to which the tubes are welded by the Foster Wheeler Power Products (FWPP) bore welding process. The factors to be considered in uprating the design to meet the current operating conditions for a 1000 MW unit are discussed. (author)

  8. State of the art of the advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Chawla, R.

    1987-01-01

    A review is given of the present status of the works concerned with an advanced pressurized water reactor (APWR). It includes the following items: reactor physics, thermal and hydraulic investigations and other engineering aspects as well as an analysis of electricity generation cost and long-term problems of embedding the APWR in a plutonium economy. As a summary it can be stated that there are discernible no principal obstacles of technically accomplishing an APWR, but there will be necessary considerable expenses in research and development works if it should be intended to start commercial service of an APWR up to the end of this century. (author)

  9. Self-Propagating Frontal Polymerization in Water at Ambient Pressure

    Science.gov (United States)

    Olten, Nesrin; Kraigsley, Alison; Ronney, Paul D.

    2003-01-01

    Advances in polymer chemistry have led to the development of monomers and initiation agents that enable propagating free-radical polymerization fronts to exist. These fronts are driven by the exothermicity of the polymerization reaction and the transport of heat from the polymerized product to the reactant monomer/solvent/initiator solution. The thermal energy transported to the reactant solution causes the initiator to decompose, yielding free radicals, which start the free radical polymerization process as discussed in recent reviews. The use of polymerization processes based on propagating fronts has numerous applications. Perhaps the most important of these is that it enables rapid curing of polymers without external heating since the polymerization process itself provides the high temperatures necessary to initiate and sustain polymerization. This process also enables more uniform curing of arbitrarily thick samples since it does not rely on heat transfer from an external source, which will necessarily cause the temperature history of the sample to vary with distance from the surface according to a diffusion-like process. Frontal polymerization also enables filling and sealing of structures having cavities of arbitrary shape without having to externally heat the structure. Water at atmospheric pressure is most convenient solvent to employ and the most important for practical applications (because of the cost and environmental issues associated with DMSO and other solvents). Nevertheless, to our knowledge, steady, self-propagating polymerization fronts have not been reported in water at atmospheric pressure. Currently, polymerization fronts require a high boiling point solvent (either water at high pressures or an alternative solvent such as dimethyl sulfoxide (DMSO) (boiling point 189 C at atmospheric pressure.) Early work on frontal polymerization, employed pressures up to 5000 atm in order to avoid boiling of the monomer/solvent/initiator solution. High

  10. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  11. Pressurized water reactor iodine spiking behavior under power transient conditions

    International Nuclear Information System (INIS)

    Ho, J.C.

    1992-01-01

    The most accepted theory explaining the cause of pressurized water reactor iodine spiking is steam formation and condensation in damaged fuel rods. The phase transformation of the primary coolant from water to steam and back again is believed to cause the iodine spiking phenomenon. But due to the complex nature of the phenomenon, a comprehensive model of the behavior has not yet been successfully developed. This paper presents a new model based on an empirical approach, which gives a first-order estimation of the peak iodine spiking magnitude. Based on the proposed iodine spiking model, it is apparent that it is feasible to derive a correlation using the plant operating data base to monitor and control the peak iodine spiking magnitude

  12. Disintegration of liquid metals by low pressure water blasting

    International Nuclear Information System (INIS)

    Heshmatpour, B.; Copeland, G.L.

    1981-01-01

    The feasibility of disintegrating metals by a low cost system and subsequently incorporating them into grout mixtures has been demonstrated. A low pressure water blasting technique consisting of multiple nozzles and a converging-line jet stream was developed to disintegrate liquid metals and produce coarse metal powder and shot. Molten iron resulted in spherical shot, while copper, aluminum, and tin produced irregular shaped particles. The particle size was between 0.05 and 3 mm (0.002 and 0.1 in.), and about half the particles were smaller than 1 mm (0.04 in.) in all cases. The water consumption was rather low, while the production rate was relatively high. The method proved to be simple and reliable. The coarse metal powders were suspendable in grout fluids, indicating that they are probably disposable by the shale hydrofracture technique

  13. Corrosion fatigue cracking behavior of Inconel 690 (TT) in secondary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Xiao Jun; Chen Luyao; Qiu Shaoyu; Chen Yong; Lin Zhenxia; Fu Zhenghong

    2015-01-01

    Inconel 690 (TT) is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690 (TT) was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio. (authors)

  14. Dual pressurized light water reactor producing 2000 M We

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

  15. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  16. Apparatus for localizing disturbances in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Sykora, D.

    1989-01-01

    The invention according to CS-PS 177386, entitled ''Apparatus for increasing the efficiency and passivity of the functioning of a bubbling-vacuum system for localizing disturbances in nuclear power plants with a pressurized water reactor'', concerns an important area of nuclear power engineering that is being developed in the RGW member countries. The invention solves the problems of increasing the reliability and intensification during the operation of the above very important system for guaranteeing the safety of the standard nuclear power plants of Soviet design. The essence of the invention consists in the installation of a simple passively operating supplementary apparatus. Consequently, the following can be observed in the system: first an improvement and simultaneous increase in the reliability of its function during the critical transition period, which follows the filling of the second space with air from the first space; secondly, elimination of the hitherto unavoidable initiating role of the active sprinkler-condensation device present; thirdly, a more effective performance and subjection of the elements to disintegration of the water flowing from the bubbling condenser into the first space; and fourthly, an enhanced utilization of the heat-conducting ability of the water reservoir of the bubbling condenser. Representatives of the supplementary apparatus are autonomous and local secondary systems of the sprinkler-sprayer without an insert, which spray the water under the effect of gravity. 1 fig

  17. Acceptance test procedure for High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    The overall objective of the acceptance test is to demonstrate a combined system. This includes associated tools and equipment necessary to perform cleaning in the 105 K East Basin (KE) for achieving optimum reduction in the level of contamination/dose rate on canisters prior to removal from the KE Basin and subsequent packaging for disposal. Acceptance tests shall include necessary hardware to achieve acceptance of the cleaning phase of canisters. This acceptance test procedure will define the acceptance testing criteria of the high pressure water jet cleaning fixture. The focus of this procedure will be to provide guidelines and instructions to control, evaluate and document the acceptance testing for cleaning effectiveness and method(s) of removing the contaminated surface layer from the canister presently identified in KE Basin. Additionally, the desired result of the acceptance test will be to deliver to K Basins a thoroughly tested and proven system for underwater decontamination and dose reduction. This report discusses the acceptance test procedure for the High Pressure Water Jet

  18. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    International Nuclear Information System (INIS)

    Wiesenack, W.

    1996-01-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project's data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup

  19. High Pressure Water Stripping Using Multi-Orifice Nozzles

    Science.gov (United States)

    Hoppe, David

    1999-01-01

    The use of multi-orifice rotary nozzles greatly increases the speed and stripping effectiveness of high pressure water blasting systems, but also greatly increases the complexity of selecting and optimizing the operating parameters. The rotational speed of the nozzle must be coupled with its transverse velocity as it passes across the surface of the substrate being stripped. The radial and angular positions of each orifice must be included in the analysis of the nozzle configuration. Orifices at the outer edge of the nozzle head move at a faster rate than the orifices located near the center. The energy transmitted to the surface from the impact force of the water stream from an outer orifice is therefore spread over a larger area than energy from an inner orifice. Utilizing a larger diameter orifice in the outer radial positions increases the total energy transmitted from the outer orifice to compensate for the wider distribution of energy. The total flow rate from the combination of all orifices must be monitored and should be kept below the pump capacity while choosing orifice to insert in each position. The energy distribution from the orifice pattern is further complicated since the rotary path of all the orifices in the nozzle head pass through the center section. All orifices contribute to the stripping in the center of the path while only the outer most orifice contributes to the stripping at the edge of the nozzle. Additional orifices contribute to the stripping from the outer edge toward the center section. With all these parameters to configure and each parameter change affecting the others, a computer model was developed to track and coordinate these parameters. The computer simulation graphically indicates the cumulative affect from each parameter selected. The result from the proper choices in parameters is a well designed, highly efficient stripping system. A poorly chosen set of parameters will cause the nozzle to strip aggressively in some areas

  20. ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Historical background and information: - 28-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. - 37-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the Candu Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency with reactor physics codes. 2 - Application of the data: The libraries in this data collection are designed for characterising used fuel from Candu pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. 3 - Source and scope of data: The Candu libraries are updated with cross sections from a variety of different sources. Capture

  1. Research on axial total pressure distributions of sonic steam jet in subcooled water

    International Nuclear Information System (INIS)

    Wu Xinzhuang; Li Wenjun; Yan Junjie

    2012-01-01

    The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters (6.0 mm, 8.0 mm and 10.0 mm). The inlet steam pressure, and pool subcooling subcooled water temperature were in the range of 0.2-0.6 MPa and 420-860 ℃, respectively. The effect of steam pressure, subcooling water temperature and nozzle size on the axial pressure distributions were obtained, and also the characteristics of the maximum pressure and its position were studied. The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure, but the influence could be ignored for high steam pressure. Moreover, a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature, and the discrepancies of predictions and experiments are within ±15%. (authors)

  2. Burnup credit activities being conducted in the United States

    International Nuclear Information System (INIS)

    Lake, W.

    1998-01-01

    The paper describes burnup credit activities being conducted in the U.S. where burnup credit is either being used or being planned to be used for storage, transport, and disposal of spent nuclear fuel. Currently approved uses of burnup credit are for wet storage of PWR fuel. For dry storage of spent PWR fuel, burnup credit is used to supplement a principle of moderator exclusion. These storage applications have been pursued by the private sector. The Department of Energy (DOE) which is an organization of the U.S. Federal government is seeking approval for burnup credit for transport and disposal applications. For transport of spent fuel, regulatory review of an actinide-only PWR burnup credit method is now being conducted. A request by DOE for regulatory review of actinide and fission product burnup credit for disposal of spent BWR and PWR fuel is scheduled to occur in 1998. (author)

  3. Core burn-up calculation method of JRR-3

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Yamashita, Kiyonobu

    2007-01-01

    SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)

  4. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    Tran, Viet-Phu; Tran, Hoai-Nam; Yamamoto, Akio; Endo, Tomohiro

    2017-01-01

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO_2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

  5. Automated generation of burnup chain for reactor analysis applications

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Viet-Phu [VINATOM, Hanoi (Viet Nam). Inst. for Nuclear Science and Technology; Tran, Hoai-Nam [Duy Tan Univ., Da Nang (Viet Nam). Inst. of Research and Development; Yamamoto, Akio; Endo, Tomohiro [Nagoya Univ., Nagoya-shi (Japan). Dept. of Materials, Physics and Energy Engineering

    2017-05-15

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO{sub 2} and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

  6. Blind prediction exercise on modeling of PHWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Sah, D.N.; Viswanathan, U.K.; Viswanadham, C.S.; Unnikrishnan, K.; Rath, B.N.

    2008-01-01

    A blind prediction exercise was organised on Indian Pressurised Heavy Water Reactor (PHWR) fuel to investigate the predictive capability of existing codes for their application at extended burnup and to identify areas of improvement. The blind problem for this exercise was based on a PHWR fuel bundle irradiated in Kakrapar Atomic Power Station-I (KAPS-I) up to about 15 000 MWd/tU and subjected to detailed post-irradiation examination (PIE) in the hot cells facility at BARC. Eleven computer codes from seven countries participated in this exercise. The participants provided blind predictions of fuel temperature, fission gas release, internal gas pressure and other performance parameters for the fuel pins. The predictions were compared with the experimental PIE data which included fuel temperature derived from fuel restructuring, fission gas release measured by fuel pin puncturing, internal gas pressure in pin, cladding oxidation and fuel microstructural data. The details of the blind problem and an analysis of the results of blind predictions by the codes vis-a-vis measured data are provided in this paper

  7. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Margen, P H; Ahlstroem, P E; Pershagen, B

    1961-04-15

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

  8. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

    1961-04-01

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  9. Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Park, T-K.; Jung, S-H.

    1996-01-01

    During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

  10. Performance of Fragema fuel in pressurized water reactors

    International Nuclear Information System (INIS)

    Dumont, A.; Ravier, G.; Ballot, B.

    1986-06-01

    FRAGEMA fuel operating experience in power reactors is very extensive. Performance over a range of power and burnup levels for various operating conditions is quite satisfactory. However significant development programs are presently in progress to further extend our knowledge under increasingly severe operating conditions. In particular, upcoming data acquisition programs (1985-1988) will cover site and hot cell measurements on Gd poison rods, 4.5 % overenriched fuel rods over four operating cycles, 17 x 17 AFA fuel assemblies. For these products the same surveillance strategy as the one used for the standard assembly has been adopted, in order to continuously provide more data which can be used to upgrade design models and pave the way for the development of future products

  11. Temperature/pressure and water vapor sounding with microwave spectroscopy

    Science.gov (United States)

    Muhleman, D. O.; Janssen, M. A.; Clancy, R. T.; Gulkis, S.; Mccleese, D. J.; Zurek, R.; Haberle, R. M.; Frerking, M.

    1992-01-01

    Two intense microwave spectra lines exist in the martian atmosphere that allow unique sounding capabilities: water vapor at 183 GHz and the (2-1) rotational line of CO at 230 GHz. Microwave spectra line sounding is a well-developed technique for the Earth's atmosphere for sounding from above from spacecraft and airplanes, and from below from fixed surface sites. Two simple instruments for temperature sounding on Mars (the CO line) and water vapor measurements are described. The surface sounder proposed for the MESUR sites is designed to study the boundary layer water vapor distribution and the temperature/pressure profiles with vertical resolution of 0.25 km up to 1 km with reduced resolution above approaching a scale height. The water channel will be sensitive to a few tenths of a micrometer of water and the temperature profile will be retrieved to an accuracy between 1 and 2 K. The latter is routinely done on the Earth using oxygen lines near 60 GHz. The measurements are done with a single-channel heterodyne receiver looking into a 10-cm mirror that is canned through a range of elevation angles plus a target load. The frequency of the receiver is sweep across the water and CO lines generating the two spectra at about 1-hr intervals throughout the mission. The mass and power for the proposed instrument are 2 kg and 5-8 W continuously. The measurements are completely immune to the atmospheric dust and ice particle loads. It was felt that these measurements are the ultimate ones to properly study the martian boundary layer from the surface to a few kilometers. Sounding from above requires an orbiting spacecraft with multichannel microwave spectrometers such as the instrument proposed for MO by a subset of the authors, a putative MESUR orbiter, and a proposed Discovery mission called MOES. Such an instrument can be built with less than 10 kg and use less than 15 W. The obvious advantage of this approach is that the entire atmosphere can be sounded for temperature and

  12. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  13. High converter pressurized water reactor with heavy water as a coolant

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  14. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    Finck, P.J.; Stenberg, C.G.

    1998-01-01

    The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work

  15. Triton burnup in JET - profile effects

    Energy Technology Data Exchange (ETDEWEB)

    Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van (Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking); Adams, J.M.; Watkins, N. (AEA Industrial Technology, Harwell Laboratory (United Kingdom))

    1991-01-01

    Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small (<<0.1 m[sup 2]/s). (author) 4 refs., 3 figs.

  16. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    International Nuclear Information System (INIS)

    Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

    2011-01-01

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  17. K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1998-08-01

    This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k inf values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon consultation with the YMP staff, a Quad Cities BWR fuel assembly was selected as a baseline assembly. This design consists of seven axial enrichment zones, three of which contain natural uranium oxide. No attempt was made to find a bounding or even typical assembly design due to the wide variety in fuel assembly designs necessary for consideration. The current work concentrates on establishing a baseline analysis, along with a small number of sensitivity studies which can be expected later if desired. As a result of similar studies of this nature, several effects are known to be important in the determination of the final k inf for spent fuel in a cask-like geometry. For a given enrichment there is an optimal burnup: for lower burnups, excess energy (and corresponding excess reactivity) is present in the fuel assembly; for larger burnups, the assembly is overburned and essentially driven by neighboring fuel assemblies. The majority of the burnup/enrichment scenarios included in this study were for some near-optimum burnup/enrichment combinations as determined from Energy Information Administration (EIA) data. Several calculations were performed for under- and over-burned fuel to show these effects

  18. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    International Nuclear Information System (INIS)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

    2003-01-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

  19. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

    2003-07-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

  20. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Pavelescu, M.; Borza, M.

    1975-01-01

    The authors approach theoretical treatment of isotopic composition changement for nuclear fuel in nuclear reactors. They show the difficulty of exhaustive treatment of burn-up problems and introduce the principal simplifying principles. Due to these principles they write and solve analytically the evolution equations of the concentration for the principal nuclides both in the case of fast and thermal reactors. Finally, they expose and comment the results obtained in the case of a power fast reactor. (author)

  1. Contemporary pressurized water reactor technology in the world

    International Nuclear Information System (INIS)

    Komarek, A.

    1991-01-01

    The recent political events enabled Czechoslovak industrial companies to come into direct contact with leading western companies involved in pressurized water ractor technology. A survey is presented of the present situation at the world market of PWR type nuclear power plant suppliers and suppliers of fuel cycle services. Information is given on the potential bids for the next Czechoslovak nuclear power plants with PWR reactors. Economic aspects of the potential bids are presented including some considerations about the participation of the Czechoslovak nuclear industry as a supplier of the reactor for the future power plants. Main technical parameters are listed of PWR units with an output about 1000 MW supplied by Westinghouse EC, ABB -Combustion Engineering and Siemens AG. At present, the bids for new Czechoslovak nuclear power plants are being evaluated. No information on terms of the bids actually coming from foreign companies is used in the article. (Z.S.). 9 figs., 5 tabs

  2. Computer code for simulating pressurized water reactor core

    International Nuclear Information System (INIS)

    Serrano, A.M.B.

    1978-01-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

  3. Code for the core simulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1978-08-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt

  4. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  5. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  6. Aging assessment of PWR [Pressurized Water Reactor] Auxiliary Feedwater Systems

    International Nuclear Information System (INIS)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab

  7. Testing of a portable ultrahigh pressure water decontamination system (UHPWDS)

    International Nuclear Information System (INIS)

    Lovell, A.; Dahlby, J.

    1996-02-01

    This report describes the tests done with a portable ultrahigh pressure water decontamination system (UHPWDS) on highly radioactively contaminated surfaces. A small unit was purchased, modified, and used for in-situ decontamination to change the waste level of the contaminated box from transuranic (TRU) waste to low- level waste (LLW). Low-level waste is less costly by as much as a factor of five or more if compared with TRU waste when handling, storage, and disposal are considered. The portable unit we tested is commercially available and requires minimal utilities for operation. We describe the UHPWDS unit itself, a procedure for its use, the results of the testing we did, and conclusions including positive and negative aspects of the UHPWDS

  8. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  9. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    International Nuclear Information System (INIS)

    Lin Chaung; Shen Chihming

    2000-01-01

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller

  10. Design and analysis of pressurized water reactor systems

    International Nuclear Information System (INIS)

    Juhn, P.E.; Kim, Y.H.

    1979-01-01

    To help develop nuclear engineering technologies in local industry sectors, technical and economical data on pressurized water reactor systems and components have been collected, systematically analyzed and computerized to a certain degree. Codes and standards necessary for engineering design of PWR systems have been surveyed and clarified in terms of NSSS, turbine-generator system and BOP, then again rearranged with respect to quality classes and seismic classes. Some design manuals, criteria and guidelines regarding design, construction, test and operation of PWR plants have also been surveyed and collected. Benchmark cost calculation for the construction of a 900 MWe PWR plant, according to the standard format, was carried out, and computer model on construction costs was improved and updated by considering the local supply of labor and materials. And for the indigeneous development of PWR equipment and materials, such data as delivery schedule and manufacturers of 52 systems and 36,000 components have also been reviewed herein. (author)

  11. A dual pressurized water reactor producing 2000 MWe

    International Nuclear Information System (INIS)

    Kang, K. M.; Suh, K. Y.

    2010-01-01

    The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

  12. Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

    International Nuclear Information System (INIS)

    Tingkuan, C.; Xuanzheng, C.

    1987-01-01

    It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

  13. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  14. Reducing energy consumption and leakage by active pressure control in a water supply system

    NARCIS (Netherlands)

    Bakker, M.; Rajewicz, T.; Kien, H.; Vreeburg, J.H.G.; Rietveld, L.C.

    2013-01-01

    WTP Gruszczyn supplies drinking water to a part of the city of Pozna?, in the Midwest of Poland. For the optimal automatic pressure control of the clear water pumping station, nine pressure measuring points were installed in the distribution network, and an active pressure control model was

  15. Basal interstitial water pressure in laboratory debris flows over a rigid bed in an open channel

    Directory of Open Access Journals (Sweden)

    N. Hotta

    2012-08-01

    Full Text Available Measuring the interstitial water pressure of debris flows under various conditions gives essential information on the flow stress structure. This study measured the basal interstitial water pressure during debris flow routing experiments in a laboratory flume. Because a sensitive pressure gauge is required to measure the interstitial water pressure in shallow laboratory debris flows, a differential gas pressure gauge with an attached diaphragm was used. Although this system required calibration before and after each experiment, it showed a linear behavior and a sufficiently high temporal resolution for measuring the interstitial water pressure of debris flows. The values of the interstitial water pressure were low. However, an excess of pressure beyond the hydrostatic pressure was observed with increasing sediment particle size. The measured excess pressure corresponded to the theoretical excess interstitial water pressure, derived as a Reynolds stress in the interstitial water of boulder debris flows. Turbulence was thought to induce a strong shear in the interstitial space of sediment particles. The interstitial water pressure in boulder debris flows should be affected by the fine sediment concentration and the phase transition from laminar to turbulent debris flow; this should be the subject of future studies.

  16. Conceptual cask design with burnup credit

    International Nuclear Information System (INIS)

    Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong

    2003-01-01

    Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)

  17. Analytical study on water hammer pressure in pressurized conduits with a throttled surge chamber for slow closure

    Directory of Open Access Journals (Sweden)

    Yong-liang Zhang

    2010-06-01

    Full Text Available This paper presents an analytical investigation of water hammer in a hydraulic pressurized pipe system with a throttled surge chamber located at the junction between a tunnel and a penstock, and a valve positioned at the downstream end of the penstock. Analytical formulas of maximum water hammer pressures at the downstream end of the tunnel and the valve were derived for a system subjected to linear and slow valve closure. The analytical results were then compared with numerical ones obtained using the method of characteristics. There is agreement between them. The formulas can be applied to estimating water hammer pressure at the valve and transmission of water hammer pressure through the surge chamber at the junction for a hydraulic pipe system with a surge chamber.

  18. Water-bearing, high-pressure Ca-silicates

    Science.gov (United States)

    Németh, Péter; Leinenweber, Kurt; Ohfuji, Hiroaki; Groy, Thomas; Domanik, Kenneth J.; Kovács, István J.; Kovács, Judit S.; Buseck, Peter R.

    2017-07-01

    Water-bearing minerals provide fundamental knowledge regarding the water budget of the mantle and are geophysically significant through their influence on the rheological and seismic properties of Earth's interior. Here we investigate the CaO-SiO2-H2O system at 17 GPa and 1773 K, corresponding to mantle transition-zone condition, report new high-pressure (HP) water-bearing Ca-silicates and reveal the structural complexity of these phases. We document the HP polymorph of hartrurite (Ca3SiO5), post-hartrurite, which is tetragonal with space group P4/ncc, a = 6.820 (5), c = 10.243 (8) Å, V = 476.4 (8) Å3, and Z = 4, and is isostructural with Sr3SiO5. Post-hartrurite occurs in hydrous and anhydrous forms and coexists with larnite (Ca2SiO4), which we find also has a hydrous counterpart. Si is 4-coordinated in both post-hartrurite and larnite. In their hydrous forms, H substitutes for Si (4H for each Si; hydrogrossular substitution). Fourier transform infrared (FTIR) spectroscopy shows broad hydroxyl absorption bands at ∼3550 cm-1 and at 3500-3550 cm-1 for hydrous post-hartrurite and hydrous larnite, respectively. Hydrous post-hartrurite has a defect composition of Ca2.663Si0.826O5H1.370 (5.84 weight % H2O) according to electron-probe microanalysis (EPMA), and the Si deficiency relative to Ca is also observed in the single-crystal data. Hydrous larnite has average composition of Ca1.924Si0.851O4H0.748 (4.06 weight % H2O) according to EPMA, and it is in agreement with the Si occupancy obtained using X-ray data collected on a single crystal. Superlattice reflections occur in electron-diffraction patterns of the hydrous larnite and could indicate crystallographic ordering of the hydroxyl groups and their associated cation defects. Although textural and EPMA-based compositional evidence suggests that hydrous perovskite may occur in high-Ca-containing (or low silica-activity) systems, the FTIR measurement does not show a well-defined hydroxyl absorption band for this

  19. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 2: Human reliability analysis and human performance evaluation; Technical issues related to rulemakings; Risk-informed, performance-based initiatives; High burn-up fuel research

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues related to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research

  20. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 2: Human reliability analysis and human performance evaluation; Technical issues related to rulemakings; Risk-informed, performance-based initiatives; High burn-up fuel research

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues related to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  1. Analytical and numerical study of radiation effect up to high burnup in power reactor fuels

    International Nuclear Information System (INIS)

    Lemes, M; Denis, A; Soba, A

    2012-01-01

    In the present work the behavior of fuel pellets for power reactors in the high burnup range (average burnup higher than 50 MWd/kgHM) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup, as long as a new microstructure develops, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behaviour. The evolution of porosity in the high burnup structure (HBS) is assumed to be determinant of the retention capacity of the fission gases released by the matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Starting from several works published in the open literature, a model was developed to describe the behaviour and evolution of porosity at local burnup values ranging from 60 to 300 MWd/KgHM. The model is mathematically expressed by a system of non-linear differential equations that take into account the open and closed porosity, the interactions between pores and the free surface and phenomena like pore's coalescence and migration and gas venting. Interactions of different orders between open and closed pores, growth of pores radius by vacancies trapping, the evolution of the pores number density, the internal pressure and over pressure within the pores, the fission gas retained in the matrix and released to the free volume are analyzed. The results of the simulations performed in the present work are in excellent agreement with experimental data available in the open literature and with results calculated by other authors (author)

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

    International Nuclear Information System (INIS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-01-01

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

  3. Taipower's reload safety evaluation methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Ping-Hue; Yang, Y.S.

    1996-01-01

    For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

  4. Imaging optical probe for pressurized steam-water environment

    International Nuclear Information System (INIS)

    Donaldson, M.R.; Pulfrey, R.E.

    1979-01-01

    An air-cooled imaging optical probe, with an outside diameter of 25.4 mm, has been developed to provide high resolution viewing of flow regimes in a steam-water environment at 343 0 C and 15.2 MPa. The design study considered a 3-m length probe. A 0.3-m length probe prototype was fabricated and tested. The optical probe consists of a 3.5-mm diameter optics train surrounded by two coaxial coolant flow channels and two coaxial insulating dead air spaces. With air flowing through the probe at 5.7 g/s, thermal analysis shows that no part of the optics train will exceed 93 0 C when a 3-m length probe is immersed in a 343 0 C environment. Computer stress analysis plus actual tests show that the probe can operate successfully with conservative safety factors. The imaging optical probe was tested five times in the design environment at the semiscale facility at the INEL. Two-phase flow regimes in the high temperature, high pressure, steam-water blowdown and reflood experiments were recorded on video tape for the first time with the imaging optical probe

  5. Influence of graphite discs, chamfers, and plenums on temperature distributions in high burnup fuel

    International Nuclear Information System (INIS)

    Ranger, A.; Tayal, M.; Singh, P.

    1990-04-01

    Previous studies have demonstrated the desirability to increase the fuel burnups in CANDU reactors from 7-9 GW.d/t to 21 GW.d/t. At high burnups, one consideration in fuel integrity is fission gas pressure, which is predicted to reach about 13 MPa. The gas pressure can be kept below the coolant pressure (about 10 MPa) via a variety of options such as bigger chamfers, deeper dishes, central hole, and plenums. However, it is important to address the temperature perturbations produced by the bigger chamfers and plenums which in turn, affect the gas pressure. Another consideration in fuel integrity is to reduce the likelihood of fuel failures via environmentally assisted cracking. Insertion of graphite discs between neighbouring pellets will lower the pellet temperatures, hence, lower fission gas release and lower expansion of the pellet. Therefore, it is desired to quantify the effect of graphite discs on pellet temperatures. Thermal analyses of different fuel element geometries: with and without chamfers, graphite discs, and plenums were performed. The results indicate that the two-dimensional distributions of temperatures due to the presence of chamfers, graphite discs, or plenums can have a significant impact on the integrity of high burnup fuel as we have been able to quantify in this paper

  6. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  7. Robustness of parameter-less remote real-time pressure control in water distribution systems

    CSIR Research Space (South Africa)

    Page, Philip R

    2017-06-01

    Full Text Available One way of reducing water leakage, pipe bursts and water consumption in a water distribution system (WDS) is to manage the pressure to be as low as possible. This can be done by adjusting a pressure control valve (PCV) in real-time in order to keep...

  8. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    Energy Technology Data Exchange (ETDEWEB)

    Burdin, J.R.

    1995-10-25

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  9. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    International Nuclear Information System (INIS)

    Burdin, J.R.

    1995-01-01

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  10. Time step length versus efficiency of Monte Carlo burnup calculations

    International Nuclear Information System (INIS)

    Dufek, Jan; Valtavirta, Ville

    2014-01-01

    Highlights: • Time step length largely affects efficiency of MC burnup calculations. • Efficiency of MC burnup calculations improves with decreasing time step length. • Results were obtained from SIE-based Monte Carlo burnup calculations. - Abstract: We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy

  11. A Monte Carlo burnup code linking MCNP and REBUS

    International Nuclear Information System (INIS)

    Hanan, N. A.

    1998-01-01

    The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented

  12. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1993-01-01

    The use of burnup credit in the criticality safety analysis of the GA-4 Cask increases the cask's capacity from three spent fuel assemblies to four, resulting in reduced public and occupational risk and reduced life cycle costs. GA's criticality calculations for burnup credit, including the associated uncertainties and analytical bias, establish the minimum burnup required as a function of initial enrichment to maintain K eff ≤ 0.95 under any conceivable condition. The minimum burnup requirement as a function of initial enrichment has been determined to be 15,000 MWd/MTU for 3.5 wt% U-235 fuel, 20,000 MWd/MTU for 4.0 wt% U-235 fuel and 25,000 MWd/MTU for 4.5 wt% U-235 fuel. The minimum burnup requirement as a function of enrichment is well below the typical burnup levels seen in the current and projected spent fuel inventory. (J.P.N.)

  13. Implementation of burnup credit in spent fuel management systems

    International Nuclear Information System (INIS)

    Dyck, H.P.

    2001-01-01

    Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate safety margin. The motivation for using burnup credit in criticality safety applications is based on economic considerations and additional benefits contributing to public health and safety and resource conservation. Interest in the implementation of burnup credit has been shown by many countries. In 1997, the International Atomic Energy Agency (IAEA) started a task to monitor the implementation of burnup credit in spent fuel management systems, to provide a forum to exchange information, to discuss the matter and to gather and disseminate information on the status of national practices of burnup credit implementation in the Member States. The task addresses current and future aspects of burnup credit. This task was continued during the following years. (author)

  14. Condensation effects in a pressurizer scaled from a pressurized water reactor

    International Nuclear Information System (INIS)

    Loomis, G.G.; Shaw, R.A.

    1985-01-01

    This paper presents results from an experimental investigation of phenomena associated with pressurizer auxiliary spray during an abnormal plant transient in a commercial PWR. If normal pressurizer spray is unavailable (main coolant pumps are off) or the pressurizer power operated relief valve cannot be used during abnormal transients, pressurizer auxiliary spray can be used to reduce primary system pressure. Results from both transient integral experiments involving pressurizer auxiliary spray during tube rupture and separate effects spray experiments are presented. The experimental investigation was conducted in the Semiscale MOD-2B facility. Phenomenon of interest that occurred in the pressurizer during the pressurized auxiliary spray was desuperheating of the pressurizer steam space and quenching of metal walls followed by dropwise condensation of the pressurizer steam. The data from both the transient integral experiments and the separate effects experiments are compared to RELAP5 computer calculations and the capability of existing models in the code is discussed

  15. Theory analysis and simple calculation of travelling wave burnup scheme

    International Nuclear Information System (INIS)

    Zhang Jian; Yu Hong; Gang Zhi

    2012-01-01

    Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)

  16. Burnup calculation in microcells of high conversion reactors

    International Nuclear Information System (INIS)

    Gomez, S.E.; Salvatore, M.; Patino, N.E.; Abbate, M.J.

    1991-01-01

    The development of high converter reactors (HCR) requires careful burnup calculations because their main goals are reach high discharge burnup levels (Up to 50 GWd/T) and a close to one conversion ratio. Then, it is necessary a revision of design elements used for this type of calculation. In this work, a burnup module (BUM) developed in order to use nuclear data directly from evaluated data files is presented; these was included in the AMPX system. (author)

  17. Classification of criticality calculations with correlation coefficient method and its application to OECD/NEA burnup credit benchmarks phase III-A and II-A

    International Nuclear Information System (INIS)

    Okuno, Hiroshi

    2003-01-01

    A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the Nuclear Energy Agency of the Organisation for Economic Cooperation and Development (OECD/NEA). Phase III-A benchmark was a series of criticality calculations for irradiated Boiling Water Reactor (BWR) fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated Pressurized Water Reactor (PWR) fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark calculation results were classified according to the criterion that the values of the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of benchmark calculation results belonged to the same group, one calculation result was found predictable from the other. An example was shown for each of the Benchmarks. While the evaluated nuclear data seemed the main factor for the classification, further investigations were required for finding other factors. (author)

  18. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Yung Joon

    1994-02-01

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  19. High pressure, low pressure and hot water heating systems in hospitals. Hochdruck-, Niederdruck- und Warmwasserheizungsanlagen im Krankenhaus

    Energy Technology Data Exchange (ETDEWEB)

    Riedle, K [H. Riedle GmbH, Wiesbaden (Germany)

    1994-07-01

    In hospital nowadays the limitation of the use of steam boilers and their direct supply network to the possible minimum is aimed at when the heating system is exchanged or retrofitted. Independent of the fact whether high pressure or low pressure steam or hot water is used the optimum water treatment should be carried out with a minimum of chemical substances. Here hydroquinone, neutralizing amines, carbohydrazide, sodium sulphite and tannins can be used. The dimensioning of hot water heating circuits is shown with examples. (BWI)

  20. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

    1990-12-01

    A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

  1. Effects of ambient temperature and water vapor on chamber pressure and oxygen level during low atmospheric pressure stunning of poultry.

    Science.gov (United States)

    Holloway, Paul H; Pritchard, David G

    2017-08-01

    The characteristics of the vacuum used in a low atmospheric pressure stunning system to stun (render unconscious) poultry prior to slaughter are described. A vacuum chamber is pumped by a wet screw compressor. The vacuum pressure is reduced from ambient atmospheric pressure to an absolute vacuum pressure of ∼250 Torr (∼33 kPa) in ∼67 sec with the vacuum gate valve fully open. At ∼250 Torr, the sliding gate valve is partially closed to reduce effective pumping speed, resulting in a slower rate of decreasing pressure. Ambient temperature affects air density and water vapor pressure and thereby oxygen levels and the time at the minimum total pressure of ∼160 Torr (∼21 kPa) is varied from ∼120 to ∼220 sec to ensure an effective stun within the 280 seconds of each cycle. The reduction in total pressure results in a gradual reduction of oxygen partial pressure that was measured by a solid-state electrochemical oxygen sensor. The reduced oxygen pressure leads to hypoxia, which is recognized as a humane method of stunning poultry. The system maintains an oxygen concentration of air always reduces the oxygen concentrations to a value lower than in dry air. The partial pressure of water and oxygen were found to depend on the pump down parameters due to the formation of fog in the chamber and desorption of water from the birds and the walls of the vacuum chamber. © The Author 2017. Published by Oxford University Press on behalf of Poultry Science Association.

  2. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  3. Correction of Pressure Drop in Steam and Water System in Performance Test of Boiler

    Science.gov (United States)

    Liu, Jinglong; Zhao, Xianqiao; Hou, Fanjun; Wu, Xiaowu; Wang, Feng; Hu, Zhihong; Yang, Xinsen

    2018-01-01

    Steam and water pressure drop is one of the most important characteristics in the boiler performance test. As the measuring points are not in the guaranteed position and the test condition fluctuation exsits, the pressure drop test of steam and water system has the deviation of measuring point position and the deviation of test running parameter. In order to get accurate pressure drop of steam and water system, the corresponding correction should be carried out. This paper introduces the correction method of steam and water pressure drop in boiler performance test.

  4. Supercritical water gasification with decoupled pressure and heat transfer modules

    KAUST Repository

    Dibble, Robert W.; Ng, Kim Choon; Sarathy, Mani

    2017-01-01

    decouples the function of containing high pressure from the high temperature function. The present invention allows the heat transfer function to be conducted independently from the pressure transfer function such that the system equipment can be designed

  5. Method of reactivity control in pressure tube reactor

    International Nuclear Information System (INIS)

    Fukumura, Nobuo.

    1988-01-01

    Purpose: To provide a method of controlling reactivity in a pressure tube reactor at high conversion ratio intended for high burn-up degree. Method: Control tubes are inserted in heavy water moderator. Light water is filled in the tubes at the initial burning stage. Along with the advance of the burning, the light water is gradually removed and replaced with gases of less reactive nuclear reactivity with neutrons such as air or gaseous carbon dioxide. The tubes are made of less neutron absorbing material such as aluminum. By filling light water, infinite multiplication factor is reduced to suppress the reactivity at the initial burning stage. As light water is gradually removed and replaced with air, etc., it provides an effect like that elimination of heavy water moderator to increase the conversion ratio. Accordingly, nuclear fission materials are produced additionally by so much to extend the burn-up degree. In this way, it can provide excellent effect in realizing high burn-up ratio and high conversion ratio. (Kamimura, M.)

  6. IFPE/IFA-533, Fuel Thermal Behaviour at High Burnup, Halden Reactor

    International Nuclear Information System (INIS)

    Gyori, Cs.; Turnbull, J.A.

    1997-01-01

    Description: After twelve years irradiation in the Halden Boiling Water Reactor two fuel rods (Rod 807 and Rod 808) were re-instrumented with fuel centre thermocouples and reloaded into the reactor in order to investigate the fuel thermal behaviour at high burnup. The fuel rods were pre-irradiated with four other rods in the upper cluster of IFA-409 (IFA=Instrumented Fuel Assembly) from May 1973 to June 1985. After base irradiation the four neighbouring rods were re-instrumented with pressure transducers and ramp tested in IFA-535.5 and IFA-535.6 providing useful data about fission gas release (FGR) presented in the Fuel Performance Database as well (Ref. 1). The two rods re-instrumented with fuel centre thermocouples have been irradiated as IFA-533.2 from April 1992. As the irradiation history of IFA-533.2 in the first months was very similar to the history of the ramp tests, the fuel temperature and FGR data measured in the different IFAs can complement each other, although the fuel-cladding gap sizes were slightly different and due to re-instrumentation the internal gas conditions were also dissimilar

  7. Performance of materials in the component cooling water systems of pressurized water reactors

    International Nuclear Information System (INIS)

    Lee, B.S.

    1993-01-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

  8. Modelling of high burnup structure in UO2 fuel with the RTOP code

    International Nuclear Information System (INIS)

    Likhanskii, V.; Zborovskii, V.; Evdokimov, I.; Kanyukova, V.; Sorokin, A.

    2008-01-01

    The present work deals with self-consistent physical approach aimed to derive the criterion of fuel restructuring avoiding correlations. The approach is based on study of large over pressurized bubbles formation on dislocations, at grain boundaries and in grain volume. At first, stage of formation of bubbles non-destroyable by fission fragments is examined using consistent modelling of point defects and fission gas behavior near dislocation and in grain volume. Then, evolution of formed large non-destroyable bubbles is considered using results of the previous step as initial values. Finally, condition of dislocation loops punching by sufficiently large over pressurized bubbles is regarded as the criterion of fuel restructuring onset. In the present work consideration of large over pressurized bubbles evolution is applied to modelling of the restructuring threshold depending on temperature, burnup and grain size. Effect of grain size predicted by the model is in qualitative agreement with experimental observations. Restructuring threshold criterion as an analytical function of local burnup and fuel temperature is derived and compared with HBRP project data. To predict rim-layer width formation depending on fuel burnup and irradiation conditions the model is implemented into the mechanistic fuel performance code RTOP. Calculated dependencies give upper estimate for the width of restructured region. Calculations show that one needs to consider temperature distribution within pellet which depends on irradiation history in order to model rim-structure formation

  9. Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

    International Nuclear Information System (INIS)

    El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

    2004-01-01

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

  10. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  11. A novel coordinated control for Integrated Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Zhao, Yuxin; Du, Xue; Xia, Genglei; Gao, Feng

    2015-01-01

    Highlights: • Proposed IPWR coordinated control strategy to avoid flow instability of OTSG. • Tuned PID controller parameters by Fuzzy kernel wavelet neural network with kernel trick and adaptive variable step-size. • Transition process exhibit the effectiveness of the novel IPWR control system. - Abstract: Integrated Pressurized Water Reactor (IPWR) has the characteristic of strong coupling, nonlinearity and complicated dynamic performance, which requires high standards of the control strategy and controller design. Most of IPWR systems utilize control strategy of ideal steady-state and PID controller, even though this strategy causes flow instability in the once through steam generator (OTSG) in low load conditions. Besides, the simple form of PID limits the performance developing which could not appropriately satisfy the requirements for quality. Motivated by these drawbacks, this paper proposes an IPWR coordinated control strategy and adopts PID controller to control each subsystem. The control strategy considers the system as a two-level hierarchical control system, and considers coordinating controller and bottom controllers. In the period of controller design, this strategy utilizes PID controller to control each subsystem, and modifies the controller parameters in real time by Fuzzy-KWNN algorithm, which adaptively achieves the system adjustment. Finally, simulation results are presented to exhibit the effectiveness of the proposed IPWR control system

  12. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  13. Modeling and simulation of pressurized water reactor power plant

    International Nuclear Information System (INIS)

    Wang, S.J.

    1983-01-01

    Two kinds of balance of plant (BOP) models of a pressurized water reactor (PWR) system are developed in this work - the detailed BOP model and the simple BOP model. The detailed model is used to simulate the normal operational performance of a whole BOP system. The simple model is used to combine with the NSSS model for a whole plant simulation. The trends of the steady state values of the detailed model are correct and the dynamic responses are reasonable. The simple BOP model approach starts the modelling work from the overall point of view. The response of the normalized turbine power and the feedwater inlet temperature to the steam generator of the simple model are compared with those of the detailed model. Both the steady state values and the dynamic responses are close to those of the detailed model. The simple BOP model is found adequate to represent the main performance of the BOP system. The simple balance of plant model was coupled with a NSSS model for a whole plant simulation. The NSSS model consists of the reactor core model, the steam generator model, and the coolant temperature control system. A closed loop whole plant simulation for an electric load perturbation was performed. The results are plausible. The coupling effect between the NSSS system and the BOP system was analyzed. The feedback of the BOP system has little effect on the steam generator performance, while the performance of the BOP system is strongly affected by the steam flow rate from the NSSS

  14. Fuzzy power control algorithm for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Y.J.; Lee, B.W.

    1994-01-01

    A fuzzy power control algorithm is presented for automatic reactor power control in a pressurized water reactor (PWR). Automatic power shape control is complicated by the use of control rods with a conventional proportional-integral-differential controller because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability needed for load-following operations including frequency control. In an attempt to achieve automatic power shape control without any design modifications to the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multiple-input multiple-output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of PWRs during the load-following operations

  15. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-01

    The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

  16. Thermal shield support degradation in pressurized water reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Fry, D.N.

    1986-01-01

    Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

  17. Current state of research on pressurized water reactor safety

    International Nuclear Information System (INIS)

    Couturier, Jean; Schwarz, Michel; Roubaud, Sebastien; Lavarenne, Caroline; Mattei, Jean-Marie; Rigollet, Laurence; Scotti, Oona; Clement, Christophe; Lancieri, Maria; Gelis, Celine; Jacquemain, Didier; Bentaib, Ahmed; Nahas, Georges; Tarallo, Francois; Guilhem, Gilbert; Cattiaux, Gerard; Durville, Benoit; Mun, Christian; Delaval, Christine; Sollier, Thierry; Stelmaszyk, Jean-Marc; Jeffroy, Francois; Dechy, Nicolas; Chanton, Olivier; Tasset, Daniel; Pichancourt, Isabelle; Barre, Francois; Bruna, Gianni; Evrard, Jean-Michel; Gonzalez, Richard; Loiseau, Olivier; Queniart, Daniel; Vola, Didier; Goue, Georges; Lefevre, Odile

    2018-03-01

    For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study - loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. -, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

  18. Remerschen nuclear power station with BBR pressurized water reactor

    International Nuclear Information System (INIS)

    Hoffmann, J.P.

    1975-01-01

    On the basis of many decades of successful cooperation in the electricity supply sector with the German RWE utility, the Grand Duchy of Luxemburg and RWE jointly founded Societe Luxembourgeoise d'Energie Nucleaire S.A. (SENU) in 1974 in which each of the partners holds a fifty percent interest. SENU is responsible for planning, building and operating this nuclear power station. Following an international invitation for bids on the delivery and turnkey construction of a nuclear power station, the consortium of the German companies of Brown, Boveri and Cie. AG (BBC), Babcock - Brown Boveri Reaktor GmbH (BBR) and Hochtief AG (HT) received a letter of intent for the purchase of a 1,300 MW nuclear power station equipped with a pressurized water reactor. The 1,300 MW station of Remerschen will be largely identical with the Muelheim-Kaerlich plant under construction by the same consortium near Coblence on the River Rhine since early 1975. According to present scheduling, the Remerschen nuclear power station could start operation in 1981. (orig.) [de

  19. Design and safety of the Sizewell pressurized water reactor

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the UK the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear power focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The long-term consequences can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking. (author)

  20. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  1. Numerical study of optimal equilibrium cycles for pressurized water reactors

    International Nuclear Information System (INIS)

    Mahlers, Y.P.

    2003-01-01

    An algorithm based on simulated annealing and successive linear programming is applied to solve equilibrium cycle optimization problems for pressurized water reactors. In these problems, the core reload scheme is represented by discrete variables, while the cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are treated as continuous variables. The enrichments are considered to be distinct in all feed fuel assemblies. The number of batches and their sizes are not fixed and also determined by the algorithm. An important feature of the algorithm is that all the parameters are determined by the solution of one optimization problem including both discrete and continuous variables. To search for the best reload scheme, simulated annealing is used. The optimum cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are determined for each reload pattern examined using successive linear programming. Numerical results of equilibrium cycle optimization for various values of the effective price of electricity and fuel reprocessing cost are studied

  2. Software to study the control strategy of pressurized water reactor

    International Nuclear Information System (INIS)

    Oliveira, Jose Ricardo de

    2002-01-01

    The computational program, result of this work, is a tool developed for the study of the control of Pressurized Water Reactors (PWR) constituted by only one coolant loop. The implementation of a user friendly interface for input/output data, makes the program also suitable for training and teaching applications. As design premise, it was considered enough just the modeling of the primary circuit, using as interface with the secondary circuit, a simplified differential equation of the temperature associated with the secondary power. All the incorporated dynamic equations to the model were developed using basic laws of conservation, boundary conditions and hypotheses appropriated to the control study. To arrive to the final model, core thermal and hydraulic characteristics and design data were obtained from of the available bibliography and adapted for a conceptual peculiar design of a small PWR. The whole program and all input/output interfaces were developed using the software Matlab, version 5.L Sub-routines of numeric integration based on the Runge-Kutta 4 method were applied, to solve the set of ordinary differential equations. (author)

  3. Evolution of general design requirements for french pressurized water reactors

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Rollinger, F.

    1988-10-01

    The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5

  4. Design description of the European pressurized water reactor

    International Nuclear Information System (INIS)

    Leverenz, R.

    1999-01-01

    The EPR (the European Pressurized Water Reactor) is an evolutionary PWR developed by Nuclear Power International and its parent companies, Framatome and Siemens, in co-operation with Electricite de France and German Utilities. NPI can rely on the huge experience gained by its parent companies; they have constructed more than 100 nuclear power plants throughout the world. The total installed capacity exceeds 100,000 MW - about 25% of the total world-wide figure. Following the conceptual design phase of the so-called Common Product conducted by NPI, Framatome and Siemens, from 1989 through 1991, Electricite de France (EDF) and several major German utilities decided to merge their own development programmes, - the N4 Plus and REP 2000 projects on the French side and the further development of the KONVOI technology on the German side, - with the NPI project. From that time on, the NPI project became one single common development line for both countries. In parallel, EDF and the German utilities decided to establish, together with other European utilities, specifications that would represent common utility views on the design and performance of future nuclear power plants. These are documented in the European Utility Requirements (EURs). The basic design has been completed in 1997, and in 1998 a design optimization is being carried out with the goal to even increase the economic competitiveness of nuclear power. This paper provides a brief design description of the EPR. (author)

  5. Contributions to the energetical role of advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1984-06-01

    Three articles written by the author in the past two years and all concerned with energy production of the Advanced Pressurized Water Reactor (APWR) are collated to form this report. The standpoint is made that the APWR using mixed oxide-fuel (about 8% Pusub(fiss) and Usub(depl)) would entail similar capital costs as a PWR, but with conversion rates of 0.85< C<0.95. Given this assumption the analysis shows that the high specific plutonium inventories (being proportional to the conversion ratio in the interested range) result in a strong damping effect on the growth rate, the effective uranium utilization, and on the total nuclear power level. Over one century an APWR strategy is 'far from equilibrium' and to describe this dynamic situation using static APWR data is not appropriate. If nuclear fission energy is to play a substantial and not just a marginal role in providing energy for the future (in the region of 5 - 8 TWsub(el) in 50 years time, corresponding to a share of 20 - 25% of the total world demand), clearly the fast breeder strategy offers the better chance of achieving the goal. (Auth.)

  6. High protein flexibility and reduced hydration water dynamics are key pressure adaptive strategies in prokaryotes

    KAUST Repository

    Martinez, N.

    2016-09-06

    Water and protein dynamics on a nanometer scale were measured by quasi-elastic neutron scattering in the piezophile archaeon Thermococcus barophilus and the closely related pressure-sensitive Thermococcus kodakarensis, at 0.1 and 40 MPa. We show that cells of the pressure sensitive organism exhibit higher intrinsic stability. Both the hydration water dynamics and the fast protein and lipid dynamics are reduced under pressure. In contrast, the proteome of T. barophilus is more pressure sensitive than that of T. kodakarensis. The diffusion coefficient of hydration water is reduced, while the fast protein and lipid dynamics are slightly enhanced with increasing pressure. These findings show that the coupling between hydration water and cellular constituents might not be simply a master-slave relationship. We propose that the high flexibility of the T. barophilus proteome associated with reduced hydration water may be the keys to the molecular adaptation of the cells to high hydrostatic pressure.

  7. High protein flexibility and reduced hydration water dynamics are key pressure adaptive strategies in prokaryotes

    KAUST Repository

    Martinez, N.; Michoud, Gregoire; Cario, A.; Ollivier, J.; Franzetti, B.; Jebbar, M.; Oger, P.; Peters, J.

    2016-01-01

    Water and protein dynamics on a nanometer scale were measured by quasi-elastic neutron scattering in the piezophile archaeon Thermococcus barophilus and the closely related pressure-sensitive Thermococcus kodakarensis, at 0.1 and 40 MPa. We show that cells of the pressure sensitive organism exhibit higher intrinsic stability. Both the hydration water dynamics and the fast protein and lipid dynamics are reduced under pressure. In contrast, the proteome of T. barophilus is more pressure sensitive than that of T. kodakarensis. The diffusion coefficient of hydration water is reduced, while the fast protein and lipid dynamics are slightly enhanced with increasing pressure. These findings show that the coupling between hydration water and cellular constituents might not be simply a master-slave relationship. We propose that the high flexibility of the T. barophilus proteome associated with reduced hydration water may be the keys to the molecular adaptation of the cells to high hydrostatic pressure.

  8. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1983-01-01

    Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt

  9. Determination of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Kristak, J.; Vobecky, M.

    1973-01-01

    Samples containing a known content of 235 U were irradiated with several different neutron doses and activities were determined of radionuclides including 125 Sb, 144 Ce, 134 Cs, 154 Eu, 103 Ru, 95 Zr. The values thus obtained were divided by the 137 Cs activity value. The resulting neutron dose-dependent value is plotted into a calibration graph. The degree of nuclear fuel burn-up is obtained from the graph using an experimentally determined ratio of the activities of the above radionuclides. (B.S.)

  10. High Burnup Fuel Performance and Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

    2007-03-15

    The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

  11. Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

    International Nuclear Information System (INIS)

    Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

    1983-01-01

    The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

  12. Spreading pressures of water and n-propanol on polymer surfaces

    NARCIS (Netherlands)

    Busscher, H.J.; Kip, Gerhardus A.M.; van Silfhout, Arend; Arends, J.

    1986-01-01

    Spreading pressures of water and n-propanol on polytetrafluoroethylene (PTFE), polystyrene (PS), polymethylmethacrylate (PMMA), polycarbonate (PC), and glass are determined from ellipsometrically measured adsorption isotherms by graphical integration, yielding for water 9, 37, 26, 33, and 141

  13. Technological and licensing challenges for high burnup fuel

    International Nuclear Information System (INIS)

    Gross, H.; Urban, P.; Fenzlein, C.

    2002-01-01

    Deregulation of electricity markets is driving electricity prices downward as well in the U.S. as in Europe. As a consequence high burnup fuel will be demanded by utilities using either the storage or the reprocessing option. At a minimum, burnups consistent with the current political enrichment limit of 5 w/o will be required for both markets.Significant progress has been achieved in the past by Siemens in meeting the demands of utilities for increased fuel burnup. The technological challenges posed by the increased burnup are mainly related to the corrosion and hydrogen pickup of the clad, the high burnup properties of the fuel and the dimensional changes of the fuel assembly structure. Clad materials with increased corrosion resistance appropriate for high burnup have been developed. The high burnup behaviour of the fuel has been extensively investigated and the decrease of thermal conductivity with burnup, the rim effect of the pellet and the increase of fission gas release with burnup can be described, with good accuracy, in fuel rod computer codes. Advanced statistical design methods have been developed and introduced. Materials with increased corrosion resistance are also helpful controlling the dimensional changes of the fuel assembly structure. In summary, most of the questions about the fuel operational behaviour and reliability in the high burnup range have been solved - some of them are still in the process of verification - or the solutions are visible. This fact is largely acknowledged by regulators too. The main licensing challenges for high burnup fuel are currently seen for accident condition analyses, especially for RIA and LOCA. (author)

  14. Actinides burnup in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

  15. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

  16. Modeling and diagnostic techniques applicable to the analysis of pressure noise in pressurized water reactors and pressure-sensing systems

    International Nuclear Information System (INIS)

    Mullens, J.A.; Thie, J.A.

    1984-01-01

    Pressure noise data from a PWR are interpreted by means of a computer-implemented model. The model's parameters, namely hydraulic impedances and noise sources, are either calculated or deduced from fits to data. Its accuracy is encouraging and raises the possibility of diagnostic assistance for nuclear plant monitoring. A number of specific applications of pressure noise in the primary system of a PWR and in a pressure sensing system are suggested

  17. On the pressure field of nonlinear standing water waves

    Science.gov (United States)

    Schwartz, L. W.

    1980-01-01

    The pressure field produced by two dimensional nonlinear time and space periodic standing waves was calculated as a series expansion in the wave height. The high order series was summed by the use of Pade approximants. Calculations included the pressure variation at great depth, which was considered to be a likely cause of microseismic activity, and the pressure distribution on a vertical barrier or breakwater.

  18. Living PSA issues in France on pressurized water reactors

    International Nuclear Information System (INIS)

    Dewailly, J.; Deriot, S.; Dubreuil Chambardel, A.; Francois, P.; Magne, L.

    1993-09-01

    Two Probabilistic Safety Assessments (PSAs) carried out in France on 900 and 1300 MWe Pressurized Water Reactor units ended in 1990. These PSAs determined the core damage frequency for all plant operating conditions ranging from cold shutdown for refuelling to full power operation. Since 1990, these PSAs have been used increasingly as tools for applications such as accident precursor analysis, risk-based Technical Specifications, and maintenance optimization. In turn, these applications are used to enhance the initial PSAs. The notion of a ''living'' PSA which can be used and updated is slowly taking form. The accident precursor analysis consists in applying PSA event trees to obtain quick information on the potential consequences of a precursor event and on the corresponding probabilities of occurrence. A feedback on PSAs is provided by comparing them with actual operating incidents. The computation of the allowed outage time during power operation, based on the computerized models of Probabilistic Safety Assessments, requires adjustments: calculation of hourly risk of core damage under different reactor conditions without equipment unavailabilities. The proposed method also turns out to be an aid in determining the safe shutdown condition and procedure. Furthermore, when introducing a sufficient level of detail, PSA reliability models make it possible to compute contributions and to perform sensitivity studies in order to highlight those components for which a maintenance effort should be made. From the experience acquired up to now, there was felt to be a strong need to create guidelines for using PSAs that would simplify their implementation by the experts in charge of determining Technical Specifications, of maintenance programs, etc. who are not generally specialists in PSAs. For this purpose, it is necessary to improve the intelligibility of the models made in order for them to be used and to offer user's guides adapted to each application. Documents

  19. Study on low pressure evaporation of fresh water generation system model

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Shik; Wibowo, Supriyanto; Shin, Yong Han; Jeong, Hyo Min [Gyeongsang National University, Tongyeong (Korea, Republic of); Fajar, Berkah [University of Diponegoro, Semarang (Indonesia)

    2012-02-15

    A low pressure evaporation fresh water generation system is designed for converting brackish water or seawater into fresh water by distillation in low pressure and temperature. Distillation through evaporation of feed water and subsequent vapor condensation as evaporation produced fresh water were studied; tap water was employed as feed water. The system uses the ejector as a vacuum creator of the evaporator, which is one of the most important parts in the distillation process. Hence liquid can be evaporated at a lower temperature than at normal or atmospheric conditions. Various operating conditions, i.e. temperature of feed water and different orifice diameters, were applied in the experiment to investigate the characteristics of the system. It was found that these parameters have a significant effect on the performance of fresh water generation systems with low pressure evaporation.

  20. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)