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Sample records for jen-upm pwr fuel

  1. Improving fuel-rod performance. [PWR; BWR

    SciTech Connect

    Ocken, H.; Knott, S.

    1981-03-01

    To reduce the risk of fuel-rod failures, utilities operate their nuclear reactors within conservative limits on power increases proposed by nuclear-fuel vendors. Of particular concern to US utilities is that adopting these limits results in an industrywide average plant capacity loss of 3% in BWR designs and 0.3% in PWR designs. To replace lost BWR capacity by other generating means currently costs the utilities $150 million annually, and losses for PWRs are about $20 million. Efforts are therefore being made to identify the factors responsible for Zircaloy degradation under PCI condition and to improve nuclear-fuel-rod design and reactor operation.

  2. PWR fuel features to preclude externally induced damage

    SciTech Connect

    Shallenberger, J.M.; Wilson, J.F.; Knott, R.P.

    1987-01-01

    Over the past several years there have been instances of pressurized water reactor (PWR) fuel damage attributed to factors external to the fuel. These externally induced causes include debris in the reactor coolant and baffle jetting. These causes of PWR fuel damage account for --50% of the total number of damaged rods. This paper discusses two features that significantly reduce the potential for fuel damage due to debris and baffle jetting. These two features are the debris filter bottom nozzle (DFBN) and the antivibration clip.

  3. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect

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

    2000-08-21

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

  4. A comprehensive in-pile test of PWR fuel bundle

    NASA Astrophysics Data System (ADS)

    Kang, Rixin; Zhang, Shucheng; Chen, Dianshan

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3 × 3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 μm. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation.

  5. Consolidation and disposal of PWR fuel inserts

    SciTech Connect

    Wakeman, B.H. )

    1992-08-01

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

  6. Timing analysis of PWR fuel pin failures

    SciTech Connect

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

    1992-09-01

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

  7. Experience in PWR and BWR mixed-oxide fuel management

    SciTech Connect

    Schlosser, G.J.; Krebs, W.; Urban, P. )

    1993-04-01

    Germany has adopted the strategy of a closed fuel cycle using reprocessing and recycling. The central issue today is plutonium recycling by the use of U-Pu mixed oxide (MOX) in pressurized water reactors (PWRs) and boiling water reactors (BWRs). The design of MOX fuel assemblies and fuel management in MOX-containing cores are strongly influenced by the nuclear properties of the plutonium isotopes. Optimized MOX fuel assembly designs for PWRs currently use up to three types of MOX fuel rods having different plutonium contents with natural uranium or uranium tailings as carrier material but without burnable absorbers. The MOX fuel assembly designs for BWRs use four to six rod types with different plutonium contents and Gd[sub 2]O[sub 3]/UO[sub 2] burnable absorber rods. Both the PWR and the BWR designs attain good burnup equivalence and compatibility with uranium fuel assemblies. High flexibility exists in the loading schemes relative to the position and number of MOX fuel assemblies in the reloads and in the core as a whole. The Siemens experience with MOX fuel assemblies is based on the insertion of 318 MOX fuel assemblies in eight PWRs and 168 in BWRs and pressurized heavy water reactors so far. The primary operating results include information on the cycle length, power distribution, reactivity coefficients, and control rod worth of cores containing MOX fuel assemblies.

  8. PWR and BWR spent fuel assembly gamma spectra measurements

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  9. PWR and BWR spent fuel assembly gamma spectra measurements

    SciTech Connect

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

    2016-07-17

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

  10. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGES

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

    2016-07-17

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

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

    SciTech Connect

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

    2013-07-01

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

  12. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

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

    SciTech Connect

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

    2001-04-04

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

  14. Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

    SciTech Connect

    Feltus, M.A.; Pozsgai, C. )

    1992-01-01

    Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative nature of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.

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

    SciTech Connect

    Naegeli, Robert Earl

    2004-06-01

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

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

    SciTech Connect

    Cardoni, Jeffrey

    2010-11-01

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

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

    SciTech Connect

    Bailey, W.J.; Tokar, M.

    1982-12-01

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

  18. THERMAL HISTORY OF CLADDING IN A 21 PWR WASTE PACKAGE LOADED WITH AVERAGE FUEL

    SciTech Connect

    H.M. Wade

    2000-01-25

    The purpose of this calculation is to evaluate a mid-assembly axial fuel cladding temperature profile of a 21 pressurized water reactor (PWR) spent nuclear fuel (SNF) waste package (WP) loaded with average fuel assemblies and emplaced in a monitored geologic repository. This calculation is intended to evaluate Viability Assessment (VA) and Enhanced Design Alternatives (EDA) II design configurations in support of performance assessment. This calculation was developed by Waste Package Operations (WPO) under Office of Civilian Radioactive Waste Management (OCRWM) procedure AP-3.12Q, Revision 0.

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

    SciTech Connect

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

    2014-09-30

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

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

  1. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    SciTech Connect

    Hori, Keiichi; Miyazaki, Keiji; Akiyama, Yoshiei; Nishioka, Hiromasa; Takeda, Naoki

    1996-08-01

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method.

  2. IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC

    NASA Astrophysics Data System (ADS)

    PetroviĆ, B. G.

    1991-01-01

    The following sections are included: * INTRODUCTION * IN-CORE FUEL MANAGEMENT CALCULATIONS * In-Core Fuel Management * Methodological Problems of In-Core Fuel Management * In-Core Fuel Management Analytical Tools * PENN STATE FUEL MANAGEMENT PACKAGE * Penn State Fuel Management Package (PFMP) * Assembly Data Description (ADD) * Linking PSU-LEOPARD and MCRAC: An Example * MULTICYCLE REACTOR ANALYSIS CODE (MCRAC) * Main Features and Options of MCRAC code * Core geometry * Diffusion equations * 1.5-group model * Multicycle neutronic analysis * Multicycle cost analysis * Criticality search * Power-dependent xenon feedback calculations * Control rod and burnable absorber simulation * Search for LP with flat BOC power distribution * Artificial ADD option * Variable dimensioning technique * RBI version of MCRAC code * Programming changes in PC version * Fuel interchange option * MCRAC Input/Output * General input description * Sample input * Sample output * EXPERIENCE WITH MCRAC CODE * CONCLUSIONS * REFERENCES

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

    SciTech Connect

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

    2015-09-30

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

  4. Application of the RCP01 Code to Depletion of a PWR Spent Nuclear Fuel Sample

    SciTech Connect

    Joo, Hansem

    2002-01-01

    An essential component of a proposed burnup credit methodology for commercial PWR spent nuclear fuel (SNF) is the validation of the tools used for isotopic and criticality calculations. A number of benchmark experiments have been analyzed to establish the validation of the tools and to determine biases and corrections. To benchmark the RCP01 Monte Carlo computer code, an isotopic validation study was conducted for one of the benchmark experiments, a SNF sample taken from the Calvert Cliffs PWR Unit-1 (CCPU1). Modeling considerations and nuclear data associated with the RCP01 transport/depletion calculations are discussed. The accuracy of RCP01 calculations is demonstrated to be very good when RCP01 results are compared to destructive chemical assay data for major actinides and important fission products in the SNF sample.

  5. Pressure-vessel-damage fluence reduction by low-leakage fuel management. [PWR

    SciTech Connect

    Cokinos, D.; Aronson, A.L.; Carew, J.F.; Kohut, P.; Todosow, M.; Lois, L.

    1983-01-01

    As a result of neutron-induced radiation damage to the pressure vessel and of an increased concern that in a PWR transient the pressure vessel may be subjected to pressurized thermal shock (PTS), detailed analyses have been undertaken to determine the levels of neutron fluence accumulation at the pressure vessels of selected PWR's. In addition, various methods intended to limit vessel damage by reducing the vessel fluence have been investigated. This paper presents results of the fluence analysis and the evaluation of the low-leakage fuel management fluence reduction method. The calculations were performed with DOT-3.5 in an octant of the core/shield/vessel configuration using a 120 x 43 (r, theta) mesh structure.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

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

  7. A Study on the Conceptual Design of a 1,500 MWe Passive PWR with Annular Fuel

    SciTech Connect

    Kwi Lim Lee; Soon Heung Chang

    2004-07-01

    In this study, the preliminary conceptual design of a 1500 MWe pressurized water reactor (PWR) with annular fuel has been performed. This design is derived from the AP1000 which is a 1000 MWe PWR with two-loop. However, the present design is a 1500 MWe PWR with three-loop, passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. The preliminary design parameters of this reactor have been determined through simple relation to those of AP1000 for reactor, reactor coolant system, and passive safety injection system. Using the MATRA code, we analyze the core designs for two alternatives on fuel assembly types: solid fuel and annular fuel. The performance of reactor cooling systems is evaluated through the accident of the cold leg break in the core makeup tank loop by using MARS2.1 code. This study presents the developmental strategy, preliminary design parameters and safety analysis results. (authors)

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

    SciTech Connect

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

    1989-02-01

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

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

    SciTech Connect

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

    2010-03-01

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

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

    SciTech Connect

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

    2016-06-10

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

  11. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

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

    DOE PAGES

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

    2016-06-10

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

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

    SciTech Connect

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

    1980-01-01

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

  14. Modeling and design of a reload PWR core for a 48-month fuel cycle

    SciTech Connect

    McMahon, M.V.; Driscoll, M.J.; Todreas, N.E.

    1997-05-01

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

  15. Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

    NASA Astrophysics Data System (ADS)

    Burdo, James S.

    This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the

  16. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  17. Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

    SciTech Connect

    Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

    2012-07-01

    The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

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

    SciTech Connect

    Crowell, Shannon L.; Alzheimer, James M.

    2011-08-08

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

  19. Analysis of MERCI decay heat measurement for PWR UO{sub 2} fuel rod

    SciTech Connect

    Jaboulay, J.C.; Bourganel, S.

    2012-01-15

    Decay heat measurements, called the MERCI experiment, were conducted at Commissariat a l'Energie Atomique (CEA)/Saclay to characterize accurately residual power at short cooling time and verify its prediction by decay code and nuclear data. The MOSAIC calorimeter, developed and patented by CEA/Grenoble (DTN/SE2T), enables measurement of the decay heat released by a pressurized water reactor (PWR) fuel rod sample between 200 and 4 W within a precision of 1%. The MERCI experiment included three phases. At first, a UO{sub 2} fuel rod sample was irradiated in the CEA/Saclay experimental reactor OSIRIS. The burnup achieved at the end of irradiation was similar to 3.5 GWd/tonne. The second phase was the transfer of the fuel rod sample from its irradiation location to a hot cell, to be inserted inside the MOSAIC calorimeter. It took 26 min to carry out the transfer. Finally, decay heat released by the PWR sample was measured from 27 min to 42 days after shutdown. Post irradiation examinations were performed to measure concentrations of some heavy nuclei (U, Pu) and fission products (Cs, Nd). The decay heat was predicted using a calculation scheme based on the PEPIN2 depletion code, the TRIPOLI-4 Monte Carlo code, and the JEFF3.1.1 nuclear data file. The MERCI experiment analysis shows that the discrepancy between the calculated and the experimental decay heat values is included between -10% at 27 min and +6% at 12 h, 30 min otter shutdown. From 4 up to 42 days of cooling time, the difference between calculation and measurement is about ± 1%, i.e., experimental uncertainty. The MERCI experiment represents a significant contribution for code validation; the time range above 10{sup 5} s has not been validated previously. (authors)

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

    SciTech Connect

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

    1995-03-01

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

  1. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    SciTech Connect

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

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

    SciTech Connect

    Ade, Brian J; Gauld, Ian C

    2011-10-01

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

  3. High mechanical performance of Areva upgraded fuel assemblies for PWR in USA

    SciTech Connect

    Gottuso, Dennis; Canat, Jean-Noel; Mollard, Pierre

    2007-07-01

    The merger of the product portfolios of the former Siemens and Framatome fuel businesses gave rise to a new family of PWR products which combine the best features of the different technologies to enhance the main performance of each of the existing products. In this way, the technology of each of the three main fuel assembly types usually delivered by AREVA NP, namely Mark-BW{sup TM}, HTP{sup TM} and AFA 3G{sup TM} has been enriched by one or several components from the others which contributes to improve their robustness and to enhance their performance. The combined experience of AREVA's products shows that the ROBUST FUELGUARD{sup TM}, the HMP{sup TM} end grid, the MONOBLOC{sup TM} guide tube, a welded structure, M5{sup R} material for every zirconium component and an upper QUICK-DISCONNECT{sup TM} are key features for boosting fuel assembly robustness. The ROBUST FUELGUARD benefits from a broad experience demonstrating its high efficiency in stopping debris. In addition, its mechanical strength has been enhanced and the proven blade design homogenizes the downstream flow distribution to strongly reduce excitation of fuel rods. The resistance to rod-to-grid fretting resistance of AREVA's new products is completed by the use of a lower HMP grid with 8 lines of contact to insure low wear. The Monobloc guide tube with a diameter maximized to strengthen the fuel assembly stiffness, excludes through its uniform outer geometry any local condition which could weaken guide tube straightness. The application of a welded cage to all fuel assemblies of the new family of products in combination with stiffer guide tubes and optimized hold-down assures each fuel assembly enhanced resistance to distortion. The combination of these features has been widely demonstrated as an effective method to reduce the risk of incomplete RCCA insertion and significantly reduce assembly distortion. Thanks to its enhanced performance, M5 alloy insures that all fuel assemblies in the family

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

    SciTech Connect

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

    2000-09-01

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

  5. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect

    Bi, G.; Liu, C.; Si, S.

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no

  6. Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

    SciTech Connect

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

    2015-07-01

    A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

  7. PWR core and spent fuel pool analysis using scale and nestle

    SciTech Connect

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  8. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    SciTech Connect

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  9. Evaluation of storing Shippingport Core II spent blanket fuel assemblies in the T Plant PWR Core II fuel pool without active cooling

    SciTech Connect

    Gilbert, E.R.; Lanning, D.D.; Dana, C.M.; Hedengren, D.C.

    1994-10-01

    PWR Core II fuel pool chiller-off test was conducted because it appeared possible that acceptable pool-water temperatures could be maintained without operating the chillers, thus saving hundreds of thousands of dollars in maintenance and replacement costs. Test results showed that the water-cooling capability is no longer needed to maintain pool temperature below 38{degrees}C (100{degrees}F).

  10. Investigation of the Effect of Fixed Absorbers on the Reactivity of PWR Spent Nuclear Fuel for Burnup Credit

    SciTech Connect

    Wagner, John C.; Sanders, Charlotta E.

    2002-08-15

    The effect of fixed absorbers on the reactivity of pressurized water reactor (PWR) spent nuclear fuel (SNF) in support of burnup-credit criticality safety analyses is examined. A fuel assembly burned in conjunction with fixed absorbers may have a higher reactivity for a given burnup than an assembly that has not used fixed absorbers. As a result, guidance on burnup credit, issued by the U.S. Nuclear Regulatory Commission's Spent Fuel Project Office, recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommendation eliminates a large portion of the currently discharged SNF from loading in burnup credit casks and thus severely limits the practical usefulness of burnup credit. Therefore, data are needed to support the extension of burnup credit to additional SNF. This research investigates the effect of various fixed absorbers, including integral burnable absorbers, burnable poison rods, control rods, and axial power shaping rods, on the reactivity of PWR SNF. Trends in reactivity with relevant parameters (e.g., initial fuel enrichment, burnup and absorber type, exposure, and design) are established, and anticipated reactivity effects are quantified. Where appropriate, recommendations are offered for addressing the reactivity effects of the fixed absorbers in burnup-credit safety analyses.

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

    SciTech Connect

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

    2006-10-10

    A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the

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

    SciTech Connect

    None, None

    1997-04-01

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

  13. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    NASA Astrophysics Data System (ADS)

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  14. International comparison of a depletion calculation benchmark devoted to fuel cycle issues results from the phase 1 dedicated to PWR-UOx fuels

    SciTech Connect

    Roque, B.; Kilger, R.; Laugier, F.; Marimbeau, P.; Riffard, C.; Thro, J. F.; Yudkevich, M.; Hesketh, K.; Sartori, E.

    2006-07-01

    This paper presents the results from the first phase of an international depletion calculations comparison devoted to PWR-UOx fuel cycle issues. This 'benchmark' has been defined within the NEA/OECD Working Party on Scientific Issues in Reactors Systems (WPRS). The aim is to investigate a large range of isotopes, physics quantities and fuel types applied to fuel and back-end cycle configurations. The results analyses have shown that there is a good agreement between participants for the mass calculation of many isotopes. However, it is interesting to observe that better agreement is obtained for isotopes which benefit from experimental validation. In this benchmark, the poorest agreement is obtained in calculating activation products originating from fuel impurities. Some discrepancies on neutron emission rates were also observed, mainly due to the discrepancies on masses calculations. Good agreement was obtained for the total decay heat calculation. (authors)

  15. Parametric study of CHF data. Volume 2. A generalized subchannel CHF correlation for PWR and BWR fuel assemblies. Final report

    SciTech Connect

    Reddy, D.G.; Fighetti, C.F.

    1983-01-01

    The primary objective of this research was to develop a generalized subchannel CHF correlation based on the local fluid conditions obtained with the COBRA-IIIC thermal hydraulic subchannel code and covering PWR and BWR normal operating conditions as well as hypothetical loss-of-coolant accident (LOCA) conditions. In view of the importance of the local conditions predicted by the COBRA-IIIC code in the development of CHR correlation, the secondary objective was to improve the predictive capability of the COBRA-IIIC subchannel code. In the first phase of this study, the sensitivity of local enthalpies and local mass fluxes predicted by the COBRA-IIIC subchannel code to subcooled void correlation, bulk void correlation, two-phase friction multiplier correlation and turbulent mixing parameter was determined. In the second phase, based on the local conditions obtained with the COBRA-IIIC subchannel code, an accurate generalized subchannel CHF correlation was developed utilizing 3607 CHF data points from 65 test sections simulating PWR and BWR fuel assemblies.

  16. Criticality evaluation of control component credited mixed zone spent and fresh fuel storage in high density PWR racks

    SciTech Connect

    Bilovsky, V.; Redmond, E.; Walker, C.; Ivanov, K.

    2006-07-01

    To expand the set of assemblies that qualify for storage in high-density racks, a mixed zone analysis may be performed where repeating pattern configurations within the rack are prescribed. In a mixed zone analysis, assemblies that are more reactive (low burnup) are stored adjacent to less reactive (highly burned) assemblies, thereby meeting the same overall criticality requirements as with the uniform burnup/enrichment analysis. The Arkansas Nuclear One (ANO) Plant has faced several challenges with respect to their spent fuel storage that reach beyond simply the number of spent fuel assemblies and available storage cells. These issues have resulted in the need for ANO to use an advanced storage strategy. In addition to using the mixed zone burnup approach in the high-density racks, ANO also proposed a new solution involving credit for control components in the spent fuel pool. ANO submitted an amendment of their spent fuel pool technical specifications to the Nuclear Regulatory Commission (NRC) based on the evaluation performed by Holtec International that was subsequently approved. This paper presents a description of the overall methodology used for supporting the submittal, and provides further discussion regarding the reactivity effect of control rods in a PWR spent fuel pool. (authors)

  17. Scoping design analyses for optimized shipping casks containing 1-, 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect

    Bucholz, J.A.

    1983-01-01

    This report details many of the interrelated considerations involved in optimizing large Pb, Fe, or U-metal spent fuel shipping casks containing 1, 2, 3, 5, 7, or 10-year-old PWR fuel assemblies. Scoping analyses based on criticality, shielding, and heat transfer considerations indicate that some casks may be able to hold as many as 18 to 21 ten-year-old PWR fuel assemblies. In the criticality section, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. In another section, details of many n/..gamma.. shielding optimization studies are presented, including the optimal n/..gamma.. design points and the actual shielding requirements for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. Multigroup source terms based on ORIGEN2 calculations at these and other decay times are also included. Lastly, the numerical methods and experimental correlations used in the steady-state and transient heat transfer analyses are fully documented, as are pertinent aspects of the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. (While only casks for square, intact PWR fuel assemblies were considered in this study, the SCOPE code may also be used to design and analyze casks containing canistered spent fuel or other waste material. An abbreviated input data guide is included as an appendix).

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

    NASA Astrophysics Data System (ADS)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

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

  19. Summary report on optimized designs for shipping casks containing 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect

    Bucholz, J.A.

    1983-04-01

    The purpose of this study was to develop new conceptual designs for large Pb, Fe, and U-shielded spent fuel casks which have been optimized for the shipment of 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel assemblies. Design specifications for about 100 cases of potential interest are presented along with a brief 20-page synopsis of the associated analyses. Optimized shielding requirements are presented for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. With respect to criticality, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. Steady-state and transient heat transfer analyses for casks under nominal and accident conditions were performed using the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. Based on criticality, shielding, and heat transfer considerations, it appears that optimized cask designs could be developed to carry 15 to 18 five-year-old PWR fuel assemblies or as many as 18 to 21 ten-year-old PWR fuel assemblies. 4 figures, 4 tables.

  20. Improvement of the thermal margins in the Swedish Ringhals-3 PWR by introducing new fuel assemblies with thorium

    SciTech Connect

    Lau, C. W.; Demaziere, C.; Nylen, H.; Sandberg, U.

    2012-07-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)

  1. Rod consolidation of RG and E's (Rochester Gas and Electric Corporation) spent PWR (pressurized water reactor) fuel

    SciTech Connect

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister.

  2. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    SciTech Connect

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  3. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    SciTech Connect

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-07-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

  4. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    SciTech Connect

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  5. On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

    NASA Astrophysics Data System (ADS)

    Restani, R.; Horvath, M.; Goll, W.; Bertsch, J.; Gavillet, D.; Hermann, A.; Martin, M.; Walker, C. T.

    2016-12-01

    Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

  6. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

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

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  8. Surveillance of PLUS7{sup TM} fuel for PWR nuclear power plant

    SciTech Connect

    Jang, Y. K.; Kim, J. I.; Shin, J. C.; Chung, J. G.; Chung, S. K.; Kim, M. S.; Lee, T. H.; Yoon, Y. B.; Kim, T. W.

    2012-07-01

    The surveillance program on the advanced nuclear fuel of PLUS{sup TM} developed for Optimized Power Reactors of 1000 MWe (OPR1000s) and Advanced Power Reactors of 1400 MWe (APR1400s) in Korea was completed in the early of 2011. This fuel had been jointly developed through the extensive out-of-pile tests with Westinghouse for three years since 1999. The irradiation tests for the in-reactor verification using four lead test assemblies (LTAs) had been started in Ulchin unit 3 in 2002. During the overhaul period after each irradiation test, the eight (8) burnup-dependent parameters were measured without disassembling using the precise measurement systems in pool-side. After three cycle irradiations, one test assembly was disassembled and the rod-wise inspection on twenty rods was performed. During this stage, five (5) parameters were measured and evaluated. Among these twenty rods, ten rods including skeleton were sent to hot-cell test facility for further detailed examination and are currently being examined. After in-reactor verifications during two cycles, this fuel was commercially supplied to eight (8) OPR1000s sequentially. Currently all eight (8) OPR1000s were replaced with this fuel. In addition, this fuel is going to be supplied to four (4) APR1400s being constructed in Braka, UAE as well as four(4) OPR1000s and four(4) APR1400s being constructed in Korea. In the meanwhile, the surveillance program for the commercially supplied fuel has been launched to confirm growth, creep, corrosion and deformation, etc. obtained during LTA irradiation. Four (4) limiting fuel assemblies, that is, two (2) assemblies to be discharged after 2 cycle irradiations and the other two (2) after 3 cycle irradiations were selected for this surveillance program. Irradiation data of commercially supplied fuels are compared and confirmed to LTA irradiation performance behaviors on this paper. Among the eight (8) burnup-dependent parameters, the interesting ones were irradiation

  9. THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

    SciTech Connect

    Croft, Stephen; Favalli, Andrea; Swinhoe, Martyn T.

    2012-06-19

    Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to a particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence of Gd

  10. Experimental Test Plan for PWR Sister Rods in the High Burnup Spent Fuel Data Project

    SciTech Connect

    Montgomery, Rose; Scaglione, John M; Bevard, Bruce Balkcom; Hanson, Brady; Billone, Dr. Michael

    2016-01-01

    The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.

  11. Uranium resource utilization improvements in the once-through PWR fuel cycle

    SciTech Connect

    Matzie, R A

    1980-04-01

    In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U/sub 3/O/sub 8/ consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout.

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

    NASA Astrophysics Data System (ADS)

    Walter, Daniel John

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

  13. In-situ measurement of the effect of LiOH on the stability of fuel cladding oxide film in simulated PWR primary water environment

    SciTech Connect

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

    1995-12-31

    Development of new improved fuel cladding materials is a long process, partly because of the lack of fast and reliable in-situ techniques for investigations of cladding degradation in high temperature water environments. This paper describes results gained with the Contact Electric Resistance (CER) technique on the electric resistance of oxides growing on zirconium based fuel cladding materials. LiOH decreased the electric resistance of the oxides when about 70 ppm was injected in PWR water at 300 C. When PWR water contains boric acid and LiOH from the beginning of the exposure the fuel cladding material is covered by a hydroxide layer that protects the amorphous oxide layer and later hinders the increase of the resistance of the crystalline oxide layer. The dependency of electric resistance of the oxides on LiOH concentration is shown to correlate inversely with the effect of LiOH on weight gain. The kinetics of the breakdown process of electric resistance indicate that a phase transformation rather than a diffusion limited process is the mechanism of degradation. The growth rate of the electric resistance of the oxide in the early stage of oxide formation is shown to correlate well with the in-reactor weight gain of similar alloys. In-situ monitoring of the electric resistance of the oxide during growth is shown to give the same ranking order as long term in-reactor weight gain tests, but in a fraction of the testing time needed for weight gain tests.

  14. Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly

    NASA Astrophysics Data System (ADS)

    da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2014-04-01

    Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

  15. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

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

    1987-10-01

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

  16. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

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

  17. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    SciTech Connect

    Foad, Basma; Takeda, Toshikazu

    2015-12-31

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  18. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    NASA Astrophysics Data System (ADS)

    Foad, Basma; Takeda, Toshikazu

    2015-12-01

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO2 and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  19. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    SciTech Connect

    Lafleur, Adrienne M; Charlton, William S; Menlove, Howard O; Swinhoe, Martyn T

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  20. Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

    NASA Astrophysics Data System (ADS)

    Riglet-Martial, Ch.; Sercombe, J.; Lamontagne, J.; Noirot, J.; Roure, I.; Blay, T.; Desgranges, L.

    2016-11-01

    The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that irradiated fuels (of nominal stoichiometry close to 2.000) under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U4O9 type in its cold section, of lower temperature. The parameters governing the oxidation states of UO2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

  1. Direct Experimental Evaluation of the Grain Boundaries Gas Content in PWR fuels: New Insight and Perspective of the ADAGIO Technique

    SciTech Connect

    Pontillon, Y.; Noirot, J.; Caillot, L.

    2007-07-01

    Over the last decades, many analytical experiments (in-pile and out-of-pile) have underlined the active role of the inter-granular gases on the global fuel transient behavior under accidental conditions such as RIA and/or LOCA. In parallel, the improvement of fission gas release modeling in nuclear fuel performance codes needs direct experimental determination/validation regarding the local gas distribution inside the fuel sample. In this context, an experimental program, called 'ADAGIO' (French acronym for Discriminating Analysis of Accumulation of Inter-granular and Occluded Gas), has been initiated through a joint action of CEA, EDF and AREVA NP in order to develop a new device/technique for quantitative and direct measurement of local fission gas distribution within an irradiated fuel pellet. ADAGIO technique is based on the fact that fission gas inventory (intra and inter-granular parts) can be distinguished by controlled fuel oxidation, since grain boundaries oxidize faster than the bulk. The purpose of the current paper is to present both the methodology and the associated results of the ADAGIO program performed at CEA. It has been divided into two main parts: (i) feasibility (UO{sub 2} and MOX fuels), (ii) application on high burn up UO{sub 2} fuel. (authors)

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

    SciTech Connect

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

    2001-11-15

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

  3. Implementation of the active neutron Coincidence Collar for the verification of unirradiated PWR and BWR fuel assemblies

    SciTech Connect

    Menlove, H.O.; Keddar, A.

    1982-01-01

    An active neutron interrogation technique has been developed for the measurement of the /sup 235/U content in fresh fuel assemblies. The method employs an AmLi neutron source to induce fission reactions in the fuel assembly and coincidence counting of the resulting fission reaction neutrons. When no interrogation source is present, the passive neutron coincidence rate gives a measure of the /sup 238/U by the spontaneous fission reactions. The system can be applied to the fissile content determination in fresh fuel assemblies for accountability, criticality control, and safeguards purposes. Field tests have been performed by International Atomic Energy Agency (IAEA) staff using the Coincidence Collar to verify the /sup 235/U content in light-water-reactor fuel assemblies. The results gave an accuracy of 1 to 2% in the active mode (/sup 235/U) and 2 to 3% in the passive mode (/sup 238/U) under field conditions.

  4. In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

    SciTech Connect

    Swinhoe, Martyn Thomas; De Baere, Paul

    2015-04-17

    A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hour of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.

  5. Space and Time Distribution of Pu Isotopes inside The First Experimental Fuel Pin Designed for PWR and Manufactured in Indonesia

    NASA Astrophysics Data System (ADS)

    Suwardi; Setiawan, J.; Susilo, J.

    2017-01-01

    The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared and planned to be tested in power ramp irradiation. An irradiation test should be designed to allow an experiment can be performed safely and giving maximum results of many performance aspects of design and manufacturing. Performance analysis to the fuel specimen shows that the specimen is not match to be used for power ramp testing. Enlargement by 0.20 mm of pellet diameter has been proposed. The present work is evaluation of modified design for important aspect of isotopic Pu distribution during irradiation test, because generated Pu isotopes in natural UO2 fuel, contribute more power relative to the contribution by enriched UO2 fuel. The axial profile of neutrons flux have been chosen from both experimental measurement and model calculation. The parameters of ramp power has been obtained from statistical experiment data. A simplified and typical base-load commercial PHWR profile of LHR history has been chosen, to determine the minimum irradiation time before ramp test can be performed. The data design and Mat pro XI materials properties models have been chosen. The axial profile of neutrons flux has been accommodated by 5 slices of discrete pin. The Pu distribution of slice-4 with highest power rate has been chosen to be evaluated. The radial discretion of pellet and cladding and numerical parameter have been used the default best practice of TU. The results shows that Pu 239 increased rapidly. The maximum burn up of slice 4 at upper the median slice, it reached nearly 90% of maximum value at about 6000 h with peak of 0.8%a Pu/HM at 22000 h, which is higher than initial U 235. Each 240, 241 and 240 Pu grows slower and ends up to 0.4, 0.2 and 0.18 % respectively. This results can be used for verification of other aspect of fuel behavior in the modeling results and also can be used as guide and comparison to the future post irradiation examination for Pu isotopes distribution.

  6. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    NASA Astrophysics Data System (ADS)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  7. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    SciTech Connect

    Suwardi; Dewayatna, W.; Briyatmoko, B.

    2012-06-06

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK{center_dot}CEN and Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  8. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    NASA Astrophysics Data System (ADS)

    Suwardi, Dewayatna, W.; Briyatmoko, B.

    2012-06-01

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK.CEN & Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott [2]. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  9. Impact of boron dilution accidents on low boron PWR safety

    SciTech Connect

    Papukchiev, A.; Liu, Y.; Schaefer, A.

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As an inadvertent reduction of the boron concentration during a boron dilution accident could introduce positive reactivity and have a negative impact on PWR safety, design changes to reduce boron concentration in the reactor coolant are of general interest. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) load has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) to 518 ppm. For the assessment of the potential safety advantages, a boron dilution accident due to small break loss-of-coolant-accident (SBLOCA) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The results from the comparative analyses showed that the impact of the boron dilution accident on the new PWR design safety is significantly lower in comparison with the standard design. The new reactor design provided at least 4, 4% higher reactivity margin to recriticality during the whole accident which is equivalent to the negative reactivity worth of additional 63% of all control rods fully inserted in to the core. (authors)

  10. Influence Of Low Boron Core Design On PWR Transient Behavior

    SciTech Connect

    Aleksandrov Papukchiev, Angel; Yubo Liu; Schaefer, Anselm

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, the concentration of boron in primary coolant is limited by the requirement of having a negative moderator density coefficient. As high boron concentrations have significant impact on reactivity feedback properties, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) content has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) by approx. 50% compared to current German PWR technology. For the assessment of the potential safety advantages, a Loss-of-Feedwater Anticipated Transient Without Scram (ATWS LOFW) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The most significant difference in the transient performance of both designs is the total primary fluid mass released through the pressurizer (PRZ) valves. It is reduced by a factor of four for the low boron reactor, indicating its improved density reactivity feedback. (authors)

  11. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    SciTech Connect

    M. Krug; R. Shogan; A. Fero; M. Snyder

    2004-11-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR.

  12. PWR representative behavior during a LOCA

    SciTech Connect

    Allison, C.M.

    1981-01-01

    To date, there has been substantial analytical and experimental effort to define the margins between design basis loss-of-coolant accident (LOCA) behavior and regulatory limits on maximum fuel rod cladding temperature and deformation. As a result, there is extensive documentation on the modeling of fuel rod behavior in test reactors and design basis LOCA's. However, modeling of that behavior using representative, non-conservative, operating histories is not nearly as well documented in the public literature. Therefore, the objective of this paper is (a) to present calculations of LOCA induced behavior for Pressurized Water Reactor (PWR) core representative fuel rods, and (b) to discuss the variability in those calculations given the variability in fuel rod condition at the initiation of the LOCA. This analysis was limited to the study of changes in fuel rod behavior due to different power operating histories. The other two important parameters which affect that behavior, initial fuel rod design and LOCA coolant conditions were held invarient for all of the representative rods analyzed.

  13. WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

    SciTech Connect

    J.W. Davis

    1996-07-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so.

  14. A PWR Thorium Pin Cell Burnup Benchmark

    SciTech Connect

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

    2000-05-01

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

  15. Update on the PWR axial burnup profile database

    SciTech Connect

    Cacciapouti, R.F.; Volkinburg, S.V.

    1995-12-01

    A pressurized water reactor database was developed to evaluate the axial burnup profiles of various reactor types. The data showed that the various types exhibit similar behavior, especially at the top and bottom of the assembly. From the existing data, bounding axial burnup profiles can be developed to envelope the various pressurized water reactor assembly deigns. The database encompasses most of the PWR fuel designs and contains sufficient data to provide reliable statistics.

  16. Evaluation of zinc addition to PWR primary coolant

    SciTech Connect

    Pathania, R.; Yagnik, S.; Gold, R.E.; Dove, M.; Kolstad, E.

    1995-12-31

    Laboratory studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. Because of the potential benefits of zinc addition in reducing radiation fields and mitigating PWSCC of Alloy 600 a project was initiated to qualify zinc addition to a PWR. The objective of this work was to evaluate the effect of zinc addition on radiation fields, PWSCC of Alloy 600 and fuel cladding corrosion at the Farley-2 PWR. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor prior to zinc addition at Farley-2. This paper provides an overview of the scope of the zinc addition demonstration at Farley-2 and the fuel cladding corrosion tests at Halden. The zinc concentration in the Farley-2 coolant is approximately 40 ppb and that in Halden is 50 ppb. The paper presents initial results from these studies which are still in progress.

  17. PWR Cross Section Libraries for ORIGEN-ARP

    SciTech Connect

    McGraw, Carolyn; Ilas, Germina

    2012-01-01

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

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

    SciTech Connect

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

    2012-06-06

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

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

    NASA Astrophysics Data System (ADS)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

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

    SciTech Connect

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

    2006-07-01

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

  1. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    SciTech Connect

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  2. Concept of Small Sized Integrated PWR with Double Pressure Vessels

    SciTech Connect

    Kinoshita, I.; Ueda, N.; Nishi, Y.; Matsumura, T.

    2002-07-01

    For early deployment of small sized nuclear reactors, it is better to reduce the BOP cost with new ideas than introducing innovative technologies for core, fuel and materials. In this report, a concept of the integrated, forced convective and small PWR with double pressure vessels has been proposed. The electric output of this reactor is 150 MW. Conventional technologies are adopted for core and fuel. Refueling, maintenance and repairing are made in a special ship with complete facilities and skilled experts. The pressure vessel with the core, control rod drive mechanisms (CRDM), main circulating pumps (MCP), steam generators (SG) and other reactor internals are transferred between the reactor building and the ship. Technical feasibility for safety and maintainability has been discussed qualitatively. The construction cost has been roughly estimated. (authors)

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

    SciTech Connect

    2013-11-01

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

  4. Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

    SciTech Connect

    Fridman, E.; Shwageraus, E.; Galperin, A.

    2006-07-01

    The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of this work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)

  5. Annual report, FY 1979 Spent fuel and fuel pool component integrity.

    SciTech Connect

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.

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

    SciTech Connect

    McGraw, C.; Ilas, G.

    2012-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-06-01

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

  8. New results from the NSRR experiments with high burnup fuel

    SciTech Connect

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  9. Horizontal Drop of 21- PWR Waste Package

    SciTech Connect

    A.K. Scheider

    2007-01-31

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design.

  10. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    SciTech Connect

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-12-31

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR`s) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design.

  11. Integrated Radiation Transport and Thermo-Mechanics Simulation of a PWR Assembly

    SciTech Connect

    Clarno, Kevin T; Hamilton, Steven P; Philip, Bobby; Sampath, Rahul S; Allu, Srikanth; Berrill, Mark A; Barai, Pallab; Banfield, James E

    2012-01-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step towards incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source terms, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses, of traditional (single-pin) nuclear fuel performance simulation. AMPFuel was used to model an entire 17 x 17 Pressurized Water Reactor (PWR) fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins, the 25 guide tubes, top and bottom structural regions, and the upper and lower (neutron) reflector regions. The final full-assembly calculation was executed on Jaguar (Cray XT5) at the Oak Ridge Leadership Computing Facility using 40,000 cores in under 10 hours to model over 162 billion degrees of freedom for 10 loading steps.

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

    SciTech Connect

    Hardin, Ernest; Hadgu, Teklu; Clayton, Daniel James

    2015-03-01

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

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

    SciTech Connect

    Billone, M. C.; Burtseva, T. A.

    2016-08-30

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

  14. The spent fuel safety experiment

    SciTech Connect

    Harms, G.A.; Davis, F.J.; Ford, J.T.

    1995-08-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The SFSX provides a direct measurement of the reactivity effects of spent PWR fuel using a well-characterized, spent fuel sample. The SFSX also provides an experimental measurement of the end-effect, i.e., the reactivity effect of the variation of the burnup profile at the ends of PWR fuel rods. The design of the SFSX is optimized to yield accurate benchmark measurements of the effects of interest, well above experimental uncertainties.

  15. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect

    Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

    2013-07-01

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

  16. Crevice chemistry control in PWR steam generators

    SciTech Connect

    Sawochka, S.G.; Choi, S.S.; Millett, P.J.; Bates, J.; Gardner, J.

    1995-12-31

    To establish a basis for predicting and eventually controlling crevice solution chemistry in PWR steam generators, hideout tests were performed at several units. Results indicated that impurity hideout rates varied with the species and with bulk water concentration. Field evaluations of crevice impurity inventory models based on the hideout rate data indicated that further model refinements were necessary, e.g., more frequent quantification of the relation of hideout rates and bulk water concentration. An alternate crevice inventory model based on a real-time mass balance approach also began to be pursued. Modeling results currently are being used at several PWRs to establish a chloride injection rate consistent with development of a near neutral crevice solution to minimize IGA/SCC. Hideout return data are being used to independently establish predictions of crevice chemistry and to substantiate the hideout rate and mass balance model predictions.

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

    DTIC Science & Technology

    2013-06-01

    commercial nuclear power occurred on April 26 th , 1986 at the then Soviet Chernobyl Nuclear Power Plant in what is now the Ukraine. The reactor...correlations for the two cases will not be significantly different. 2.10 Specific Power Specific power is the ratio of a core’s rated thermal power to its...in one day per kgIHM—i.e. the daily burnup. Since the core thermal power and cycle length are the same for all cases considered herein—yet the core

  18. PWR internal flow modeling with fuel assemblies details

    SciTech Connect

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

    2012-07-01

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

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

    SciTech Connect

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

    1987-05-21

    For a severe PWR accident that leads to a loss of feedwater to the steam generators, such as might occur in a station blackout, fission product decay heating will cause a water boiloff. Without effective cooling of the core, steam will begin to oxidize the Zircaloy cladding. The noble gases and volatile fission products, such as Cs and I, that are major contributors to the radiological source term, will be released from the damaged fuel shortly after cladding failure. The accident environment when these volatile fission products escape was simulated in STEP-3 using four fuel elements from the Belgonucleaire BR3 reactor. The primary objective was to examine the releases in samples collected as close to the test zone as possible. In this paper, an analysis of the temperatures and hydrogen generation is compared with the measurements. The analysis is needed to estimate releases and characterize conditions at the source for studies of fission product transport.

  20. PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

    SciTech Connect

    Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C; Murphy, Brian D; Mueller, Don

    2007-09-01

    The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.

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

    NASA Astrophysics Data System (ADS)

    Hartini, Entin; Andiwijayakusuma, Dinan

    2014-09-01

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

  2. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  3. High Cycle Thermal Fatigue in French PWR

    SciTech Connect

    Blondet, Eric; Faidy, Claude

    2002-07-01

    Different fatigue-related incidents which occurred in the world on the auxiliary lines of the reactor coolant system (SIS, RHR, CVC) have led EDF to search solutions in order to avoid or to limit consequences of thermodynamic phenomenal (Farley-Tihange, free convection loop and stratification, independent thermal cycling). Studies are performed on mock-up and compared with instrumentation on nuclear power stations. At the present time, studies allow EDF to carry out pipe modifications and to prepare specifications and recommendations for next generation of nuclear power plants. In 1998, a new phenomenal appeared on RHR system in Civaux. A crack was discovered in an area where hot and cold fluids (temperature difference of 140 deg. C) were mixed. Metallurgic studies concluded that this crack was caused by high cycle thermal fatigue. Since 1998, EDF is making an inventory of all mixing areas in French PWR on basis of criteria. For all identified areas, a method was developed to improve the first classifying and to keep back only potential damage pipes. Presently, studies are performing on the charging line nozzle connected to the reactor pressure vessel. In order to evaluate the load history, a mock-up has been developed and mechanical calculations are realised on this nozzle. The paper will make an overview of EDF conclusions on these different points: - dead legs and vortex in a no flow connected line; - stratification; - mixing tees with high {delta}T. (authors)

  4. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    SciTech Connect

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  5. Design of Recycle PWR with Heavy Water Moderation

    SciTech Connect

    Hibi, K.; Uchita, M.

    2002-07-01

    This study shows the conceptual plant design of the recycle PWR (RPWR), which is an innovative MOX-PWR with breeding ratios around 1.1 moderated by heavy water. Most of the plant systems of RPWR can employ the systems of PWRs. RPWR has no acid boron systems and has a small tritium removal system. The construction and operation costs are similar to the current PWRs. While, heavy water cost will be decreased drastically with up-to-date producing methods. The reliability for the plant systems of RPWR is high and R and D cost for realizing RPWR is very low because the core design of RPWR is fundamentally based on the current PWR technology. (authors)

  6. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    SciTech Connect

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A.

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

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

    SciTech Connect

    Faidy, C.

    1997-04-01

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

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

    SciTech Connect

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

    2016-05-01

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

  9. Effect of coolant chemistry on PWR radiation transport processes. Progress report on reactor loop studies

    SciTech Connect

    Brown, D.J.; Flynn, G.; Haynes, J.W.; Kitt, G.P.; Large, N.R.; Lawson, D.; Mead, A.P.; Nichols, J.L.; Woodwark, D.R.

    1986-05-01

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. There are strong indications that the in-core deposition behavior of corrosion product species is not fully accounted for by the solubility model based on nickel ferrite; boric acid plays a role apart from its influence on pH, and corrosion products are adsorbed to some extent in the zirconium oxide film on the fuel cladding. In DWL, soluble species appear to be dominant in deposition processes. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. 13 figs.

  10. Whole-core comet solutions to a 3-dimensional PWR benchmark problem with gadolinium

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2012-07-01

    A pressurized water reactor (PWR) benchmark problem with gadolinium was used to determine the accuracy and computational efficiency of the coarse mesh radiation transport method COMET. The benchmark problem contains 193 square fuel assemblies. The COMET solution (eigenvalue, assembly averaged and fuel pin averaged fission density distributions) was compared with those obtained from the corresponding Monte Carlo reference solution using the same 2-group material cross section library. The comparison showed that both the core eigenvalue and fission density distribution averaged over each assembly and fuel pin predicated by COMET agree very well with the corresponding MCNP reference solution if the incident flux response expansion used in COMET is truncated at 2nd order in the two spatial and the two angular variables. The benchmark calculations indicate that COMET has Monte Carlo accuracy. In, particular, the eigenvalue difference between the codes ranged from 17 pcm to 35 pcm, being within 2 standard deviations of the calculational uncertainty. The mean flux weighted relative differences in the assembly and fuel pin fission densities were 0.47% and 0.65%, respectively. It was also found that COMET's full (whole) core computational speed is 30,000 times faster than MCNP in which only 1/8 of the core is modeled. It is estimated that COMET would have been about over 6 orders of magnitude faster than MCNP if the full core were also modeled in MCNP. (authors)

  11. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect

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

    1997-12-01

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

  12. TREAT source-term experiment STEP-1 simulating a PWR LOCA

    SciTech Connect

    Simms, R.; Baker, L. Jr.; Blomquist, C.A.; Ritzman, R.L.

    1986-01-01

    In a hypothetical pressurized water reactor (PWR) large-break loss-of-coolant accident (LOCA) in which the emergency core cooling system fails, fission product decay heating causes water boil-off and reduced heat removal. Zircaloy cladding is oxidized by the steam. The noble gases and volatile fission products such as cesium and iodine that constitute a principal part of the source term will be released from the damaged fuel at or shortly after the time of cladding failure. TREAT test STEP-1 simulated the LOCA environment when the volatile fission products would be released using four fuel elements from the Belgonucleaire BR3 reactor. The principal objective was to collect a portion of the releases carried by the flow stream in a region as close as possible to the test zone. In this paper, the test is described and the results of an analysis of the thermal and steam/hydrogen environment are compared with the test measurements in order to provide a characterization for analysis of fission product releases and aerosol formation. The results of extensive sample examinations are reported separately.

  13. Advanced Nuclear Fuel Development in Japan

    NASA Astrophysics Data System (ADS)

    Yamawaki, Michio

    2003-06-01

    The verification test programs of high burnup BWR and PWR fuels have been carried out by Nuclear Power Engineering Corporation under the sponsorship of Ministry of Economy, Trade and Industry since 1986. BWR and PWR fuel assemblies of high burnup range of up to about 48 GWd/t and 53 GWd/t, respectively were examined by hot cell PIEs and many segment rods of local burnup range of up to more than 60GWd/t were power ramped in test reactors. Though some fuel rods showed minor failure after power ramp tests beyond commercial reactor condition, the results have shown good performance of the high burnup fuels in general. In BWR power ramp tests, the new failure mode of segment rods and the decrease of the failure threshold for higher burnup fuels have been found. Other than oxide fuel, new type fuels such as metallic, nitride and hydride fuels are under research and development in Japan for fast breeder reactors and, in case of hydride fuel, for both fast reactors and LWRs. Topics on some of these new type fuels will be also presented.

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

    NASA Astrophysics Data System (ADS)

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

    2014-06-01

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

  15. Nuclear fuel performance: Trends, remedies and challenges

    NASA Astrophysics Data System (ADS)

    Rusch, C. A.

    2008-12-01

    It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

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

    SciTech Connect

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  17. Estimating probable flaw distributions in PWR steam generator tubes

    SciTech Connect

    Gorman, J.A.; Turner, A.P.L.

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  18. Fuel characteristics required for LWR fuel rod calculations

    NASA Astrophysics Data System (ADS)

    De Meulemeester, E.

    1982-04-01

    BELGONUCLEAIRE gradually increasing in-reactor experience has enabled to assess the relative importance of attributes defined in specifications and drawings for both UO 2 and MO 2 fuels. On the basis of that experience, design codes have been benchmarked and were thereafter applied to cover the range of parameters and irradiation histories to be encountered or evaluated. To illustrate the effects of fuel characteristics on fuel behaviour, sensitivity calculations were performed on the basis of actual fuel irradiated in BWR's (DODEWAARD, GARIGLANO and OYSTER CREEK) and PWR's (BR3, DOEL, SENA, TIHANGE and MAINE YANKEE). The major characteristics are : fuel structure, UO 2 versus mixed oxide fuel; fuel accomodation (depending on the fuel microstructure and chemical composition); fuel density and densification stability; open porosity; pellet end geometry; pellet L/D ratio, gap size. Although the influence of the various parameters is not additive, these examples enable to determine the relative influence of each characteristic and to conclude to what accuracy it should be measured (in demo fuel) or controlled (in production fuel).

  19. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

  20. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    SciTech Connect

    Ishijima, Kiyomi; Fuketa, Toyoshi

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

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

    SciTech Connect

    Cheverton, R.D.

    1982-01-01

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

  2. VERA Core Simulator Methodology for PWR Cycle Depletion

    SciTech Connect

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

    2015-01-01

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

  3. 77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-02

    ... criticality safety analyses of pressurized water reactor spent nuclear fuel (SNF) in transportation packages... Doc No: 2012-10618] NUCLEAR REGULATORY COMMISSION [NRC-2012-0100] Burnup Credit in the Criticality... the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.'' This...

  4. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    SciTech Connect

    Sonat Sen; Gilles Youinou

    2013-02-01

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this case the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)

  5. Neutron collar calibration and evaluation for assay of LWR fuel assemblies containing burnable neutron absorbers

    SciTech Connect

    Henriksen, P.W.; Menlove, H.O.; Stewart, J.E.; Qiao, S.Z.; Wenz, T.R. ); Verrecchia, G.P.D. . Safeguards Directorate)

    1990-11-01

    The neutron coincidence collar is used to verify the uranium content in light water reactor fuel assemblies. An AmLi neutron source actively interrogates the fuel assembly to measure the {sup 235}U content and the {sup 238}U content can be verified from a passive neutron coincidence measurement. This report gives the collar calibration data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies both with and without cadmium liners. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and various fuel assembly sizes. The data were collected using the Los Alamos BWR and PWR test assemblies as well as fuel assemblies from several fuel fabrication facilities. 11 refs., 15 figs., 14 tabs.

  6. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

    SciTech Connect

    Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa; Xu, Yiban; Cao, Liping

    2016-01-01

    The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

  7. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

    SciTech Connect

    A. K. MAJI; B. MARSHALL; ET AL

    2000-10-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  8. Comparison of PWR - Burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results

    SciTech Connect

    Oberle, P.; Broeders, C. H. M.; Dagan, R.

    2006-07-01

    The increasing tendency towards fuel lifetime extension in thermal nuclear reactors motivated validation work for available evaluation tools for nuclear fuel burnup calculations. In this study two deterministic codes with different transport solvers and one Monte Carlo method are investigated. The code system KAPROS/KARBUS uses the classical deterministic First Collision Probability method utilizing a cylinderized Wigner-Seitz cell. In the SCALES.0/TRITON/NEWT code the Extended Step Characteristic method is applied. In a first step the two deterministic codes are compared with experimental results from the KWO-Isotope Correlation Experiment up to 30 MWD/kg HM burnup, published in 1981. Two pin cell calculations are analyzed by comparison of calculated and experimental results for important heavy isotope vectors. The results are very satisfactory. Subsequently, further validation at higher burnup (< 80 MWD/kg HM) is provided by comparison of the two deterministic codes and the Monte Carlo based burnup code MONTEBURNS for PWR UO{sub 2} fuel assembly calculations. Possible reasons for differences in the results are analyzed and discussed. Especially the influence of cross section data and processing is presented. (authors)

  9. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    SciTech Connect

    Kavaklioglu, K.; Ikonomopoulos, A. )

    1993-01-01

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint.

  10. 103. PWR2 CORE SUPPORT FLANGE BEING SEATED ON REACTOR VESSEL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    103. PWR-2 CORE SUPPORT FLANGE BEING SEATED ON REACTOR VESSEL FLANGE, APRIL 14, 1964 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

  11. Subchannel analysis of multiple CHF events. [PWR; BWR

    SciTech Connect

    Reddy, D.G.; Fighetti, C.F.

    1982-08-01

    The phenomenon of multiple CHF events in rod bundle heat transfer tests, referring to the occurrence of CHF on more than one rod or at more than one location on one rod is examined. The adequacy of some of the subchannel CHF correlations presently used in the nuclear industry in predicting higher order CHF events is ascertained based on local coolant conditions obtained with the COBRA IIIC subchannel code. The rod bundle CHF data obtained at the Heat Transfer Research Facility of Columbia University are examined for multiple CHF events using a combination of statistical analyses and parametric studies. The above analyses are applied to the study of three data sets of tests simulating both PWR and BWR reactor cores with uniform and non-uniform axial heat flux distributions. The CHF correlations employed in this study include: (1) CE-1 correlation, (2) B and W-2 correlation, (3) W-3 correlation, and (4) Columbia correlation.

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

    SciTech Connect

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

    1997-04-01

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

  13. Pump and valve fastener serviceability in PWR nuclear facilities

    SciTech Connect

    Moisidis, N.T.; Ratiu, M.D.

    1996-02-01

    The results of several studies conducted on corrosion of carbon and low-alloy steels in borated water have shown that impingement of borated steam on ferritic steels or contact with a moist paste of boric acid can lead to high corrosion rates due to high local concentrations of boric acid on the surface. The corrosion process of the flange fasteners of pumps and valves is considered a material compatibility and equipment maintenance problem. Therefore, the nuclear utilities of pressurized water reactor (PWR) power plants can prevent this damage by implementing appropriate fastener steel replacement and extended inspections to detect and correct the cause of leakage. A 3-phase corrosion protection program is presented for implementation based on system operability, outage-related accessibility, and cost of fastener replacement versus maintenance frequency increase. A selection criterion for fastener material is indicated based on service limitation: preloading and metal temperature.

  14. Modeling local chemistry in PWR steam generator crevices

    SciTech Connect

    Millett, P.J.

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

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

    SciTech Connect

    Dukelow, J S; Harrison, D G; Morgenstern, M

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  16. Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

    DOE PAGES

    George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; ...

    2014-12-01

    Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resultingmore » operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.« less

  17. Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

    SciTech Connect

    George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; Godfrey, Andrew T.; Gehin, Jess C.; Powers, Jeffrey J.

    2014-12-01

    Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resulting operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.

  18. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    SciTech Connect

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

    1998-09-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  19. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    SciTech Connect

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  20. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    J.K. Knudson

    2003-10-02

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

  1. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    A. Alsaed

    2005-07-28

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

  2. Spent fuel disassembly hardware and other non-fuel bearing components: characterization, disposal cost estimates, and proposed repository acceptance requirements

    SciTech Connect

    Luksic, A.T.; McKee, R.W.; Daling, P.M.; Konzek, G.J.; Ludwick, J.D.; Purcell, W.L.

    1986-10-01

    There are two categories of waste considered in this report. The first is the spent fuel disassembly (SFD) hardware. This consists of the hardware remaining after the fuel pins have been removed from the fuel assembly. This includes end fittings, spacer grids, water rods (BWR) or guide tubes (PWR) as appropriate, and assorted springs, fasteners, etc. The second category is other non-fuel-bearing (NFB) components the DOE has agreed to accept for disposal, such as control rods, fuel channels, etc., under Appendix E of the standard utiltiy contract (10 CFR 961). It is estimated that there will be approximately 150 kg of SFD and NFB waste per average metric ton of uranium (MTU) of spent uranium. PWR fuel accounts for approximately two-thirds of the average spent-fuel mass but only 50 kg of the SFD and NFB waste, with most of that being spent fuel disassembly hardware. BWR fuel accounts for one-third of the average spent-fuel mass and the remaining 100 kg of the waste. The relatively large contribution of waste hardware in BWR fuel, will be non-fuel-bearing components, primarily consisting of the fuel channels. Chapters are devoted to a description of spent fuel disassembly hardware and non-fuel assembly components, characterization of activated components, disposal considerations (regulatory requirements, economic analysis, and projected annual waste quantities), and proposed acceptance requirements for spent fuel disassembly hardware and other non-fuel assembly components at a geologic repository. The economic analysis indicates that there is a large incentive for volume reduction.

  3. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    SciTech Connect

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

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

    NASA Astrophysics Data System (ADS)

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

    2014-04-01

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

  5. Severe accident modeling of a PWR core with different cladding materials

    SciTech Connect

    Johnson, S. C.; Henry, R. E.; Paik, C. Y.

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

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

    SciTech Connect

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

    2015-09-01

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

  7. Fuel flexible fuel injector

    DOEpatents

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  8. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    SciTech Connect

    Bardet, Philippe; Ricciardi, Guillaume

    2016-01-31

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  9. A Feasibility Study of an Integral PWR for Space Applications

    SciTech Connect

    Grandis, S. De; Finzi, E.; Lombardi, C.V.; Mandelli, D.; Padovani, E.; Passoni, M.; Ricotti, M.E.; Santini, L.

    2004-07-01

    Fission space power systems are well suited to provide safe, reliable, economic and robust energy sources, in the order of 100 KWe. A preliminary feasibility study of a nuclear fission reactor is here presented with the following requirements: i) high reliability, ii) R and D program of moderate cost, iii) to be deployed within a reasonable period of time (e.g. 2015), iv) to be operated and controlled for a long time (10 years) without human intervention, v) possibly to be also used as a byproduct for some particular terrestrial application (or at least to share common technologies), vi) to start with stationary application. The driving idea is to extend as much as possible the PWR technology, by recurring to an integral type reactor. Two options are evaluated for the electricity production: a Rankine steam cycle and a Rankine organic fluid cycle. The neutronics calculation is based on WIMS code benchmarked with MCNP code. The reactivity control is envisaged by changing the core geometry. The resulting system appears viable and of reasonable size, well fit to the present space vector capabilities. Finally, a set of R and D needs has been identified: cold well, small steam turbines, fluid leakage control, pumps, shielding, steam generator in low-gravity conditions, self pressurizer, control system. A R and D program of reasonable extent may yield the needed answers, and some demanding researches are of interest for the new generation Light Water Reactors. (authors)

  10. Effect of component aging on PWR control rod drive systems

    SciTech Connect

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-01-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock Wilcox (B W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging.

  11. Effect of component aging on PWR control rod drive systems

    SciTech Connect

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-06-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock & Wilcox (B & W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging.

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

    SciTech Connect

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

    2015-09-01

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

  13. Evaluation of surface modification techniques for PWR steam generator channel heads. Final report

    SciTech Connect

    Spalaris, C.N.

    1986-06-01

    Surface modification which were developed under a previous EPRI program and then applied to Boiling Water Reactor replacement piping, were modified for treating PWR steam generator channel head surfaces. Surface modifications have been shown to reduce out-of-core activity build up in BWR and thought to be equally effective in PWR circuits as well. Prototypical surface test specimens were used to develop techniques appropriate to PWR alloy substrates which were then applied to treat the surfaces of a spare, full size PWR channel head in a field demonstration. Modified surfaces cut from test specimens and pieces removed from the field demonstration were submitted to metallurgical investigations. No damage to the substrate alloys was detected as a result of the surface modification processes. Combination of mechanical and electropolishing action improved the as fabricated finish by at least a factor of 3 for the Inconel plate and factors of 20 for the stainless weld overlay. Field demonstration yielded a factor of 10 improvement in the weld overlay and 30 to 40% in the divider plate. Because these surfaces are known to be responsible for 57% of the area radioactivity in PWR steam generators in service, prepolishing is expected to reduce radiation fields substantially. 31 figs.

  14. Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

    SciTech Connect

    Wagner, J.C.

    2000-10-12

    Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and

  15. Pressurized water reactor fuel assembly subchannel void fraction measurement

    SciTech Connect

    Akiyama, Yoshiei; Hori, Keiichi; Miyazaki, Keiji; Mishima, Kaichiro; Sugiyama, Shigekazu

    1995-12-01

    The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

  16. BR-100 spent fuel shipping cask development

    SciTech Connect

    McGuinn, E.J.; Childress, P.C.

    1990-01-01

    Continued public acceptance of commercial nuclear power is contingent to a large degree on the US Department of Energy (DOE) establishing an integrated waste management system for spent nuclear fuel. As part of the from-reactor transportation segment of this system, the B W Fuel Company (BWFC) is under contract to the DOE to develop a spent-fuel cask that is compatible with both rail and barge modes of transportation. Innovative design approaches were the keys to achieving a cask design that maximizes payload capacity and cask performance. The result is the BR-100, a 100-ton rail/barge cask with a capacity of 21 PWR or 52 BWR ten-year cooled, intact fuel assemblies. 3 figs.

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

    SciTech Connect

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

    2014-01-01

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

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

    SciTech Connect

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

    1993-06-01

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

  19. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    SciTech Connect

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of /sup 235/U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the /sup 235/U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described.

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

    SciTech Connect

    Grenouillet, Jean-Jacques

    2007-07-01

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

  1. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    NASA Astrophysics Data System (ADS)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. This approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.

  2. Spent fuel characteristics & disposal considerations

    SciTech Connect

    Oversby, V.M.

    1996-06-01

    The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

  3. PWR FLECHT SEASET 21-rod bundle flow blockage task data and analysis report. NRC/EPRI/Westinghouse Report No. 11. Appendices K-P

    SciTech Connect

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  4. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect

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

    1999-10-01

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

  5. Steady-state fuel behavior modeling of nitride fuels in FRAPCON-EP

    NASA Astrophysics Data System (ADS)

    Feng, Bo; Karahan, Aydın; Kazimi, Mujid S.

    2012-08-01

    Fuel material properties and mechanistic fission gas models in FRAPCON-EP were updated to model the steady-state behavior of high-porosity nitride fuel operating at temperatures below half of the melting point. The fuel thermal conductivity and fuel thermal expansion models were updated with correlations for UN and (U,Pu)N fuels. Hot-pressing of the as-fabricated porosity was modeled as a function of the hydrostatic pressure and creep rate. The solid fission product swelling was assumed to increase linearly with burnup. Fission gas swelling constitutive models were updated to appropriately capture the intragranular gas bubble evolution in nitride fuel. Intergranular gas swelling was neglected due to the assumed high porosity of the fuel. The fission gas release behavior was modeled by fitting the fission gas diffusion coefficient in UN to FRAPCON's default fission gas release model. This fitted gas diffusion coefficient reflects the effects of porosity, burnup, operating temperature, fission rate, and bubble sink strength. Fission gas release and fuel swelling benchmarks against irradiation data were performed. The updated code was applied to UN fuel in typical PWR geometry and operating conditions, with an extended cycle length of 24 months. The results show that swelling of the nitride fuel up to 60 MWd/kg burnup did not lead to excessive straining of the cladding. Furthermore, this study showed that a porous (>15% porosity) nitride fuel pellet could achieve a much higher margin to failure from the cladding collapse and grid-to-rod fretting.

  6. Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs

    SciTech Connect

    Greenspan, E

    2006-04-30

    The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity – in particular for BWR’s, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR’s and BWR’s without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR’s and BWR’s were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density – on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR’s more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel – reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ~2/3 that of the MOX fuel and the

  7. Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

    SciTech Connect

    Wilson, C.L.; Hesson, G.M.; Pilger, J.P.; King, L.L.; Panisko, F.E.

    1993-09-01

    This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuel bundle is cooled.

  8. Fuel consolidation demonstration: Consolidation concept development

    SciTech Connect

    Not Available

    1990-02-01

    EPRI, Northeast utilities Service Company (NUSCO), DOE, Baltimore Gas Electric Company, and Combustion Engineering, Inc. (C-E) are engaged in a program to develop a system for consolidating spent fuel, in which the consolidated fuel will be licensable by NRC for storage in the spent-fuel storage pool. Fuel consolidation offers a means of substantially increasing the capacity of spent-fuel storage pools. Consolidation equipment design, development, construction, and testing are being performed by C-E in Windsor, Connecticut. Seismic and structural evaluation of the capability of the Millstone Unit 2 spent-fuel pool and building to accommodate the increased fuel capacity is being conducted by NUSCO. NUSCO plans to obtain a license to store consolidated fuel in the Millstone-2 spent-fuel storage pool. NUSCO also plans to perform a hot demonstration of the integrated consolidation system with spent fuel at Millstone-2. This report describes the consolidation system design that forms the basis for the detailed design of the equipment comprising the system, including information on the fabrication and testing of the equipment. Appendix B describes an evaluation of the ability of the system under development to consolidate LWR spent-fuel assemblies other than the 14 {times} 14 fuel of C-E design stored at Millstone-2. A comparison was made of fuel-assembly designs on the basis of information available in open literature. It was concluded that with appropriate dimensional modifications the spent-fuel consolidation system equipment design is applicable to almost all PWR fuel-assembly configurations. 8 refs., 20 figs.

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

    SciTech Connect

    Chung, H. M.; Energy Technology

    2006-01-31

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

  10. Crack growth rates of nickel alloy welds in a PWR environment.

    SciTech Connect

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

    2006-05-31

    In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of {approx}5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

  11. Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics

    NASA Astrophysics Data System (ADS)

    Nuttin, A.; Capellan, N.; David, S.; Doligez, X.; El Mhari, C.; Méplan, O.

    2014-06-01

    Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.

  12. Spent fuel pin temperature PC code

    SciTech Connect

    Fischer, L.E.

    1985-03-01

    During an annual outage, a Pressurized Water Reactor (PWR) may discharge 60 or more spent fuel bundles into its storage pool. Most early PWRs were built to store 3 to 5 years of spent fuel in their pools and are beginning to exceed their capacities. One method currently being developed and licensed for expanding spent fuel storage capabilities is the dry storage of spent fuel in large casks. To reduce the probability of gross failures of fuel cladding during dry storage in casks, the fuel pin temperatures must be shown to remain within acceptable limits. LLNL has developed, for the Nuclear Regulatory Commission, a personal computer (PC) code for calculating fuel pin temperatures on the IBM PC. The code uses the Wooton-Epstein Correlation to calculate the pin temperatures and has been benchmarked against test data. An iterative type of solution is used to calculate the fuel pin temperatures for specified heat fluxes and pin configurations. The PC code is useful in performing confirmatory analyses and comparing the results with those submitted by applicants applying for storage licenses. 5 references, 2 tables.

  13. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect

    Chodak, III, Paul

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  14. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  15. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    SciTech Connect

    Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir; Rael, Carlos D.; Desimone, David J.

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

  16. Preliminary study on direct recycling of spent BWR fuel in BWR system

    NASA Astrophysics Data System (ADS)

    Waris, A.; Sumbono, Prayudhatama, Dythia; Novitrian, Su'ud, Zaki

    2012-06-01

    Spent fuel management is considered to be one of the main problems in energy nuclear utilization. Recycling after reprocessing is one of the options for dealing with nuclear reactor spent fuel. Reprocessing is very costly and needs remote handling since spent fuel is very hazard high level waste. On top of that, only a small number of countries can manage a reprocessing plant. If country likes Indonesia decide to "go nuclear", it should find another way to deal with the nuclear spent fuel. Korea has proposed the DUPIC (Direct Utilization of Spent PWR fuel In CANDU) concept. Nevertheless, DUPIC concept requires two types of nuclear power plants, i.e., pressurized water reactor (PWR) and CANadian Deuterium Uranium reactor (CANDU). In this study, we evaluate a scheme of direct recycling of spent BWR fuel in BWR system, under the concept that we have called as a SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario. Several spent BWR fuel compositions in loaded BWR fuel has been evaluated to achieve the criticality of reactor.

  17. Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

    SciTech Connect

    Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

    2016-01-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  18. Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

    SciTech Connect

    Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

    2015-12-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  19. Spent fuel dry storage technology development: thermal evaluation of sealed storage cask containing spent fuel

    SciTech Connect

    Schmitten, P.F.; Wright, J.B.

    1980-08-01

    A PWR spent fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in a instrumented above surface storage cell during December 1978 for thermal testing. Instrumentation provided to measure canister, liner and concrete temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and liner and in three radial positions in the concrete. Temperatures from the SSC test assembly have been recorded throughout the past 16 months. Canister and liner temperatures have reached their peak values of 200{sup 0}F and 140{sup 0}F, respectively. Computer predictions of the transient and steady-state temperatures show good agreement with the test data.

  20. SPENT NUCLEAR FUEL WASTE PACKAGE FILLER TESTING TECHNICAL GUIDELINES DOCUMENT

    SciTech Connect

    J.A. Cogar

    1996-08-28

    The purpose of this work is to provide supporting calculations for determination of the radiation source terms specific to subsurface shielding design and analysis. These calculations are not intended to provide the absolute values of the source terms, which are under the charter of the Waste Package Operations (WPO) Group. Rather, the calculations focus on evaluation of the various combinations of fuel enrichment, burnup and cooling time for a given decay heat output, consistent with the waste package (WP) thermal design basis. The objective is to determine the worst-case combination of the fuel characteristics (enrichment, burnup and cooling time) which would give the maximum radiation fields for subsurface shielding considerations. The calculations are limited to PWR fuel only, since the WP design is currently evolving with thinner walls and a reduced heat load as compared to the viability assessment (VA) reference design. The results for PWR fuel will provide a comparable indication of the trend for BWR fuel, as their characteristics are similar. The source term development for defense high-level waste and other spent nuclear fuel (SNF) is the responsibility of the WPO Group, and therefore, is not included this work. This work includes the following items responsive to the stated purpose and objective: (1) Determine the possible fuel parameters (initial enrichment, burnup and cooling time), that give the same decay heat value as specified for the waste package thermal design; (2) Obtain the neutron and gamma source terms for the various combinations of the fuel parameters for use in radiation field calculations; and (3) Calculate radiation fields on the surfaces of the waste package and its transporter to quantify the effects of the fuel parameters with the same decay heat value for use in identifying the worst-case combination of the fuel parameters.

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

    NASA Astrophysics Data System (ADS)

    Yilmaz, Mine Ozdemir

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

  2. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3. International Agreeement Report

    SciTech Connect

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

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser.

  3. MC21 analysis of the MIT PWR benchmark: Hot zero power results

    SciTech Connect

    Kelly Iii, D. J.; Aviles, B. N.; Herman, B. R.

    2013-07-01

    MC21 Monte Carlo results have been compared with hot zero power measurements from an operating pressurized water reactor (PWR), as specified in a new full core PWR performance benchmark from the MIT Computational Reactor Physics Group. Included in the comparisons are axially integrated full core detector measurements, axial detector profiles, control rod bank worths, and temperature coefficients. Power depressions from grid spacers are seen clearly in the MC21 results. Application of Coarse Mesh Finite Difference (CMFD) acceleration within MC21 has been accomplished, resulting in a significant reduction of inactive batches necessary to converge the fission source. CMFD acceleration has also been shown to work seamlessly with the Uniform Fission Site (UFS) variance reduction method. (authors)

  4. The electrochemistry in 316SS crevices exposed to PWR-relevant conditions

    NASA Astrophysics Data System (ADS)

    Vankeerberghen, M.; Weyns, G.; Gavrilov, S.; Henshaw, J.; Deconinck, J.

    2009-04-01

    The chemical and electrochemical conditions within a crevice of Type 316 stainless steel in boric acid-lithium hydroxide solutions under PWR-relevant conditions were modelled with a computational electrochemistry code. The influence of various variables: dissolved hydrogen, boric acid, lithium hydroxide concentration, crevice length, and radiation dose rate was studied. It was found with the model that 25 ccH 2/kg (STP) was sufficient to remain below an electrode potential of -230 mV she, commonly accepted sufficient to prevent stress corrosion cracking under BWR conditions. In a PWR plant various operational B-Li cycles are possible but it was found that the choice of the cycle did not significantly influence the model results. It was also found that a hydrogen level of 50 ccH 2/kg (STP) would be needed to avoid substantial lowering of the pH inside a crevice.

  5. Generation and behavior of metal oxide colloids in PWR steam systems

    SciTech Connect

    Varsanik, R.G.

    1984-10-01

    This work reviews the curently available literature and research work on the generation and behavior of metal oxide colloids in PWR steam systems. The work of E. Matijevic et al on the generation and adhesion of iron and copper oxides is described. The role of colloid chemistry in the control of plant sludge and corrosion products is described. Factors affecting the adherence and re-entrainment of colloidal metal oxides along with possible methods for the control of metal oxide deposition are reviewed.

  6. Spent Nuclear Fuel Structural Response when Subject to an End Impact Accident

    SciTech Connect

    Adkins, Harold E.; Koeppel, Brian J.; Tang, David T.

    2004-11-19

    The US Nuclear Regulatory Commission (USNRC) is tasked with licensing of safe spent fuel storage and transportation systems. A subset of this responsibility is to investigate and understand the structural performance of these systems. In a joint effort between staff at the Pacific Northwest National Laboratory (PNNL) and the USNRC, computational studies were performed to predict the structural response of spent nuclear fuel when subject to an end impact accident. In this study, the structural performance of a typical Pressurized Water Reactor (PWR) fuel assembly is evaluated utilizing the ANSYS®/LS-DYNA® finite element analysis (FEA) code.

  7. Determination of soluble chromium in simulated PWR coolant by differential-pulse adsorptive stripping voltammetry.

    PubMed

    Torrance, K; Gatford, C

    1987-11-01

    An analytical method has been developed for the determination of dissolved chromium at concentrations less than 2 mug/l. in PWR coolant by differential-pulse adsorptive stripping voltammetry at a hanging mercury drop electrode. Concentrations above 2 mug/l. can be determined by appropriate dilution of the sample. The method is based on measurement of the current associated with reduction of a chromium(III)-DTPA (diethylenetriaminepenta-acetic acid) complex adsorbed at the surface of the mercury drop. The effects of boric acid, pH, DTPA concentration, accumulation potential and time were investigated together with the oxidation state of the chromium. No interference was observed from other transition metal ions expected to be present in PWR coolant. No alternative chemical technique of similar sensitivity was available for comparison with the results obtained in solutions containing <1 mug/l. chromium. Recoveries from simulated coolant solutions were greater than 95% and the relative standard deviations for single determinations were in the range 12-25%. The statistical limit of detection at the 95% confidence level was 0.023 mug/l. This method of analysis should prove valuable in corrosion studies and is uniquely capable of following the changes in soluble chromium concentration in PWR coolant that follow operational changes in the reactor.

  8. Differential pulse stripping voltammetry for the determination of nickel and cobalt in simulated PWR coolant.

    PubMed

    Torrance, K; Gatford, C

    1985-04-01

    The determination of ionic nickel and cobalt in simulated PWR coolant at concentrations below 1 microg/l. by differential pulse stripping voltammetry at a hanging mercury-drop electrode has been investigated. The high sensitivity for these ions results from the adsorptive accumulation of their dimethylglyoximate complexes on the mercury drop. Boric acid does not interfere and if the samples are adjusted to pH 9 with an ammonia-ammonium chloride buffer, both nickel and cobalt can be determined in the same run. The relative standard deviations at concentrations below 2 microg/l. are of the order of 5-7% and the limits of detection for nickel and cobalt are about 8 and 2 ng/l. respectively. These performance statistics show that this method is the most sensitive method currently available for determination of soluble nickel and cobalt in PWR coolant and it should prove to be most valuable in any corrosion studies of the materials of construction of the primary circuit of a PWR.

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

    SciTech Connect

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

    2004-07-01

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

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

    SciTech Connect

    Wagner, J.C.

    2001-08-02

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

  11. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    SciTech Connect

    Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R.; Rossa, R.; Liljenfeldt, H.

    2015-07-01

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)

  12. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    DOEpatents

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  13. Fossil fuels -- future fuels

    SciTech Connect

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

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

    SciTech Connect

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

    2009-07-12

    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

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

    SciTech Connect

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

    2006-10-20

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

  16. RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

    SciTech Connect

    David Andrs; Ray Berry; Derek Gaston; Richard Martineau; John Peterson; Hongbin Zhang; Haihua Zhao; Ling Zou

    2012-05-01

    The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to

  17. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    SciTech Connect

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  18. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    DOE PAGES

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-01-17

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

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

    SciTech Connect

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

    1989-01-01

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

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

    SciTech Connect

    Ellis, Ronald James

    2015-04-01

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

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

    SciTech Connect

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

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  2. URSULA2 computer program. Volume 2. Applications (sensitivity studies and demonstration calculations). Final report. [PWR

    SciTech Connect

    Keeton, L.W.; Marchland, E.O.; Singhal, A.K.; Spalding, D.B.

    1980-01-01

    The URSULA2 computer program has been developed for the thermal-hydraulic analysis of steam generators for PWR nuclear power plants. It computes three-dimensional distributions of velocity, pressure, enthalpy, etc., in the shell of the generator, and the distributions of primary-fluid temperature within the tubes. The code is applicable to both steady and unsteady flows and is equiped with three physical models: the equal velocity homogeneous model, a slip (or two-fluid) model, and an algebraic slip model. Applications, sensitivity studies, and demonstration calculations are presented.

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

    SciTech Connect

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

    1980-01-01

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

  4. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    SciTech Connect

    LEWIS, M.E.

    2000-04-06

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

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

    SciTech Connect

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

    1997-09-01

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

  6. Prospects for improvement of fuel performance and fuel cycle cost in BWRs and PWRs in the nineties

    SciTech Connect

    Ritter, G.L.; Woods, K.N.; Patten, T.W.; Patterson, J.F.

    1988-01-01

    Advanced Nuclear Fuels Corporation (ANF) has developed fuel assembly designs for BWRs and PWRs that can reduce fuel cycle costs by increasing uranium utilization in the reactors. Design features which have been used to increase uranium utilization include natural uranium axial blankets, 9x9 lattices with multiple water rods for BWRs and optimized water to fuel ratio lattices for PWRs. Gadolinia burnable absorber has been used in both BWR and PWR fuel designs to control power peaking which facilitates reactor operation for high burnup and extended cycles. New developments in fuel design and performance will further increase uranium utilization in the nineties. A high performance spacer grid design is being developed by ANF for use in PWRs and BWRs to increase coolant mixing efficiency with a resulting increase in critical heat flux and increased thermal margin which can be used to reduce fuel cycle costs. The development of more corrosion resistant cladding and other fuel design changes should enable fuel to attain assembly burnups of up to 45,000 MWd/MTU for BWRs and 60,000 MWd/MTU for PWRs with two year cycle lengths.

  7. PWR FLECHT SEASET 21-rod-bundle flow-blockage task: data and analysis report. NRC/EPRI/Westinghouse report No. 11, main report and appendices A-J

    SciTech Connect

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  8. Alternative Fuels

    EPA Pesticide Factsheets

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  9. On the Application of CFD Modeling for the Prediction of the Degree of Mixing in a PWR During a Boron Dilution Transient

    SciTech Connect

    Lycklama, Jan-Aiso; Hoehne, Thomas

    2006-07-01

    In a Pressurized Water Reactor, negative reactivity is present in the core by means of Boric acid as a soluble neutron absorber in the coolant water. During a so-called Boron Dilution Transient (BDT), a de-borated slug of coolant water is transported from the cold leg into the reactor vessel, and the borated coolant water is diluted by mixing with this un-borated water. The resulting decrease in the boron concentration leads to an insertion of positive reactivity in the core, which may lead to a reactivity excursion. The associated power peak may damage the fuel rods. The mixing of borated and un-borated water in downcomer and lower plenum is an important process, because it mitigates the degree of reactivity insertion. In the present study the application of Computational Fluid Dynamics (CFD) for the prediction of this mixing of un-borated with borated water in the RPV has been assessed. The analyses have been compared with the measurement data from the Rossendorf coolant mixing model (ROCOM) experiment. The ROCOM test facility represents the primary cooling system of a KONVOI type of PWR (1300 MW{sub el}). In spite of the complicated spatial, temporal, and geometrical aspects of the flow in the RPV, the agreement between the calculated and the experimental data is good. The CFD model tends to slightly under predict the degree of mixing in the RPV resulting in a slight under-prediction of the boron concentration at the core. (authors)

  10. A study on the expulsion of iodine from spent-fuel solutions

    SciTech Connect

    Sakurai, Tsutomu; Takahashi, Akira; Ishikawa, Niroh

    1995-02-01

    During dissolution of spent nuclear fuels, some radioiodine remains in spent-fuel solutions. Its expulsion to dissolver off-gas is important to minimize iodine escape to the environment. In our current work, the iodine remaining in spent-fuel solutions varied from 0 to 10% after dissolution of spent PWR-fuel specimens (approximately 3 g each). The amount remaining probably was dependent upon the dissolution time required. The cause is ascribable to the increased nitrous acid concentration that results from NOx generated during dissolution. The presence of nitrous acid was confirmed spectrophotometrically in an NO-HNO{sub 3} system at 100{degrees}C. Experiments examining NOx concentration versus the quantity of iodine in a simulated spent-fuel solution indicate that iodine (I{minus}) in spent fuels is subjected to the following three reactions: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid arising from NOx, and (3) formation of colloidal iodine (AgI, PdI{sub 2}), the major iodine species in a spent-fuel solution. Reaction (2) competes with reaction (3) to control the quantity of iodine remaining in solution. The following two-step expulsion process to remove iodine from a spent-fuel solution was derived from these experiments: Step One - Heat spent-fuel solutions without NOx sparging. When aged colloidal iodine is present, an excess amount of iodate should be added to the solution. Step Two - Sparge the fuel solution with NOx while heating. Effect of this new method was confirmed by use of a spent PWR-fuel solution.

  11. Application of the MELCOR code to design basis PWR large dry containment analysis.

    SciTech Connect

    Phillips, Jesse; Notafrancesco, Allen; Tills, Jack Lee

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  12. Regeneratively Cooled Liquid Oxygen/Methane Technology Development Between NASA MSFC and PWR

    NASA Technical Reports Server (NTRS)

    Robinson, Joel W.; Greene, Christopher B.; Stout, Jeffrey B.

    2012-01-01

    The National Aeronautics & Space Administration (NASA) has identified Liquid Oxygen (LOX)/Liquid Methane (LCH4) as a potential propellant combination for future space vehicles based upon exploration studies. The technology is estimated to have higher performance and lower overall systems mass compared to existing hypergolic propulsion systems. NASA-Marshall Space Flight Center (MSFC) in concert with industry partner Pratt & Whitney Rocketdyne (PWR) utilized a Space Act Agreement to test an oxygen/methane engine system in the Summer of 2010. PWR provided a 5,500 lbf (24,465 N) LOX/LCH4 regenerative cycle engine to demonstrate advanced thrust chamber assembly hardware and to evaluate the performance characteristics of the system. The chamber designs offered alternatives to traditional regenerative engine designs with improvements in cost and/or performance. MSFC provided the test stand, consumables and test personnel. The hot fire testing explored the effective cooling of one of the thrust chamber designs along with determining the combustion efficiency with variations of pressure and mixture ratio. The paper will summarize the status of these efforts.

  13. Phenomenon analysis of stress corrosion cracking in the vessel head penetrations of French PWR`s

    SciTech Connect

    Pichon, C.; Buisine, D.; Faidy, C.; Gelpi, A.; Vaindirlis, M.

    1995-12-31

    During a hydrotest in 1991, a leak was detected on,a reactor vessel head (RVH) penetration of a French PWR. This leak was due to a phenomenon of Primary Water Stress Corrosion Cracking (PWSCC) affecting these penetrations in Alloy 600. The destructive and non-destructive examinations undertaken during the following months highlighted the generic nature of the degradations. In order to well understand this phenomenon and implement the most suitable maintenance policy, a large scale scientific program was decided and performed jointly by Electricite de France and FRAMATOME. The paper will present all the results obtained in this program concerning the parameters governing the PWSCC. In particular the following fields will be developed: (1) the material, its microstructure in line with the manufacturing and its susceptibility to PWSCC; (2) the stresses and their evaluations by measurements, mock up corrosion tests and Finite Element Analysis (FEA); (3) the effect of surface finish on crack initiation; and (4) the crack growth rate. This phenomenon analysis will be useful for evaluating the risk of PWSCC on other Alloy 600 areas in PWR`s primary system.

  14. The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes

    SciTech Connect

    Pignatel, Jean-Francois

    2002-07-01

    Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, and leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor. (author)

  15. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-01-01

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  16. Fuel pin

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  17. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-10-03

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  18. Analysis of spent fuel assay with a lead slowing down spectrometer

    SciTech Connect

    Gavron, Victor I; Smith, L Eric; Ressler, Jennifer J

    2008-01-01

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it is possible to design a system that will provide around 1% statistical precision in the determination of the {sup 239}Pu, {sup 241}Pu and {sup 235}U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of {sup 238}U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle.

  19. Analysis of spent fuel assay with a lead slowing down spectrometer

    SciTech Connect

    Gavron, Victor I; Smith, L. Eric; Ressler, Jennifer J

    2010-10-29

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it is possible to design a system that will provide around 1% statistical precision in the determination of the {sup 239}Pu, {sup 241}Pu and {sup 235}U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of {sup 238}U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle.

  20. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    NASA Astrophysics Data System (ADS)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  1. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    SciTech Connect

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1985-04-01

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

  2. Fuel pump

    SciTech Connect

    Bellis, P.D.; Nesselrode, F.

    1991-04-16

    This patent describes a fuel pump. It includes: a fuel reservoir member, the fuel reservoir member being formed with fuel chambers, the chambers comprising an inlet chamber and an outlet chamber, means to supply fuel to the inlet chamber, means to deliver fuel from the outlet chamber to a point of use, the fuel reservoir member chambers also including a bypass chamber, means interconnecting the bypass chamber with the outlet chamber; the fuel pump also comprising pump means interconnecting the inlet chamber and the outlet chamber and adapted to suck fuel from the fuel supply means into the inlet chamber, through the pump means, out the outlet chamber, and to the fuel delivery means; the bypass chamber and the pump means providing two substantially separate paths of fuel flow in the fuel reservoir member, bypass plunger means normally closing off the flow of fuel through the bypass chamber one of the substantially separate paths including the fuel supply means and the fuel delivery means when the bypass plunger means is closed, the second of the substantially separate paths including the bypass chamber when the bypass plunger means is open, and all of the chambers and the interconnecting means therebetween being configured so as to create turbulence in the flow of any fuel supplied to the outlet chamber by the pump means and bypassed through the bypass chamber and the interconnecting means.

  3. Inspection procedures for experimental fuel production

    NASA Astrophysics Data System (ADS)

    Campsie, I. C.; Rattray, H. D.

    1988-04-01

    This paper describes the inspection procedures used in the development and manufacture of experimental fuel elements and their components. The examples quoted mainly apply to the PFR experimental fuel programme, although for well over a quarter of a century the procedures and techniques have been progressively developed and applied to the Magnox, SGHW, AGR, HTR, PFR and PWR fuel development programmes undertaken at the UKAEA's Springfields and Windscale Nuclear Power Development Laboratories. In contrast to production runs involving large numbers of standard components, experimental fuel is often assembled from components which, while they may look alike, may have design and material variations. Thus in addition to normal batching and bonding operations, great emphasis has to be placed on dimensional inspection, material testing and the individual identification of all items, thus maintaining traceability throughout all operations. The quality and performance of experimental items are often evaluated comparing pre- and post-test dimensional or NDT measurements. In the case of irradiation tests, several years can elapse between the measurements, therefore it is essential to ensure the reproducibility and compatibility of pre- and post-test measuring techniques and the traceability of all measured data and standards.

  4. Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

    SciTech Connect

    Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

    2012-06-06

    Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

  5. Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

    NASA Astrophysics Data System (ADS)

    Tucker, Lucas Powelson

    This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

  6. FUEL ASSEMBLY SHAKER TEST SIMULATION

    SciTech Connect

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  7. Corrosion report for the U-Mo fuel concept

    SciTech Connect

    Henager, Jr., Charles H.; Bennett, Wendy D.; Doherty, Ann L.; Fuller, E. S.; Hardy, John S.; Omberg, Ronald P.

    2014-08-28

    The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

  8. Quantification of Back-End Nuclear Fuel Cycle Metrics Uncertainties Due to Cross Sections

    SciTech Connect

    Tracy E. Stover, Jr.

    2007-11-01

    This work examines uncertainties in the back end fuel cycle metrics of isotopic composition, decay heat, radioactivity, and radiotoxicity. Most advanced fuel cycle scenarios, including the ones represented in this work, are limited by one or more of these metrics, so that quantification of them becomes of great importance in order to optimize or select one of these scenarios. Uncertainty quantification, in this work, is performed by propagating cross-section covariance data, and later number density covariance data, through a reactor physics and depletion code sequence. Propagation of uncertainty is performed primarily via the Efficient Subspace Method (ESM). ESM decomposes the covariance data into singular pairs and perturbs input data along independent directions of the uncertainty and only for the most significant values of that uncertainty. Results of these perturbations being collected, ESM directly calculates the covariance of the observed output posteriori. By exploiting the rank deficient nature of the uncertainty data, ESM works more efficiently than traditional stochastic sampling, but is shown to produce equivalent results. ESM is beneficial for very detailed models with large amounts of input data that make stochastic sampling impractical. In this study various fuel cycle scenarios are examined. Simplified, representative models of pressurized water reactor (PWR) and boiling water reactor (BWR) fuels composed of both uranium oxide and mixed oxides are examined. These simple models are intended to give a representation of the uncertainty that can be associated with open uranium oxide fuel cycles and closed mixed oxide fuel cycles. The simplified models also serve as a demonstration to show that ESM and stochastic sampling produce equivalent results, because these models require minimum computer resources and have amounts of input data small enough such that either method can be quickly implemented and a numerical experiment performed. The simplified

  9. Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

    SciTech Connect

    Simion, G.P.; VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B.; Bulmahn, K.D.

    1993-06-01

    This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement.

  10. Effect of single aging on stress corrosion cracking susceptibility of INCONEL X-750 under PWR conditions

    NASA Astrophysics Data System (ADS)

    Mishra, B.; Moore, J. J.

    1988-05-01

    Unfavorable morphology of precipitates and inclusions has been thought to be the cause of severe intergranular stress corrosion cracking (IGSCC) in double aged INCONEL* X-750 alloy used in reactor water environments. A single step aging treatment of 200 hours at 811 °C followed by furnace cooling after solution treating for 2 hours at 1075 °C has been found to provide an improved combination of strength, ductility, and resistance to SCC under simulated PWR test conditions. In this single aged condition a reprecipitated secondary carbide, together with γ' was produced at the grain boundary which resulted in a mixed fracture mode comprising dimple rupture and microvoid coalescence compared with a predominantly intergranular mode for the fully age hardened specimens. This improvement has been explained in terms of the morphology of the second phase precipitates which are produced in these heat treatment regimes.

  11. Calculation of the neutron source distribution in the VENUS PWR Mockup Experiment

    SciTech Connect

    Williams, M.L.; Morakinyo, P.; Kam, F.B.K.; Leenders, L.; Minsart, G.; Fabry, A.

    1984-01-01

    The VENUS PWR Mockup Experiment is an important component of the Nuclear Regulatory Commission's program goal of benchmarking reactor pressure vessel (RPV) fluence calculations in order to determine the accuracy to which RPV fluence can be computed. Of particular concern in this experiment is the accuracy of the source calculation near the core-baffle interface, which is the important region for contributing to RPV fluence. Results indicate that the calculated neutron source distribution within the VENUS core agrees with the experimental measured values with an average error of less than 3%, except at the baffle corner, where the error is about 6%. Better agreement with the measured fission distribution was obtained with a detailed space-dependent cross-section weighting procedure for thermal cross sections near the core-baffle interface region. The maximum error introduced into the predicted RPV fluence due to source errors should be on the order of 5%.

  12. A predictive model for corrosion fatigue crack growth rates in RPV steels exposed to PWR environments

    SciTech Connect

    Atkinson, J.D.; Chen, Z.; Yu, J.

    1995-12-31

    Corrosion fatigue crack propagation rates have been measured in A533B Class 1 plate in stagnant PWR primary water for a range of steel sulphur contents, temperature and corrosion potential values. Parametric descriptions of the data collected under constant rig conditions give good correlations for each variable and are consistent with a crack tip environment controlled process related to sulphur chemistry. A modified crack velocity equation is proposed to include temperature, sulphur content, polarization potential, frequency and {Delta}K values and it is shown how the predictions compare with the proposed ASME XI revision. Critical fatigue situations are identified for 0.003% and 0.019% sulphur steels typical of modern and old plant. The use of the equation in assessing the synergistic effect of variables is discussed.

  13. Development of cement solidification process for sodium borate waste generated from PWR plants

    SciTech Connect

    Hirofumi Okabe; Tatsuaki Sato; Yuichi Shoji; Yoshiko Haruguchi; Masaaki Kaneko; Michitaka Saso; Masumitsu Toyohara

    2013-07-01

    A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. To obtain high volume reduction and high mechanical strength of the waste, simulated concentrated borate liquid waste with a sodium / boron (Na/B) mole ratio of 0.27 was dehydrated and powdered by using a wiped film evaporator. To investigate the effect of the Na/B mole ratio on the solidification process, a sodium tetraborate decahydrate reagent with a Na/B mole ratio of 0.5 was also used. Ordinary portland cement (OPC) and some additives were used for the solidification. Solidified cement prepared from powdered waste with a Na/B mole ratio 0.24 and having a high silica sand content (silica sand/cement>2) showed to improved uniaxial compressive strength. (authors)

  14. Investigation of radial power and temperature effects in large-scale reflood experiments. [PWR

    SciTech Connect

    Motley, F.

    1983-01-01

    The largest reflood test facility in the world has been designed and constructed by the Japan Atomic Energy Research Institute (JAERI). The experimental test facility, known as the Cylindrical Core Test Facility (CCTF), models a full-height core section and the four primary loops of a Pressurized Water Reactor (PWR). The radial power distribution and temperature distribution were varied during the testing program. The test results indicate that the radial effects, while noticeable, do not appreciably alter the overall quenching behavior of the facility. The Transient Reactor Analysis Code (TRAC) correctly predicted the experimental results of several of the tests. The code results indicate that the core flow pattern adjusts multidimensionally to mitigate the effects of increased power or stored energy.

  15. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    SciTech Connect

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  16. Fog inerting effects on hydrogen combustion in a PWR ice condenser contaminant

    SciTech Connect

    Luangdilok, W.; Bennett, R.B.

    1995-05-01

    A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the upward lean flammability limits of the H{sub 2}-air-steam mixture in the ice condenser upper plenum region of a pressurized water reactor (PWR) ice condenser contaminant during postulated large loss of coolant accident (LOCA) conditions indicate that combustion may be suppressed beyond the downward flammability limit (8 percent H{sub 2} by volume). 18 refs., 3 tabs.

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

    SciTech Connect

    Pasichnyk, I.; Perin, Y.; Velkov, K.

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  18. Methods and findings of a systems interaction study of a Westinghouse PWR

    SciTech Connect

    Youngblood, R.; Hanan, N.; Fitzpatrick, R.; Xue, D.; Bozoki, G.; Fresco, A.; Papazoglou, I.; Mitra, S.; Macdonald, G.; Chelliah, E.

    1985-01-01

    This paper describes the methods and findings of a systems interaction study of a Westinghouse PWR. BNL conducted the study as a methods application that was performed to support the resolution of Unresolved Safety Issue A-17 on Systems Interactions. The method calls for a fault tree model of the plant to be developed in stages, corresponding to successively increasing levels of scope and detail. A functional model is developed first, resolved only to sufficient detail to reflect support system dependences; this guides the subsequent searches for spatial and induced-human interactions. This process has led to the identification of an active single failure causing loss of low pressure injection following a large or medium LOCA.

  19. The key to superior water chemistry at a PWR nuclear station

    SciTech Connect

    Dolan, R.; Miller, L.K.; Olejar, L.L.; Salem, E.

    1983-01-01

    This paper demonstrates how a condensate polishing unit can be successfully used to treat the feedwater for circulating-type pressurized water reactors (PWRs). Water chemistry at the Salem Generating Station, a two-unit, four-loop Westinghouse PWR located in New Jersey, is discussed. Topics considered include a plant description and the history of early operation, the role of constant surveillance, makeup water quality, the effect of freezing on gel-type anion exchange resin, a total organic carbon (TOC) survey, steam generator chemistry, steam generator inspection, condensate polisher operation, and management philosophy. The SEPREX condensate polishing process, in which the complete separation of the anion exchange resin from the cation exchange resin is achieved by flotation separation, is examined. It is concluded that the utilization of a condensate polishing process such as SEPREX provides the operating personnel at the plant with the necessary means to maintain the minimum desired level of contaminants within the steam generator.

  20. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    SciTech Connect

    Herer, C.

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

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

    SciTech Connect

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

    1982-08-01

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

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

    SciTech Connect

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

    1981-01-01

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

  3. Decontamination as a precursor to decommissioning. Status report Task 2: process evaluation. [PWR; BWR

    SciTech Connect

    Divine, J.R.; Woodruff, E.M.; McPartland, S.A.; Zima, G.E.

    1983-05-01

    As part of the US Nuclear Regulatory Commission's program to reduce occupational exposure and waste volumes, the Pacific Northwest Laboratory is studying decontamination as a precursor to decommissioning. Eleven processes or solvents were examined for their behavior in decontaminating BWR carbon steel samples. The solvents included NS-1, a proprietary solvent of Dow Chemical Corporation, designed for BWR use, and AP-Citrox, a well-known, two-step process designed for PWR stainless steel; it was used to provide a reference for later comparison to other systems and processes. The decontamination factors observed in the tests performed in a small laboratory scale recirculating loop ranged from about 1 (no effect) to 222 (about 99.6% of the initial activity removed. Coordinated corrosion measurements were made using twelve chemical solvents and eight metal alloys found in a range of reactor types.

  4. Effect of aging on the PWR Chemical and Volume Control System

    SciTech Connect

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K.

    1995-06-01

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  5. 3D Neutron Transport PWR Full-core Calculation with RMC code

    NASA Astrophysics Data System (ADS)

    Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien

    2014-06-01

    Nowadays, there are more and more interests in the use of Monte Carlo codes to calculate the detailed power density distributions in full-core reactors. With the Inspur TS1000 HPC Server of Tsinghua University, several calculations have been done based on the EDF 3D Neutron Transport PWR Full-core benchmark through large-scale parallelism. To investigate and compare the results of the deterministic method and Monte Carlo method, EDF R&D and Department of Engineering Physics of Tsinghua University are having a collaboration to make code to code verification. So in this paper, two codes are used. One is the code COCAGNE developed by the EDF R&D, a deterministic core code, and the other is the Monte Carlo code RMC developed by Department of Engineering Physics in Tsinghua University. First, the full-core model is described and a 26-group calculation was performed by these two codes using the same 26-group cross-section library provided by EDF R&D. Then the parallel and tally performance of RMC is discussed. RMC employs a novel algorithm which can cut down most of the communications. It can be seen clearly that the speedup ratio almost linearly increases with the nodes. Furthermore the cell-mapping method applied by RMC consumes little time to tally even millions of cells. The results of the codes COCAGNE and RMC are compared in three ways. The results of these two codes agree well with each other. It can be concluded that both COCAGNE and RMC are able to provide 3D-transport solutions associated with detailed power density distributions calculation in PWR full-core reactors. Finally, to investigate how many histories are needed to obtain a given standard deviation for a full 3D solution, the non-symmetrized condensed 2-group fluxes of RMC are discussed.

  6. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    SciTech Connect

    Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel.

  7. On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

    PubMed

    Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

    2016-08-01

    For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line.

  8. ALARA assessment of spent fuel and nuclear waste transportation systems

    SciTech Connect

    Sutherland, S. H.

    1980-01-01

    The effects of ALARA (as low as reasonably achievable) on transportation system costs were evaluated for LWR spent fuel, high-level commercial and defense wastes, and remotely handled TRU waste. Three dose rate specifications were used: 10 mrem/h at 2m, 5 mrem/h, and 2 mrem/h. The evaluation was done for wastes and LWR spent fuel 1, 3, 5, and 10 years old. Gamma shield materials were depleted uranium, lead, and steel; the neutron shield material was water. Results for a 7-element PWR cask show that uranium shielding is the lightest, and that the increased weight of the low dose rate casks results in 1 to 2 million dollars increase in lifetime transportation costs. 6 figures, 3 tables. (DLC)

  9. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    SciTech Connect

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  10. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOEpatents

    Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  11. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOEpatents

    Corletti, M.M.; Lau, L.K.; Schulz, T.L.

    1993-12-14

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  12. Spent-fuel composition: a comparison of predicted and measured data

    SciTech Connect

    Thomas, C.C. Jr.; Cobb, D.D.; Ostenak, C.A.

    1981-03-01

    The uncertainty in predictions of the nuclear materials content of spent light-water reactor fuel was investigated to obtain guidelines for nondestructive spent-fuel verification and assay. Values predicted by the reactor operator were compared with measured values from fuel reprocessors for six reactors (three PWR and three BWR). The study indicates that total uranium, total plutonium, fissile uranium, fissile plutonium, and total fissile content can be predicted with biases ranging from 1 to 6% and variabilities (1-sigma) ranging from 2 to 7%. The higher values generally are associated with BWRs. Based on the results of this study, nondestructive assay measurements that are accurate and precise to 5 to 10% (1sigma) or better should be useful for quantitative analyses of typical spent fuel.

  13. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    SciTech Connect

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  14. Testing and COBRA-SFS analysis of the VSC-17 ventilated concrete, spent fuel storage cask

    SciTech Connect

    McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

    1992-04-01

    A performance test of a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask loaded with 17 canisters of consolidated PWR spent fuel generating approximately 15 kW was conducted. The performance test included measuring the cask surface, concrete, air channel surface, and fuel temperatures, as well as cask surface gamma and neutron dose rates. Testing was performed using vacuum, nitrogen, and helium backfill environments. Pretest predictions of cask thermal performance were made using the COBRA-SFS computer code. Analysis results were within 15{degrees}C of measured peak fuel temperature. Peak fuel temperature for normal operation was 321{degrees}C. In general, the surface dose rates were less than 30 mrem/h on the side of the cask and 40 mrem/h on the top of the cask.

  15. Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels.

    SciTech Connect

    Wolf, S. F.

    1999-03-24

    The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns.

  16. Synthetic Fuel

    ScienceCinema

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2016-07-12

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  17. Synthetic Fuel

    SciTech Connect

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2008-03-26

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  18. Fuel cells

    NASA Astrophysics Data System (ADS)

    1984-12-01

    The US Department of Energy (DOE), Office of Fossil Energy, has supported and managed a fuel cell research and development (R and D) program since 1976. Responsibility for implementing DOE's fuel cell program, which includes activities related to both fuel cells and fuel cell systems, has been assigned to the Morgantown Energy Technology Center (METC) in Morgantown, West Virginia. The total United States effort of the private and public sectors in developing fuel cell technology is referred to as the National Fuel Cell Program (NFCP). The goal of the NFCP is to develop fuel cell power plants for base-load and dispersed electric utility systems, industrial cogeneration, and on-site applications. To achieve this goal, the fuel cell developers, electric and gas utilities, research institutes, and Government agencies are working together. Four organized groups are coordinating the diversified activities of the NFCP. The status of the overall program is reviewed in detail.

  19. Future Fuels

    DTIC Science & Technology

    2005-10-04

    tactical ground mobility and increasing operational reach • Identify, review, and assess – Technologies for reducing fuel consumption, including...T I O N S A C T I O N S TOR Focus - Tactical ground mobility - Operational reach - Not A/C, Ships, or troops Hybrid Electric Vehicle Fuel Management...Fuel Management During Combat Operations Energy Fundamentals • Energy Density • Tactical Mobility • Petroleum Use • Fuel Usage (TWV) • TWV OP TEMPO TOR

  20. Results of small break LOCA experiments in the LOFT reactor system with comparison to code calculations. [PWR

    SciTech Connect

    Adams, J.P.; Linebarger, J.H.; Leach, L.P.

    1980-01-01

    The results are presented of three small break loss-of-coolant experiments performed in the LOFT Pressurized Water Reactor (PWR) system. Experiment L3-0, performed without reactor power, represented a loss of coolant from the power operated relief valve on the top of the pressurizer. Experiments L3-1 and L3-2 were initiated with the reactor at full power (maximum linear heat generation rate approximately 52 kW/m) and represented 4-in and 1-in diameter breaks, respectively, in the reactor inlet piping of a commercial PWR. Comparisons of data to analytical model calculations with a number of different models indicate that most major phenomena were correctly calculated, but that improvements in modeling small break behavior are necessary.

  1. Nano-cavities observed in a 316SS PWR Flux Thimble Tube Irradiated to 33 and 70 dpa

    SciTech Connect

    Edwards, Danny J.; Garner, Francis A.; Bruemmer, Stephen M.; Efsing, Pal G.

    2009-02-28

    The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290ºC and 70 dpa at 315ºC were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during post-irradiation slow strain rate testing in PWR water conditions.

  2. Subchannel Thermal-Hydraulic Experimental Program (STEP). Volume 1. Mixing in a pressurized water reactor (PWR) rod bundle. Final report

    SciTech Connect

    Barber, A.R.; Zielke, L.A.

    1980-08-01

    This volume describes an experiment that was performed to determine the mixing characteristics of a pressurized water reactor (PWR) rod bundle. The objective of this project was to improve the subchannel computer code models of the reactor core. The experimental technique was isokinetic subchannel withdrawal of the entire flow from two sample subchannels. Once withdrawn, the sample fluid was condensed and its enthalpy was measured by regenerative heat exchange calorimetry. The test bundle was a 4 x 6 electrically heated array with a 50% power upset. The COBRA IIIC code was used to model the experiment and to determine the value of the thermal mixing coefficient, ..beta.., that was necessary to predict the measured results. Both single- and two-phase data were obtained over a range of PWR operating conditions. The results indicate that both single- and two-phase mixing is small. The COBRA model predicts the enthalpy data using a turbulent mixing coefficient, ..beta.. approx. = 0.002.

  3. Fossil Fuels.

    ERIC Educational Resources Information Center

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  4. Reactor-specific spent fuel discharge projections: 1985 to 2020

    SciTech Connect

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel.

  5. A Cherenkov viewing device for used-fuel verification

    NASA Astrophysics Data System (ADS)

    Attas, E. M.; Chen, J. D.; Young, G. J.

    1990-12-01

    A Cherenkov viewing device (CVD) has been developed to help verify declared inventories of used nuclear fuel stored in water bays. The device detects and amplifies the faint ultraviolet Cherenkov glow from the water surrounding the fuel, producing a real-time visible image on a phosphor screen. Quartz optics, a UV-pass filter and a microchannel-plate image-intensifier tube serve to form the image, which can be photographed or viewed directly through an eyepiece. Normal fuel bay lighting does not interfere with the Cherenkov light image. The CVD has been successfully used to detect anomalous PWR, BWR and CANDU (CANada Deuterium Uranium: registered trademark) fuel assemblies in the presence of normal-burnup assemblies stored in used-fuel bays. The latest version of the CVD, known as Mark IV, is being used by inspectors from the International Atomic Energy Agency for verification of light-water power-reactor fuel. Its design and operation are described, together with plans for further enhancements of the instrumentation.

  6. Reactor-specific spent fuel discharge projections: 1986 to 2020

    SciTech Connect

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs.

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

    SciTech Connect

    J.W. Davis

    1996-08-29

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

  8. Contribution of fuel vibrations to ex-core neutron noise during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor

    SciTech Connect

    Sweeney, F.J.; March-Leuba, J.; Smith, C.M.

    1984-01-01

    Noise measurements were performed during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor (PWR) to observe long-term changes in the ex-core neutron signatures. Increases in the ex-core neutron noise amplitude were observed throughout the 0.1- to 50.0-Hz range. In-core noise measurements indicate that fuel assembly vibrations contribute significantly to the ex-core neutron noise at nearly all frequencies in this range, probably due to mechanical or acoustic coupling with other vibrating internal structures. Space-dependent kinetics calculations show that ex-core neutron noise induced by fixed-amplitude fuel assembly vibrations will increase over a fuel cycle because of soluble boron and fuel concentration changes associated with burnup. These reactivity effects can also lead to 180/sup 0/ phase shifts between cross-core detectors. We concluded that it may be difficult to separate the changes in neutron noise due to attenuation (shielding) effects of structural vibrations from changes due to reactivity effects of fuel assembly motion on the basis of neutron noise amplitude or phase information. Amplitudes of core support barrel vibrations inferred from ex-core neutron noise measurements using calculated scale factors are likely to have a high degree of uncertainty, since these scale factors usually do not account for neutron noise generated by fuel assembly vibrations. Modifications in fuel management or design may also lead to altered neutron noise signature behavior over a fuel cycle.

  9. Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

    SciTech Connect

    Jason Hales; Various

    2014-06-01

    The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

  10. Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

    SciTech Connect

    Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.; Liu, Wenfeng; Hales, Jason; Stanek, Chris; Wirth, Brian D.

    2014-06-15

    The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

  11. Evaluation of stress corrosion cracking of irradiated 304L stainless steel in PWR environment using heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

    2016-08-01

    IASCC has been a major concern regarding the structural and functional integrity of core internals of PWR's, especially baffle-to-former bolts. Despite numerous studies over the past few decades, additional evaluation of the parameters influencing IASCC is still needed for an accurate understanding and modeling of this phenomenon. In this study, Fe irradiation at 450 °C was used to study the cracking susceptibility of 304 L austenitic stainless steel. After 10 MeV Fe irradiation to 5 dpa, irradiation-induced damage in the microstructure was characterized and quantified along with nano-hardness measurements. After 4% plastic strain in a PWR environment, quantitative information on the degree of strain localization, as determined by slip-line spacing, was obtained using SEM. Fe-irradiated material strained to 4% in a PWR environment exhibited crack initiation sites that were similar to those that occur in neutron- and proton-irradiated materials, which suggests that Fe irradiation may be a representative means for studying IASCC susceptibility. Fe-irradiated material subjected to 4% plastic strain in an inert argon environment did not exhibit any cracking, which suggests that localized deformation is not in itself sufficient for initiating cracking for the irradiation conditions used in this study.

  12. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    SciTech Connect

    Peterson, T.A.

    1997-06-01

    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  13. PwrSoC (integration of micro-magnetic inductors/transformers with active semiconductors) for more than Moore technologies

    NASA Astrophysics Data System (ADS)

    Mathuna, Cian Ó.; Wang, Ningning; Kulkarni, Santosh; Roy, Saibal

    2013-07-01

    This paper introduces the concept of power supply on chip (PwrSoC) which will enable the development of next-generation, functionally integrated, power management platforms with applications in dc-dc conversion, gate drives, isolated power transmission and ultimately, high granularity, on-chip, power management for mixed-signal, SOC chips. PwrSoC will integrate power passives with the power management IC, in a 3D stacked or monolithic form factor, thereby delivering the performance of a highefficiency dc-dc converter within the footprint of a low-efficiency linear regulator. A central element of the PwrSoC concept is the fabrication of power micro-magnetics on silicon to deliver micro-inductors and micro-transformers. The paper details the magnetics on silicon process which combines thin film magnetic core technology with electroplated copper conductors. Measured data for micro-inductors show inductance operation up to 20 MHz, footprints down to 0.5 mm2, efficiencies up to 93% and dc current carrying capability up to 600 mA. Measurements on micro-transformers show voltage gain of approximately - 1 dB at between 10 MHz and 30 MHz. Contribution to the Topical Issue “International Semiconductor Conference Dresden-Grenoble - ISCDG 2012”, Edited by Gérard Ghibaudo, Francis Balestra and Simon Deleonibus.

  14. Alternative fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J. S.; Butze, H. F.; Friedman, R.; Antoine, A. C.; Reynolds, T. W.

    1977-01-01

    Potential problems related to the use of alternative aviation turbine fuels are discussed and both ongoing and required research into these fuels is described. This discussion is limited to aviation turbine fuels composed of liquid hydrocarbons. The advantages and disadvantages of the various solutions to the problems are summarized. The first solution is to continue to develop the necessary technology at the refinery to produce specification jet fuels regardless of the crude source. The second solution is to minimize energy consumption at the refinery and keep fuel costs down by relaxing specifications.

  15. Development and Application of Laser Peening System for PWR Power Plants

    SciTech Connect

    Masaki Yoda; Itaru Chida; Satoshi Okada; Makoto Ochiai; Yuji Sano; Naruhiko Mukai; Gaku Komotori; Ryoichi Saeki; Toshimitsu Takagi; Masanori Sugihara; Hirokata Yoriki

    2006-07-01

    Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion cracking (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J

  16. EDF Nuclear Power Plants Operating Experience with MOX fuel

    SciTech Connect

    Thibault, Xavier

    2006-07-01

    EDF started Plutonium recycling in PWR in 1987 and progressively all the 20 reactors, licensed in using MOX fuel, have been loaded with MOX assemblies. At the origin of MOX introduction, these plants operated at full power in base load and the core management limited the irradiation time of MOX fuel assemblies to 3 annual cycles. Since 1995 all these reactors can operate in load follow mode. Since that time, a large amount of experience has been accumulated. This experience is very positive considering: - Receipt, handling, in core behaviour, pool storage and shipment of MOX fuel; - Operation of the various systems of the plant; - Environment impact; - Radioprotection; - Safety file requirements; - Availability for the grid. In order to reduce the fuel cost and to reach a better adequacy between UO{sub 2} fuel reprocessing flow and plutonium consumption, EDF had decided to improve the core management of MOX plants. This new core management call 'MOX Parity' achieves parity for MOX and UO{sub 2} assemblies in term of discharge burn-up. Compared to the current MOX assembly the Plutonium content is increased from 7,08% to 8,65% (equivalent to natural uranium enriched to respectively 3,25% and 3,7%) and the maximum MOX assembly burn-up moves from 42 to 52 GWd/t. This amount of burn-up is obtained from loading MOX assemblies for one additional annual cycle. Some, but limited, adaptations of the plant are necessary. In addition a new MOX fuel assembly has been designed to comply with the safety criteria taking into account the core management performances. These design improvements are based on the results of an important R and D program including numerous experimental tests and post-irradiated fuel examinations. In particular, envelope conditions compared to MOX Parity neutronic solicitations has been extensively investigated in order to get a full knowledge of the in reactor fuel behavior. Moreover, the operating conditions of the plant have been evaluated in many

  17. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect

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

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  18. CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

    SciTech Connect

    Ferrer, R.; Rhodes, J.; Smith, K.

    2012-07-01

    The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

  19. Measurement and analysis of fission gas release from BNFL's SBR MOX fuel

    NASA Astrophysics Data System (ADS)

    White, R. J.; Fisher, S. B.; Cook, P. M. A.; Stratton, R.; Walker, C. T.; Palmer, I. D.

    2001-01-01

    Puncture results are presented for seven SBR MOX fuel rods from the first prototypical commercial irradiation that was carried out in the Beznau-1 PWR. The rod average burn-up ranged from 31.2 to 35.6 MWd/kgHM. Comparison is made with the percentage of gas released from French MOX fuels and UO 2 fuel. The results show that in the burn-up range investigated, SBR MOX fuel and MIMAS MOX fuel perform similarly, releasing up to about 1% of the fission gas inventory. Comparisons with the Halden Criterion show that SBR MOX has the same release threshold as UO 2 and this suggests that the mechanisms of release in the two fuels are similar. This is further supported by calculations made with the ENIGMA fuel performance code. It is concluded that the apparent differences in fission gas release between SBR MOX and UO 2 fuel, at least in the early stages of release, can be explained by the higher temperatures experienced by MOX fuel.

  20. Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data

    SciTech Connect

    1997-11-01

    Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

  1. Key Issues for the control of refueling outage duration and costs in PWR Nuclear Power Plants

    SciTech Connect

    Degrave, Claude

    2002-07-01

    For several years, EDF, within the framework of the CIDEM1 project and in collaboration with some German Utilities, has undertaken a detailed review of the operating experience both of its own NPP and of foreign units, in order to improve the performances of future units under design, particularly the French-German European Pressurized Reactor (EPR) project. This review made it possible to identify the key issues allowing to decrease the duration of refueling and maintenance outages. These key issues can be classified in 3 categories Design, Maintenance and Logistic Support, Outage Management. Most of the key issues in the design field and some in the logistic support field have been studied and could be integrated into the design of any future PWR unit, as for the EPR project. Some of them could also be adapted to current plants, provided they are feasible and profitable. The organization must be tailored to each country, utility or period: it widely depends on the power production environment, particularly in a deregulation context. (author)

  2. Iodine partition coefficient measurements at simulated PWR steam generator conditions: Interim data report

    SciTech Connect

    Clinton, S.D.; Simmons, C.M.

    1987-05-01

    Iodine partition coefficients (defined as the ratio of the concentration of iodine species in the aqueous solution to the iodine concentration in the vapor phase) were measured at simulated PWR steam generator conditions (285C and 6.9 MPa), using carrier-free radioactive T I in the form of sodium iodide. The iodine tracer concentration was maintained at approx.6 x 10 mol/L; boric acid concentration was varied from 0 to 0.4 mol/L; and the solution pH (measured at 25C) was adjusted from 4 to 9 by the addition of lithium hydroxide. Iodine partition coefficients decrease with increasing boric acid concentration; however, the iodine volatility is essentially independent of the solution pH for a given boric acid concentration. Sparging the solutions with air at room temperature increases the iodine volatility by an order of magnitude, compared to that achieved with argon sparging. Iodine partition coefficient measurements ranged from a low of 200 (in 0.2 M boric acid sparged with air) to 400,000 (in purified water sparged with argon).

  3. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    SciTech Connect

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

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

    SciTech Connect

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

    1997-04-01

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

  5. Impact of makeup water system performance on PWR steam generator corrosion. Final report

    SciTech Connect

    Bell, M.J.; Pearl, W.L.; Sawochka, S.G.; Smith, L.A.

    1985-06-01

    The objectives of this project were to review makeup system design and performance and assess the possible relation of pressurized water reactor (PWR) steam generator corrosion to makeup water impurity ingress at fresh water sites. Project results indicated that makeup water transport of most ionic impurities can be expected to have a significant impact on secondary cycle chemistry only if condenser inleakage and other sources of impurities are maintained at very low levels. Since makeup water oxygen control techniques at most study plants were not consistent with state-of-the-art technology, oxygen input to the cycle via makeup can be significant. Leakage of colloidal silica and organics through makeup water systems can be expected to control blowdown silica levels and organic levels throughout the cycle at many plants. Attempts to correlate makeup water quality to steam generator corrosion observations were unsuccessful since (1) other impurity sources were significant compared to makeup at most study plants, (2) many variables are involved in the corrosion process, and (3) in the case of IGA, the variables have not been clearly established. However, in some situations makeup water can be a significant source of contaminants suspected to lead to both IGA and denting.

  6. Evaluation of on-line chelant addition to PWR steam generators. Steam generator cleaning project

    SciTech Connect

    Tvedt, T.J.; Wallace, S.L.; Griffin, F. Jr.

    1983-09-01

    The investigation of chelating agents for continuous water treatment of secondary loops of PWR steam generators were conducted in two general areas: the study of the chemistry of chelating agents and the study of materials compatability with chelating agents. The thermostability of both EDTA and HEDTA metal chelates in All Volatile Treatment (AVT) water chemistry were shown to be greater than or equal to the thermostability of EDTA metal chelates in phosphate-sulfite water chemistry. HEDTA metal chelates were shown to have a much greater stability than EDTA metal chelates. Using samples taken from the EDTA metal chelate thermostability study and from the Commonwealth Research Corporation (CRC) model steam generators (MSG), EDTA decomposition products were determined. Active metal surfaces were shown to become passivated when exposed to EDTA and HEDTA concentrations as high as 0.1% w/w in AVT. Trace amounts of iron in the water were found to increase the rate of passivation. Material balance and visual inspection data from CRC model steam generators showed that metal was transported through and cleaned from the MSG's. The Inconel 600 tubes of the salt water fouled model steam generators experienced pitting corrosion. Results of this study demonstrates the feasibility of EDTA as an on-line water treatment additive to maintain nuclear steam generators in a clean condition.

  7. Endurance tests of valves with cobalt-free hardfacing alloys: PWR phase

    SciTech Connect

    Murphy, E.V.; Inglis, I. )

    1992-05-01

    Atomic Energy of Canada Limited (AECL) is conducting endurance tests on valves hard-faced with four cobalt-free alloys. The first phase of the program, in which PWR primary heat transport conditions were simulated in AECL's valve test loop, has been completed. The candidate alloys are NOREM 01, NOREM 04, EB 5183 and EVERIT 50. One valve with Stellite 6 trim served as the standard. Prior to loop testing, a baseline inaugural inspection was performed. During testing the loop was shutdown at approximately 500 cycle intervals, and the valves were disassembled for examination. The examinations included seat leak tests, profilometry, nondestructive inspection and finally destructive examination. Corrosion coupons in the loop were used to monitor any material loss due solely to corrosion mechanisms. This report summarizes the final examination results and discusses the relative performance of the candidate alloys. The results indicate that, based upon the sliding wear damage assessment and seat leakage test results, all the candidate alloys perform better than the Stellite 6 control sample. On the same basis, NOREM 04 and EB 5183 are the best of the candidate alloys, although there are only minor differences in performance among the four alloys.

  8. Endurance tests of valves with cobalt-free hardfacing alloys: PWR phase. Final report

    SciTech Connect

    Murphy, E.V.; Inglis, I.

    1992-05-01

    Atomic Energy of Canada Limited (AECL) is conducting endurance tests on valves hard-faced with four cobalt-free alloys. The first phase of the program, in which PWR primary heat transport conditions were simulated in AECL`s valve test loop, has been completed. The candidate alloys are NOREM 01, NOREM 04, EB 5183 and EVERIT 50. One valve with Stellite 6 trim served as the standard. Prior to loop testing, a baseline inaugural inspection was performed. During testing the loop was shutdown at approximately 500 cycle intervals, and the valves were disassembled for examination. The examinations included seat leak tests, profilometry, nondestructive inspection and finally destructive examination. Corrosion coupons in the loop were used to monitor any material loss due solely to corrosion mechanisms. This report summarizes the final examination results and discusses the relative performance of the candidate alloys. The results indicate that, based upon the sliding wear damage assessment and seat leakage test results, all the candidate alloys perform better than the Stellite 6 control sample. On the same basis, NOREM 04 and EB 5183 are the best of the candidate alloys, although there are only minor differences in performance among the four alloys.

  9. Experimental investigation on denting in PWR steam generators: causes and corrective actions

    SciTech Connect

    Nordmann, F.; Brunet, J.P.; Duret, J.; Pinard-Legry, G.

    1983-10-01

    Denting studies have been undertaken in order to assess the influence of the most important parameters which could initiate corrosion of the carbon steel occurring in the tube-tube support plate crevices of some PWR steam generators. Tests have been carried out in model boilers where feedwater was polluted with sea or river water. Specific effects of chloride or sulfate and influence of oxygen content, magnetite addition and pH value were investigated. In magnetite prepacked crevices, denting is obtained within 1000 hrs for seawater pollution of 0.3 ppm chloride at the blowdown. In neutral chloride or in river water, denting is observed only with oxygen addition. Denting prevention is effective in the case of an on-line addition of phosphate, boric acid, or calcium hydroxide. For denting stopping, boric acid or calcium hydroxide is efficient even with a high seawater pollution. Soaks cannot stop denting if they are not followed by an on-line treatment (boric acid, calcium hydroxide). With quadrifoil holes, denting doesn't occur. In very severe test conditions, 13 percent Cr steel can be corroded, but the corrosion rate is low and oxide morphology is different from that growing on carbon steel.

  10. Survey of the literature on low-alloy steel fastener corrosion in PWR power plants

    SciTech Connect

    Hall, J.F.

    1984-12-01

    This report presents the results of a literature survey of low alloy steel fastener corrosion in PWR applications. The report addresses boric acid corrosion (accelerated general corrosion) and stress corrosion cracking of threaded fasteners used in primary pressure boundary closures, in secondary, auxiliary, and safety system closures and in component support applications. The report reviews and summarizes corrosion events that have occurred in domestic PWRs since 1968. Information provided for each event includes plant identification, year of event, major component or part affected, fastener material, fastener diameter, number of corroded studs, the service environments, the number of degraded fasteners and the results of post-service failure analyses. Possible corrective actions that are available to eliminate or mitigate the effects of the two types of corrosion are also identified. Laboratory test data, including some recent unpublished data, that are related to fastener corrosion are also discussed. The report also includes recommended additional work in the areas of boric acid corrosion, stress corrosion cracking and analytical methodologies to solve these fastener corrosion problems.

  11. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    SciTech Connect

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-07-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

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

    NASA Astrophysics Data System (ADS)

    Barbet, M.; Guillery, M.

    1981-06-01

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

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

    SciTech Connect

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

    1990-01-01

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

  14. CASMO-2 spent-fuel-rack criticality analysis

    SciTech Connect

    Napolitano, D.G.; Heinrichs, D.P.; Gorski, J.P.

    1986-01-01

    In recent years, utilities have needed to increase their spent-fuel storage capacity. Both Maine Yankee pressurized water reactor (PWR) and Vermont Yankee boiling water reactor (BWR) have increased their spent-fuel rack capacity by decreasing the canister center-to-center spacing while adding fixed poison. Licensing criticality analysis of such changes in spent-fuel rack design have been performed at Yankee Atomic Electric Co. (YAEC) using NITAWL-KENO-IV and the 123-group XSDRN library. However, KENO/Monte Carlo analysis has inherent drawbacks when applied to spent-fuel rack design and modification. These include statistical uncertainty and long computer time. In contrast, the transport theory code, CASMO-2, provides deterministic and fast criticality analysis. Also, since collapsed and transport-corrected cross sections are generated, PDQ can be used to analyze large array problems which are prohibitively expensive using KENO. In this work, the authors apply the CASMO-PDQ methodology to the Maine Yankee and Vermont Yankee high-density spent-fuel rack designs, and compare the final results against KENO.

  15. Fuel compositions

    SciTech Connect

    Zaweski, E.F.; Niebylski, L.M.

    1986-08-05

    This patent describes distillate fuel for indirect injection compression ignition engines containing, in an amount sufficient to minimize coking, especially throttling nozzle coking in the prechambers or swirl chambers of indirect injection compression ignition engines operated on such fuel, at least the combination of (i) organic nitrate ignition accelerator and (ii) an esterified cycle dehydration product of sorbitol which, when added to the fuel in combination with the organic nitrate ignition accelerator minimizes the coking.

  16. Fuel cells 101

    SciTech Connect

    Hirschenhofer, J.H.

    1999-07-01

    This paper discusses the various types of fuel cells, the importance of cell voltage, fuel processing for natural gas, cell stacking, fuel cell plant description, advantages and disadvantages of the types of fuel cells, and applications. The types covered include: polymer electrolyte fuel cell, alkaline fuel cell, phosphoric acid fuel cell; molten carbonate fuel cell, and solid oxide fuel cell.

  17. Fuel dehazers

    SciTech Connect

    Lyons, W.R.

    1986-03-01

    Hazy fuels can be caused by the emulsification of water into the fuel during refining, blending, or transportation operations. Detergent additive packages used in gasoline tend to emulsify water into the fuel. Fuels containing water haze can cause corrosion and contamination, and support microbiological growth. This results in problems. As the result of these problems, refiners, marketers, and product pipeline companies customarily have haze specifications. The haze specification may be a specific maximum water content or simply ''bright and clear'' at a specified temperature.

  18. Motor fuel

    SciTech Connect

    Burns, L.D.

    1982-07-13

    Liquid hydrocarbon fuel compositions are provided containing antiknock quantities of ashless antiknock agents comprising selected furyl compounds including furfuryl alcohol, furfuryl amine, furfuryl esters, and alkyl furoates.

  19. Alternative fuels

    SciTech Connect

    Not Available

    1991-07-01

    This paper presents the preliminary results of a review, of the experiences of Brazil, Canada, and New Zealand, which have implemented programs to encourage the use of alternative motor fuels. It will also discuss the results of a separate completed review of the Department of Energy's (DOE) progress in implementing the Alternative Motor Fuels Act of 1988. The act calls for, among other things, the federal government to use alternative-fueled vehicles in its fleet. The Persian Gulf War, environmental concerns, and the administration's National Energy Strategy have greatly heightened interest in the use of alternative fuels in this country.

  20. The Euratom Fast Collar (EFC): A Safeguards Instrument Design to Address Future Fuel Measurement Challenges

    SciTech Connect

    Evans, Louise; Swinhoe, Martyn T.; Menlove, Howard O.; Browne, Michael C.

    2012-08-13

    Summary of this presentation: (1) EFC instrument design for {sup 235}U verification measurements issued to EURATOM to issue a call for commercial tender; (2) Achieved a fast (Cd mode) measurement with less than 2% relative uncertainty in the doubles neutron counting rate in 10 minutes using a standard source strength; (3) Assay time in fast mode consistent with the needs of an inspector; (4) Extended to realistic calibration range for modern fuel designs - Relatively insensitive to gadolinia content for fuel designs with up to 32 burnable poison rods and 15 wt % gadolinia concentration, which is a realistic maximum for modern PWR fuel; (5) Improved performance over the standard thermal neutron collar with greater than twice the efficiency of the original design; (6) Novel tube pattern to reduce the impact of accidental pile-up; and (7) Joint test of prototype unit - EURATOM-LANL.

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

    SciTech Connect

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

    2015-03-04

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

  2. Spent fuel dry storage technology development: Report of consolidated thermal data

    NASA Astrophysics Data System (ADS)

    Lundberg, W. L.

    1980-09-01

    A drywell/sealed cask technique for spent fuel storage is discussed. Experiments indicate that PWR fuel with decay heat levels in excess of 2 kW could be stored in isolated drywells in Nevada test site soil without exceeding the current fuel clad temperature limit (715 F). The ability to thermally analyze near surface drywells and above ground storage casks is assessed. It is concluded that the required analysis procedures, computer programs, etc., are already developed and available. Soil thermal conductivity requires additional study to better understand the soil drying mechanism and effects of moisture. Work is also required to develop an internal canister subchannel model. In addition, the ability of the overall drywell thermal model to accommodate thermal interaction effects between adjacent drywells should be confirmed.

  3. A Model for Assessment of Failure of LWR Fuel during an RIA

    SciTech Connect

    Liu, Wenfeng; Kazimi, Mujid S.

    2007-07-01

    This paper presents a model for Pellet-Cladding Mechanical Interaction (PCMI) failure of LWR fuel during an RIA. The model uses the J-integral as a driving parameter to characterize the failure potential during PCMI. The model is implemented in the FRAPTRAN code and is validated by CABRI and NSRR simulated RIA test data. Simulation of PWR and BWR conditions are conducted by FRAPTRAN to evaluate the fuel failure potential using this model. Model validation and simulation results are compared with the strain-based failure model of PNNL and the SED/CSED model of EPRI. Our fracture mechanics model has good capability to differentiate failure from non-failure cases. The results reveal significant effects of power pulse width: a wider pulse width generally increases the threshold for fuel failure. However, this effect is less obvious for highly corroded cladding. (authors)

  4. Fuel injector

    DOEpatents

    Lambeth, Malcolm David Dick

    2001-02-27

    A fuel injector comprises first and second housing parts, the first housing part being located within a bore or recess formed in the second housing part, the housing parts defining therebetween an inlet chamber, a delivery chamber axially spaced from the inlet chamber, and a filtration flow path interconnecting the inlet and delivery chambers to remove particulate contaminants from the flow of fuel therebetween.

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

    SciTech Connect

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

    2006-07-01

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

  6. Fuel cladding behavior under rapid loading conditions

    NASA Astrophysics Data System (ADS)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  7. Subchannel Analysis with Mechanistic Methods for Thermo-Hydro Dynamics in BWR Fuel Bundles

    SciTech Connect

    Kentaro Imura; Kenji Yoshida; Isao Kataoka; Masanori Naitoh

    2006-07-01

    In order to predict the critical power or void fraction in BWR fuel bundles and the DNB heat flux of PWR fuel assemblies, the boiling transition analysis code called 'CAPE' with mechanistic models has been developed in the IMPACT project by NUPEC. The objective of the CAPE code development is to perform with good accuracy the safety evaluation for a new type or improved fuel bundle design of BWR and PWR without full-scale experiments or any tuning parameters in the analysis code. In this study, the CAPE for BWR was validated by the test analysis for 8 x 8 fuel bundles comparing with the void distribution data of the experimental data, which was carried out under several operational conditions in a BWR. The computations were carried out by changing the operational parameter such as the inlet subcooling, mass flow rate and the power output of the fuel bundles. Resultantly, the thermal equilibrium quality at the outlet ranges from 2% to 25%. From these results, though the predictive accuracy of the analytical results are in close agreement with the experimental data, it was noted that the errors were relatively outstanding in some subchannels, which was surrounded by the heated fuel rods and partially unheated walls, such as an unheated rod, a water rod and a separation wall of the channel box. The reason for this error is thought to be that the cross sectional void distribution was partially distributed in such subchannels surrounded partially by unheated wall, so the multidimensional void distribution structure might be formed in these subchannels. Under such conditions, it is very important to take into consideration the multidimensional structure of the two-phase flow in subchannel, and perhaps improve the estimation or correlations for the distribution parameter, as well as the amount of void exchange between neighboring subchannels. (authors)

  8. Fuel cell-fuel cell hybrid system

    DOEpatents

    Geisbrecht, Rodney A.; Williams, Mark C.

    2003-09-23

    A device for converting chemical energy to electricity is provided, the device comprising a high temperature fuel cell with the ability for partially oxidizing and completely reforming fuel, and a low temperature fuel cell juxtaposed to said high temperature fuel cell so as to utilize remaining reformed fuel from the high temperature fuel cell. Also provided is a method for producing electricity comprising directing fuel to a first fuel cell, completely oxidizing a first portion of the fuel and partially oxidizing a second portion of the fuel, directing the second fuel portion to a second fuel cell, allowing the first fuel cell to utilize the first portion of the fuel to produce electricity; and allowing the second fuel cell to utilize the second portion of the fuel to produce electricity.

  9. The effect of stainless steel overlay cladding on corrosion fatigue crack propagation in pressure vessel steel in PWR primary coolant

    SciTech Connect

    Bramwell, I.L.; Tice, D.R.; Worswick, D.; Heys, G.B.

    1995-12-31

    The growth of sub-critical cracks in pressure boundary materials in light water reactors is assessed using codified procedures, but the presence of the overlay-welded stainless steel cladding on the pressure vessel is not normally taken into consideration because of the difficulty in demonstrating clad integrity for the lifetime of the plant. In order to investigate any possible effect of the cladding layer on crack propagation, tests have been performed using two types of specimen. The first was sputter ion plated with a thin layer of austenitic stainless steel to simulate the electrochemical and oxide effects due to the cladding, whilst the second used an overlay clad specimen to investigate the behavior of a crack propagating from the austenitic into the ferritic material. Testing was carried out under cyclic loading conditions in well controlled simulated PWR primary water. At 288 C, the presence of stainless steel in contact with the low alloy steel did not enhance crack propagation in PWR primary coolant compared to unclad or unplated specimens. There was limited evidence that at 288 C under certain loading conditions, in both air and PWR water, there may be an effect of the cladding which reduces crack growth rates, at least for a short distance of crack propagation into the low alloy steel. Crack growth rates in the ferritic steel at 130 C were higher for both the plated and clad specimens than found in previous tests under similar conditions on the unclad material. However, the crack growth rates were bounded by current ASME 11 Appendix A recommendations for defects exposed to water and at low R ratio. There was no evidence of environmental enhancement of crack propagation in the stainless steel in clad specimens. The results indicate that the current approach of ignoring the cladding for assessment purposes is conservative at plant operating temperature.

  10. Experimental benchmark of MCNPX calculations against self-interrogation neutron resonance densitometry (SINRD) fresh fuel measurements

    SciTech Connect

    Menlove, Howard O; Swinhoe, Martyn T; La Fleur, Adrienne M; Charlton, William S; Lee, S Y; Tobin, S J

    2010-01-01

    We have investigated the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U concentration in a PWR 15 x 15 fresh LEU fuel assembly in air. Different measurement configurations were simulated in Monte Carlo N-Particle eXtended transport code (MCNPX) and benchmarked against experimental results. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,j) reaction peaks in fission chamber. Due to the low spontaneous fission rate of {sup 238}U (i.e. no curium in the fresh fuel), {sup 252}Cf sources were used to self-interrogate the fresh fuel pins. The resonance absorption of these neutrons in the fresh fuel pins can be measured using {sup 235}U fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the number of unknowns we are trying measure because the neutron source strength and detector-fuel assembly coupling cancel in the ratios. The agreement between MCNPX results and experimental measurements confirms the accuracy of the MCNPX models used. The development of SINRD to measure the fissile content in spent fuel is important to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in LWR spent fuel in water.

  11. An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect

    Scarangella, M. J.

    2012-07-01

    The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

  12. Issues in Three-Dimensional Depletion Analysis of Measured Data Near the End of a Fuel Rod

    SciTech Connect

    DeHart, Mark D; Gauld, Ian C; Suyama, Kenya

    2008-01-01

    The dynamics of reactor operation result in nonuniform axial-burnup profiles in fuel with any significant burnup. At the beginning of life in a pressurized water reactor (PWR), a near-cosine axial-shaped flux will begin depleting fuel near the axial center of a fuel assembly at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occur near the center. However, because of the high leakage near the end of the fuel assembly, burnup will drop off rapidly near the ends. Partial-length absorbers or nonuniform axial fuel loadings can further complicate the burnup profile. In a boiling water reactor, the same phenomena come into play, but the burnup profile is complicated by the significant variation of axial moderator density and by nonuniform axial loadings of burnable poison rods. Numerous studies of axial burnup effects have been published. However, most analyses performed in estimation of isotopic distributions due to axial burnup have been based on a set of two-dimensional (2-D) calculations performed for burnups that represent the axial burnup distribution in a fuel assembly. In general, this approach works quite well because the in-core axial gradient of the neutron flux is small over most of the length of the fuel rod, and the 2-D approximation is appropriate. Conversely, because the axial gradient becomes significant as one approaches either end of the fuel assembly, the 2-D approximation begins to break down at that point. It has been theorized that axial leakage will lead to a reduced fast flux relative to the thermal flux, softening the spectrum near the ends of the fuel, and that a 2-D approximation is conservative in that it provides more plutonium production. This has not been put the test, however, for two reasons--a lack of good three-dimensional (3-D) analysis methods acceptable for away-from-reactor applications and, more importantly, a

  13. ATHOS: a computer program for thermal-hydraulic analysis of steam generators. Volume 3. User's manual. [PWR

    SciTech Connect

    Singhal, A.K.; Keeton, L.W.; Przekwas, A.J.; Weems, J.S.

    1982-10-01

    ATHOS (Analysis of the Thermal Hydraulics of Steam Generators) is a computer code developed by CHAM of North America Incorporated, under the contract RP 1066-1 from the Electric Power Research Institute, Palo Alto, California. ATHOS supersedes the earlier code URSULA2. ATHOS is designed for three-dimensional, steady state and transient analyses of PWR steam generators. The current version of the code has been checked out for: three different configurations of the recirculating-type U-tube steam generators; the homogeneous and algebraic-slip flow models; and full and part load operating conditions.

  14. Evaluation of the /sup 252/Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

    SciTech Connect

    Mihalczo, J.T.

    1987-01-01

    The /sup 252/Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt light water reactor (LWR) fuel submerged in fuel storage pools, fully loaded and as they are being loaded. Measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This, in turn, could lead to more cost-effective cask designs. Evaluation of the method for this application was based on experiments already completed at a critical experiments facility using arrays of pressurized water reactor (PWR) fuel pins typical of the size of storage cask configurations, the existence of neutron detectors that can function in shipping cask environments, and the ability to construct ionization chambers containing /sup 252/Cf of adequate intensity for these measurements.

  15. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect

    Osamu KAawabata; Mitsuhiro Kajimoto

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

  16. On-line PWR RHR pump performance testing following motor and impeller replacement

    SciTech Connect

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  17. Increasing Fuel Utilization of Breed and Burn Reactors

    NASA Astrophysics Data System (ADS)

    Di Sanzo, Christian Diego

    Breed and Burn reactors (B&B), also referred to Traveling Wave Reactors, are fast spectrum reactors that can be fed indefinitely with depleted uranium only, once criticality is achieved without the need for fuel reprocessing. Radiation damage to the fuel cladding limits the fuel utilization of B&B reactors to ˜ 18-20% FIMA (Fissions of Initial Metal Atoms) -- the minimum burnup required for sustaining the B&B mode of operation. The fuel discharged from this type of cores contain ˜ 10% fissile plutonium. Such a high plutonium content poses environmental and proliferation concerns, but makes it possible to utilize the fuel for further energy production. The objectives of the research reported in this dissertation are to analyze the fuel cycle of B&B reactors and study new strategies to extend the fuel utilization beyond ˜ 18-20% FIMA. First, the B&B reactor physics is examined while recycling the fuel every 20% FIMA via a limited separation processing, using either the melt refining or AIROX dry processes. It was found that the maximum attainable burnup varies from 54% to 58% FIMA -- depending on the recycling process and on the fraction of neutrons lost via leakage and reactivity control. In Chapter 3 the discharge fuel characteristics of B&B reactors operating at 20% FIMA and 55% FIMA is analyzed and compared. It is found that the 20% FIMA reactor discharges a fuel with about ˜ 80% fissile plutonium over total plutonium content. Subsequently a new strategy of minimal reconditioning, called double cladding is proposed to extend the fuel utilization in specifically designed second-tier reactors. It is found that with this strategy it is possible to increase fuel utilization to 30% in a sodium fast reactor and up to 40% when a subcritical B&B core is driven by an accelerator-driven spallation neutron source. Lastly, a fuel cycle using Pressurized Water Reactors (PWR) to reduce the plutonium content of discharged B&B reactors is analyzed. It was found that it is

  18. Fuels research: Fuel thermal stability overview

    NASA Technical Reports Server (NTRS)

    Cohen, S. M.

    1980-01-01

    Alternative fuels or crude supplies are examined with respect to satisfying aviation fuel needs for the next 50 years. The thermal stability of potential future fuels is discussed and the effects of these characteristics on aircraft fuel systems are examined. Advanced fuel system technology and design guidelines for future fuels with lower thermal stability are reported.

  19. Fuel ethanol

    SciTech Connect

    Not Available

    1989-02-01

    This report discusses the Omnibus Trade and Competitiveness Act of 1988 which requires GAO to examine fuel ethanol imports from Central America and the Caribbean and their impact on the U.S. fuel ethanol industry. Ethanol is the alcohol in beverages, such as beer, wine, and whiskey. It can also be used as a fuel by blending with gasoline. It can be made from renewable resources, such as corn, wheat, grapes, and sugarcane, through a process of fermentation. This report finds that, given current sugar and gasoline prices, it is not economically feasible for Caribbean ethanol producers to meet the current local feedstock requirement.

  20. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    SciTech Connect

    Ellis, Ronald James

    2011-01-01

    A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO{sub 2}, clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in {sup 235}U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U{sub 3}O{sub 8}-Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B{sub 4}C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of {sup 235}U and 2.8 g of {sup 10}B. The thermal neutron flux in the flux trap region can exceed 2.5 x 10{sup 15} n/cm{sup 2} {center_dot} s while the fast flux in this region exceeds 1 x 10{sup 15} n/cm{sup 2} {center_dot} s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions

  1. Metal Matrix Microencapsulated (M3) fuel neutronics performance in PWRs

    SciTech Connect

    Fratoni, Massimiliano; Terrani, Kurt A

    2012-01-01

    superior to the conventional oxide fuel since PCMI and rod burst, respectively, have been mitigated. Under beyond design basis accident scenarios where the fuel is exposed to high temperature steam for prolonged periods, larger inventory of zirconium metal in the core could negatively affect the accident progression. A thin steam resistant layer (e.g. alumina forming alloy steel), integrated into the solid rod during fabrication by co-extrusion or hot-isostatic-pressing, offers the potential to provide additional fuel protection from steam interaction: blanketing under a range of boiling regimes and under severe accident conditions up to high temperatures well beyond what is currently possible in the conventional fuel. A crucial aspect to the viability of M3 fuel in light water reactors is the reduced heavy metal load compared to standard pellet fuel. This study evaluated the design requirements to operate a Pressurized Water Reactor (PWR) with M3 fuel in order to obtain fuel cycle length, reactivity coefficients, and power peaking factors comparable to that of standard fuel.

  2. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  3. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.

  4. COMMIX-1A analysis of fluid and thermal mixing in a model cold leg and downcomer of a PWR

    SciTech Connect

    Chen, B.C.J.; Cha, B.K.; Miao, C.C.; Sha, W.T.; Kim, J.H.; Sun, B.K.H.

    1983-01-01

    The issue of thermal shock of a PWR pressure vessel has been under considerable attention recently. A number of experimental as well as analytical studies have been performed to investigate the effect of the thermal transient on the pressure vessel due to the high pressure injection (HPI) of the cold fluid into the cold leg. This process has been called Pressurized Thermal Shock (PTS). This paper is an analytical study of PTS by using COMMIX-1A. Experimental investigations were performed at CREARE and SAI. In the CREARE experiment, a 1/5 scale model was set up to simulate a cold leg and downcomer of a PWR. Tests with several different ratios of hot loop flow versus cold HPI flow were performed to study the effect of the flow ratio on the fluid and thermal mixing process in the system, especially in the downcomer region. Analytical investigations also proceeded in parallel with the experiments. Quite a few analytical investigations were performed with the COMMIX-1A code. However, in this version of COMMIX, the effect of the numerical diffusion was not addressed.

  5. A Review of Thorium Utilization as an option for Advanced Fuel Cycle--Potential Option for Brazil in the Future

    SciTech Connect

    Maiorino, J.R.; Carluccio, T.

    2004-10-03

    Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO{sub 2}, as will be reviewed in this paper, and therefore R&D programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle.

  6. TRAC-PF1/MOD1 US/Japanese PWR conservative LOCA prediction

    SciTech Connect

    Gruen, G E; Fisher, J E

    1987-11-01

    This report documents the results of a 200%, double-ended, cold-leg-break, loss-of-coolant-accident (LOCA) calculation using the TRAC-PF1/MOD1 computer code. The reactor system represented a typical United States/Japanese pressurized water reactor with a 15 x 15 fuel bundle arrangement 12-ft long, four loops, and cold-leg Emergency Core Cooling (ECC) Systems. Conservation boundary and initial conditions were used. Reactor power was 102% of the 3250 MWt rated power, decay heat was set to 120% of American Nuclear Society Standard 5.1, highest core lifetime values for power peaking and fuel stored energy were used, and the LOCA occurred simultaneously with a loss of offsite power. Best estimate assumptions were used for the break flow model, fuel rod heat transfer and metal-water reaction correlations, and steady-state fuel temperature profiles. A flow blockage model, having the capability to account for the effects of cladding ballooning or rupturing, was not used. Except for these best estimate assumptions, the boundary and initial conditions were consistent with those used in licensing calculations. Maximum fuel rod temperatures were 1380 K (2020/sup 0/F) and 1040 K (1410/sup 0/F) on the hottest evaluation model rod and hottest best estimate rod, respectively. The high reported values or fuel cladding temperature were a direct consequence of the conservative boundary and initial conditions used for the calculation, primarily the 2% overpower condition, the core decay heat assumption, and the degraded ECCS. The calculation demonstrated successful core reflooding before 1478 K (2200/sup 0/F) cladding temperature was exceeded on any fuel rod. 7 refs., 47 figs., 5 tabs.

  7. Shippingport Spent Fuel Canister System Description

    SciTech Connect

    JOHNSON, D.M.

    2001-06-26

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSH will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available.

  8. Shippingport Spent Fuel Canister System Description

    SciTech Connect

    JOHNSON, D.M.

    2000-03-27

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available.

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

  10. Analysis of a Defected Dissimilar Metal Weld in a PWR Power Plant

    SciTech Connect

    Efsing, P.; Lagerstrom, J.

    2002-07-01

    During the refueling outage 2000, inspections of the RC-loops of one of the Ringhals PWR-units, Ringhals 4, indicated surface breaking defects in the axial direction of the piping in a dissimilar weld between the Low alloy steel nozzle and the stainless safe end in the hot leg. In addition some indications were found that there were embedded defects in the weld material. These defects were judged as being insignificant to the structural integrity. The welds were inspected in 1993 with the result that no significant indications were found. The weld it self is a double U weld, where the thickness of the material is ideally 79,5 mm. Its is constructed by Inconel 182 weld material. At the nozzle a buttering was applied, also by Inconel 182. The In-service inspection, ISI, of the object indicated four axial defects, 9-16 mm deep. During fabrication, the areas where the defects are found were repaired at least three times, onto a maximum depth of 32 mm. To evaluate the defects, 6 boat samples from the four axial defects were cut from the perimeter and shipped to the hot-cell laboratory for further examination. This examination revealed that the two deep defects had been under sized by the ISI outside the requirement set by the inspection tolerances, while the two shallow defects were over sized, but within the tolerances of the detection system. When studying the safety case it became evident that there were several missing elements in the way this problems is handled with respect to the Swedish safety evaluation code. Among these the most notable at the beginning was the absence of reliable fracture mechanical data such as crack growth laws and fracture toughness at elevated temperature. Both these questions were handled by the project. The fracture mechanical evaluation has focused on a fit for service principal. Thus defects both in the unaffected zones and the disturbed zones, boat sample cutouts, of the weld have been analyzed. With reference to the Swedish safety

  11. Alternative jet aircraft fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1979-01-01

    Potential changes in jet aircraft fuel specifications due to shifts in supply and quality of refinery feedstocks are discussed with emphasis on the effects these changes would have on the performance and durability of aircraft engines and fuel systems. Combustion characteristics, fuel thermal stability, and fuel pumpability at low temperature are among the factors considered. Combustor and fuel system technology needs for broad specification fuels are reviewed including prevention of fuel system fouling and fuel system technology for fuels with higher freezing points.

  12. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, R.E.

    1988-03-08

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream 1 and spent fuel stream 2. Spent fuel stream 1 is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream 1 and exhaust stream 2, and exhaust stream 1 is vented. Exhaust stream 2 is mixed with spent fuel stream 2 to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells. 1 fig.

  13. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, Ralph E.

    1988-01-01

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

  14. Dosimetry Evaluation of In-Core and Above-Core Zirconium Alloy Samples in a PWR

    NASA Astrophysics Data System (ADS)

    Amiri, Benjamin W.; Foster, John P.; Greenwood, Larry R.

    2016-02-01

    A description of the neutron fluence analysis of activated zirconium alloys samples at a Westinghouse 3-loop reactor is presented. These samples were irradiated in the core and in the fuel plenum region, where dosimetry measurements are relatively rare compared with regions radially outward of the core. Dosimetry measurements performed by Batelle/PNNL are compared to the calculational models. Good agreement is shown with the in-core measurements when using analysis conditions expected to best represent this region, such as an assembly-specific axial power distribution. However, the use of these conditions to evaluate dosimetry in the fuel plenum region can lead to significant underestimation of the fluence. The use of a flat axial power distribution, however, does not underestimate the fluence in the fuel plenum region.

  15. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

    A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

  16. Fuel bundle

    SciTech Connect

    Lui, C.K.

    1989-04-04

    This patent describes a method of forming a fuel bundle of a nuclear reactor. The method consists of positioning the fuel rods in the bottom plate, positioning the tie rod in the bottom plate with the key passed through the receptacle to the underside of the bottom plate and, after the tie rod is so positioned, turning the tie rod so that the key is in engagement with the underside of the bottom plate. Thereafter mounting the top plate is mounted in engagement with the fuel rods with the upper end of the tie rod extending through the opening in the top plate and extending above the top plate, and the tie rod is secured to the upper side of sid top plate thus simultaneously securing the key to the underside of the bottom plate.

  17. FUEL ELEMENT

    DOEpatents

    Howard, R.C.; Bokros, J.C.

    1962-03-01

    A fueled matrlx eontnwinlng uncomblned carbon is deslgned for use in graphlte-moderated gas-cooled reactors designed for operatlon at temperatures (about 1500 deg F) at which conventional metallic cladding would ordlnarily undergo undesired carburization or physical degeneratlon. - The invention comprlses, broadly a fuel body containlng uncombined earbon, clad with a nickel alloy contalning over about 28 percent by' weight copper in the preferred embodlment. Thls element ls supporirted in the passageways in close tolerance with the walls of unclad graphite moderator materlal. (AEC)

  18. Fabrication Characteristics of Large Grain DUPIC Fuel Using SIMFUEL

    SciTech Connect

    Park, Geun IL; Lee, Jung Won; Lee, Jae Won; Yang, Myung Seung; Song, Kee Chan

    2007-07-01

    Fabrication characteristics to improve the density and grain size of DUPIC fuel with relation to its fuel performance were experimentally evaluated using SIMFUEL as a surrogate for an actual spent PWR fuel due to the high radioactivity of a spent fuel. Hence, SIMFUELs with a burn-up of 35,000 MWd/tU and 60,000 MWd/tU were used to investigate the influence of fission products contents as an impurity on the fuel powder properties and on the density and grain size of a simulated DUPIC pellet. In order to improve the densification and grain growth of the simulated DUPIC fuel, the effect of the addition of sintering aids was investigated. The specific surface area of the OREOX powders was increased with an increase of the impurities by the dissolved oxides in UO{sub 2} among the impurity groups. The specific surface area of the powders milled after the OXREOX treatment was slightly higher than the UO{sub 2} powder used for a nuclear power plant, thus resulting in sintered pellets with a higher than 95% T.D. (theoretical density). The grain size of the sintered pellets was significantly decreased with increasing amount of the metallic and oxide precipitates. However, on adding the sintering aids such as TiO{sub 2} and Nb{sub 2}O{sub 5}, the grain size of the sintering aids-doped pellets was greatly improved by up to around 3 times that of the raw pellets and their sintered density was also increased by up to 2%. (authors)

  19. Weapons-Grade Plutonium-Thorium PWR Assembly Design and Core Safety Analysis

    SciTech Connect

    Dziadosz, David; Ake, Timothy N.; Saglam, Mehmet; Sapyta, Joe J.

    2004-07-15

    A light water reactor (LWR) fuel assembly design consisting of a blend of weapons-grade plutonium and natural thorium oxides was examined. The design meets current thermal-hydraulic and safety criteria. Such an assembly would have enough reactivity to achieve three cycles of operation. The pin power distribution indicates a fairly level distribution across the assembly, avoiding hot spots near guide tubes, corners, and other sections where excessive power would create significant loss to thermal-hydraulic margins.This work examined a number of physics and core safety analysis parameters that impact the operation and safety of power reactors. Such parameters as moderator coefficients of reactivity, Doppler coefficients, soluble boron worth, control rod worth, prompt neutron lifetime, and delayed-neutron fractions were considered. These in turn were used to examine reactor behavior during a number of operational conditions, transients, and accidents. Such conditions as shutdown from power with one rod stuck out, steam-line break accident, feedwater line break, loss of coolant flow, locked rotor accidents, control rod ejection accidents, and anticipated transients without scram (ATWSs) were examined.The analysis of selected reactor transients demonstrated that it is feasible to license and safely operate a reactor fueled with plutonium-thorium blended fuel. In most cases analyzed, the thorium mixture had less-severe consequences than those for a core comprising low-enriched uranium fuel. In the analyzed cases where the consequences were more severe, they were still within acceptable limits. The ATWS accident condition requires more analysis.

  20. Studies of Flexible MOX/LEU Fuel Cycles

    SciTech Connect

    Adams, M.L.; Alonso-Vargas, G.

    1999-03-01

    This project was a collaborative effort involving researchers from Oak Ridge National Laboratory and North Carolina State University as well as Texas A and M University. The background, briefly, is that the US is planning to use some of its excess weapons Plutonium (Pu) to make mixed-oxide (MOX) fuel for existing light-water reactors (LWRs). Considerable effort has already gone into designing fuel assemblies and core loading patterns for the transition from full-uranium cores to partial-MOX and full-MOX cores. However, these designs have assumed that any time a reactor needs MOX assemblies, these assemblies will be supplied. In reality there are many possible scenarios under which this supply could be disrupted. It therefore seems prudent to verify that a reactor-based Pu-disposition program could tolerate such interruptions in an acceptable manner. Such verification was the overall aim of this project. The task assigned to the Texas A and M team was to use the HELIOS code to develop libraries of two-group homogenized cross sections for the various assembly designs that might be used in a Westinghouse Pressurized Water Reactor (PWR) that is burning weapons-grade MOX fuel. The NCSU team used these cross sections to develop optimized loading patterns under several assumed scenarios. Their results are documented in a companion report.

  1. Nuclear Fuels.

    ERIC Educational Resources Information Center

    Nash, J. Thomas

    1983-01-01

    Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

  2. Fuel Cells

    ERIC Educational Resources Information Center

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  3. Future Fuel.

    ERIC Educational Resources Information Center

    Stover, Del

    1991-01-01

    Tough new environmental laws, coupled with fluctuating oil prices, are likely to prompt hundreds of school systems to examine alternative fuels. Literature reviews and interviews with 45 government, education, and industry officials provided data for a comparative analysis of gasoline, diesel, natural gas, methanol, and propane. (MLF)

  4. Alternative Fuels

    DTIC Science & Technology

    2009-06-11

    Swedish Biofuels AB • Cellulosic and algal feedstocks that are non-competitive with food material $ P r o d u c t P r o d u c t Traditional fuels...JP-8 BACK-UP SLIDES Unclassified 19 What Are Biofuels ? Cellulose “first generation”“second generation” C18:0 C16:1 Triglycerides (fats, oils

  5. Interim storage technology of spent fuel and high-level waste in Germany

    SciTech Connect

    Geiser, H.; Schroder, J.

    2007-07-01

    The idea of using casks for interim storage of spent fuel arose at GNS after a very controversial political discussion in 1978, when total passive safety features (including aircraft crash conditions) were required for an above ground spent fuel storage facility. In the meantime, GNS has loaded more than 1000 casks at 25 different storage sites in Germany. GNS cask technology is used in 13 countries. Spent fuel assemblies of PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high level waste containers are stored in full metal casks of the CASTOR{sup R} type. Also MOX fuel of PWR and BWR has been stored. More than two decades of storage have shown that the basic requirements (safe confinement, criticality safety, sufficient shielding and appropriate heat transfer) have been fulfilled in any case - during normal operation and in case of severe accidents, including aircraft crash. There is no indication of problems arising in the future. Of course, the experience of more than 20 years has resulted in improvements of the cask design. The CASTOR{sup R} casks have been thoroughly investigated by many experiments. There have been approx. 50 full and half scale drop tests and a significant number of fire tests, simulations of aircraft crash, investigations with anti tank weapons, and an explosion of a railway tank with liquid gas neighbouring a loaded CASTOR{sup R} cask. According to customer and site specific demands, different types of storage facilities are realized in Germany. Firstly, there are facilities for long-term storage, such as large ventilated central storage buildings away from reactor or ventilated storage buildings at the reactor site, ventilated underground tunnels or concrete platforms outside a building. Secondly, there are facilities for temporary storage, where casks have been positioned in horizontal orientation under a ventilated shielding cover outside a building. (authors)

  6. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    SciTech Connect

    Wagner, J. C.

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  7. Safety assessment of plutonium mixed oxide fuel irradiated up to 37.7 GW day tonne-1

    NASA Astrophysics Data System (ADS)

    Somers, J.; Papaioannou, D.; McGinley, J.; Sommer, D.

    2013-06-01

    In this irradiation test, the safety performance of (Th,Pu)O2 fuel was evaluated. The fuel pellets were synthesised from powders prepared using a sol gel method to give a product exhibiting an atomically homogeneous distribution of the elements. The fuel pellets, of conventional pressurised water reactor (PWR) dimensions, were encapsulated in zircaloy cladding, and irradiated during four reactor cycles, reaching a burnup of 37.7 GW day tonne-1 in the KWO pressurised water reactor at Obrigheim, Germany. The irradiation test was performed under representative conditions. Intermediate inspection of the fuel pin during reactor outages revealed a cladding creep down within the bounds observed for UO2 fuels under similar conditions. Hydriding of the cladding was found predominantly on the outer liner of the duplex cladding. Fission gas analysis revealed a release of about 0.5%, which is somewhat lower than U-MOX fuels at the same burnup, but the latter were operated at higher linear heating rate. The Xe/Kr ratio of 11 is much lower than (U,Pu)O2 fuel (typically 16), indicating significant 233U generation and fissioning thereof during the irradiation experiment. Examination of the microstructure indicates that the pellet - cladding gap is almost closed. The grain size remained similar to the fresh fuel (4 μm) and no intragranular porosity was observed.

  8. Severe accidents in spent fuel pools in support of generic safety, Issue 82

    SciTech Connect

    Sailor, V.L.; Perkins, K.R.; Weeks, J.R.; Connell, H.R.

    1987-07-01

    This investigation provides an assessment of the likelihood and consequences of a severe accident in a spent fuel storage pool - the complete draining of the pool. Potential mechanisms and conditions for failure of the spent fuel, and the subsequent release of the fission products, are identified. Two older PWR and BWR spent fuel storage pool designs are considered based on a preliminary screening study which tried to identify vulnerabilities. Internal and external events and accidents are assessed. Conditions which could lead to failure of the spent fuel Zircaloy cladding as a result of cladding rupture or as a result of a self-sustaining oxidation reaction are presented. Propagation of a cladding fire to older stored fuel assemblies is evaluated. Spent fuel pool fission product inventory is estimated and the releases and consequences for the various cladding scenarios are provided. Possible preventive or mitigative measures are qualitatively evaluated. The uncertainties in the risk estimate are large, and areas where additional evaluations are needed to reduce uncertainty are identified.

  9. Modeling depletion simulations for a high-burnup, highly heterogeneous BWR fuel assembly with scale

    SciTech Connect

    Smith, H. J.

    2012-07-01

    Extensive SCALE isotopic validation studies have been performed for various PWR fuel assembly designs and operating conditions, and to a lesser extent for BWR fuel assembly designs. However, no SCALE validation work has been documented for newer, highly heterogeneous BWR fuel assembly designs at high burnup. Isotopic benchmark calculations of the earlier, more geometrically uniform BWR fuel assemblies are less sensitive to simplification of the operating history details and certain modeling assumptions than heterogeneous fuel assemblies, particularly at high burnup. This analysis shows the capability of SCALE to simulate a complex highly heterogeneous SVEA96 Optima fuel assembly and illustrates the importance of the need for the highest possible accuracy and precision in isotope measurements intended to be used as benchmark-quality results. In addition, this analysis quantifies the impact of various modeling assumptions on the results. The sample for which the simulation results are reported here achieved a burnup 62 GWd/MTU and was analyzed as part of the MALIBU Extension program. (authors)

  10. The evaluation of iron-base hardfacing alloys on gate valves after cycling under simulated PWR conditions for one year

    SciTech Connect

    Murphy, E.V.; Inglis, I.; Ocken, H.

    1992-12-31

    Gate valves hardfaced with iron-base alloys were exposed for about one year to simulated PWR conditions. The hardfacing alloys tested were EB 5183, EVERIT 50, NOREM 01 and NOREM 04. A gate valve with Satellite 6 was included in the test program as a control standard. During the test period the valves were opened and closed 2000 times. The performance of the valves was assessed by periodic leak tests and visual and profilometric characterisation of sealing surfaces. At the end of the test program, the seats and discs were destructively examined. The various examinations indicated all the iron-base alloys were superior to Satellite 6. Based on the results of hot leakage tests, one valve with EB 5183 and the valve with NOREM 04 were the best performers.

  11. Effects of PbO on the oxide films of incoloy 800HT in simulated primary circuit of PWR

    NASA Astrophysics Data System (ADS)

    Tan, Yu; Yang, Junhan; Wang, Wanwan; Shi, Rongxue; Liang, Kexin; Zhang, Shenghan

    2016-05-01

    Effects of trace PbO on oxide films of Incoloy 800HT were investigated in simulated primary circuit water chemistry of PWR, also with proper Co addition. The trace PbO addition in high temperature water blocked the protective spinel oxides formation of the oxide films of Incoloy 800HT. XPS results indicated that the lead, added as PbO into the high temperature water, shows not only +2 valance but also +4 and 0 valances in the oxide film of 800HT co-operated with Fe, Cr and Ni to form oxides films. Potentiodynamic polarization results indicated that as PbO concentration increased, the current densities of the less protective oxide films of Incoloy 800HT decreased in a buffer solution tested at room temperature. The capacitance results indicated that the donor densities of oxidation film of Incoloy 800HT decreased as trace PbO addition into the high temperature water.

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

    SciTech Connect

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

    1981-11-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  14. Propagation of Isotopic Bias and Uncertainty to Criticality Safety Analyses of PWR Waste Packages

    SciTech Connect

    Radulescu, Georgeta

    2010-06-01

    Burnup credit methodology is economically advantageous because significantly higher loading capacity may be achieved for spent nuclear fuel (SNF) casks based on this methodology as compared to the loading capacity based on a fresh fuel assumption. However, the criticality safety analysis for establishing the loading curve based on burnup credit becomes increasingly complex as more parameters accounting for spent fuel isotopic compositions are introduced to the safety analysis. The safety analysis requires validation of both depletion and criticality calculation methods. Validation of a neutronic-depletion code consists of quantifying the bias and the uncertainty associated with the bias in predicted SNF compositions caused by cross-section data uncertainty and by approximations in the calculational method. The validation is based on comparison between radiochemical assay (RCA) data and calculated isotopic concentrations for fuel samples representative of SNF inventory. The criticality analysis methodology for commercial SNF disposal allows burnup credit for 14 actinides and 15 fission product isotopes in SNF compositions. The neutronic-depletion method for disposal criticality analysis employing burnup credit is the two-dimensional (2-D) depletion sequence TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)/NEWT (New ESC-based Weighting Transport code) and the 44GROUPNDF5 crosssection library in the Standardized Computer Analysis for Licensing Evaluation (SCALE 5.1) code system. The SCALE 44GROUPNDF5 cross section library is based on the Evaluated Nuclear Data File/B Version V (ENDF/B-V) library. The criticality calculation code for disposal criticality analysis employing burnup credit is General Monte Carlo N-Particle (MCNP) Transport Code. The purpose of this calculation report is to determine the bias on the calculated effective neutron multiplication factor, k{sub eff}, due to the bias and bias uncertainty associated with

  15. Enhancing VVER Annular Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect

    G. S. Chang

    2007-06-01

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. The merits of nuclear energy are the high-density energy, and low environmental impacts i.e. almost zero greenhouse gas emission. Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current LWR as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce the spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope 238Pu /Pu ratio. For future advanced nuclear systems, the minor actinides are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. In this paper, a typical pressurized water reactor (PWR) VVER-1000 annular fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. We concluded that the concept of MARA, involves the use of transuranic nuclides (237Np and/or 241Am), can not only drastically

  16. Fuel cell

    SciTech Connect

    Struthers, R.C.

    1983-06-28

    An improved fuel cell comprising an anode section including an anode terminal, an anode fuel, and an anolyte electrolyte, a cathode section including a cathode terminal, an electron distributor and a catholyte electrolyte, an ion exchange section between the anode and cathode sections and including an ionolyte electrolyte, ion transfer membranes separating the ionolyte from the anolyte and the catholyte and an electric circuit connected with and between the terminals conducting free electrons from the anode section and delivering free electrons to the cathode section, said ionolyte receives ions of one polarity moving from the anolyte through the membrane related thereto preventing chemical equilibrium in the anode section and sustaining chemical reaction and the generating of free electrons therein, said ions received by the ionolyte from the anolyte release different ions from the ionolyte which move through the membrane between the ionolyte and catholyte and which add to the catholyte.

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

    SciTech Connect

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

    1987-01-01

    For a severe pressurized water reactor accident that leads to a loss of feedwater to the stream generators, such as might occur in a station blackout, fission product decay heating causes a water boil-off. Without effective decay heat removal, the fuel elements will be uncovered. Eventually, steam will oxidize the overheated cladding. The noble gases and volatile fission products, such as cesium and iodine, that are major contributors to the radiological source term will be released from the damaged fuel shortly after cladding failure. The accident environment when these volatile fission products escape was simulated in STEP-3 using four fuel elements from the Belgonucleaire BR3 reactor. The primary objective was to examine the releases in samples collected as close to the test zone as possible. In this paper, an analysis of the temperatures and hydrogen generation is compared with the measurements. The analysis is needed to estimate releases and characterize conditions at the source for studies of fission product transport.

  18. Isotopic Generation and Confirmation of the PWR Application Model 

    SciTech Connect

    L.B. Wimmer

    2003-11-10

    The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section 3.5.2.1.1 of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.

  19. Fuel additives

    SciTech Connect

    Gheysens, J.L.G.

    1990-11-27

    This patent describes a composition for the improvement of hydrocarbon fuels exhibiting a boiling range of gasoline being suitable for use in spark ignition-type engines. It comprises an aromatic amine; a polyaminated detergent; a catalyst comprising a colloidal suspension or amine salt of transition/alkali/alkaline earth metal organic coordinations having at least one metal oxidehydroxide linked to an alkyl chain via a carboxyl group; and a solvent comprising an alkanol-aliphatic ether oxygenated hydrocarbon.

  20. Future Fuels

    DTIC Science & Technology

    2006-04-01

    Storage Devices, Fuel Management, Gasification, Fischer-Tropsch, Syngas , Hubberts’s Peak UNCLAS UNCLAS UNCLAS UU 80 Dr. Sujata Millick (703) 696...traction power – mission payloads – mobile electric power • Improved survivability • Inherent modularity improves maintainability & upgradability ...threatened the output of the Ploesti oil fields and refineries. In the FT process, so-called syngas (a mixture of molecular hydrogen and carbon monoxide

  1. Fuel conditioner

    SciTech Connect

    Nelson, M.L.; Nelson, O.L. Jr.

    1988-06-28

    A fuel conditioner is described comprising 10 to 80% of a polar oxygenated hydrocarbon having an average molecular weight from about 250 to about 500, an acid acid number from about 25 to about 125, and a saponification number from about 30 to about 250; and 5 to 50% of an oxygenated compatibilizing agent having a solubility parameter of from about 8.8 to about 11.5 and moderate to strong hydrogen-bonding capacity.

  2. Alcohol fuels

    SciTech Connect

    Not Available

    1990-07-01

    Ethanol is an alcohol made from grain that can be blended with gasoline to extend petroleum supplies and to increase gasoline octane levels. Congressional proposals to encourage greater use of alternative fuels could increase the demand for ethanol. This report evaluates the growth potential of the ethanol industry to meet future demand increases and the impacts increased production would have on American agriculture and the federal budget. It is found that ethanol production could double or triple in the next eight years, and that American farmers could provide the corn for this production increase. While corn growers would benefit, other agricultural segments would not; soybean producers, for example could suffer for increased corn oil production (an ethanol byproduct) and cattle ranchers would be faced with higher feed costs because of higher corn prices. Poultry farmers might benefit from lower priced feed. Overall, net farm cash income should increase, and consumers would see slightly higher food prices. Federal budget impacts would include a reduction in federal farm program outlays by an annual average of between $930 million (for double current production of ethanol) to $1.421 billion (for triple production) during the eight-year growth period. However, due to an partial tax exemption for ethanol blended fuels, federal fuel tax revenues could decrease by between $442 million and $813 million.

  3. Near-term commercial spent fuel shipping cask requirements

    SciTech Connect

    Daling, P.M.

    1984-11-01

    This report describes an analysis of the near-term commercial light water reactor (LWR) spent fuel transportation system. The objective was to determine if the existing commercial spent fuel shipping cask fleet is adequate to provide the needed transportation services for the period of time the US government would be authorized to accept spent fuel for Federal Interim Storage (FIS). A spent fuel shipping cask supply-demand analysis was performed to evaluate the existing fleet size. The results of the shipping cask handling capability study indicated that by weight, 75% of the spent fuel shipments will be by truck (overweight plus legal-weight truck). From the results of the shipping cask supply-demand analysis it was concluded that, if utilities begin large-scale applications for FIS, the five legal-weight truck (LWT) casks currently in service would be inadequate to perform all of the needed shipments as early as 1987. This further assumes that a western site would be selected for the FIS facility. If the FIS site were to be located in the East, the need for additional LWT casks would be delayed by about two years. The overweight truck (OWT) cask fleet (two PWR and two BWR versions) will be adequate through 1992 if some shipments to FIS can be made several years before a reactor is projected to lose full core reserve. This is because OWT cask requirements increase gradually over the next several years. The feasibility of shipping before losing full core reserve has not been evaluated. Cask utilization requirements in later years will be reduced if some shipments can be made prior to the time they are actually needed. The existing three rail casks are adequate to perform near-term shipments. 18 references, 4 figures, 18 tables.

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

    SciTech Connect

    Wagner, J.C.

    2001-09-28

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

  5. J-2X Fuel Turbopump Point of Departure: The Performance of the J-2s Fuel Turbopump Inducer

    NASA Technical Reports Server (NTRS)

    Sargent, S. R.; Becht, D. G.; Mulder, A. D.

    2008-01-01

    To aid the J-2X program design effort with detailed performance and environment information, the J-2S fuel turbopump (FTP) inducer has undergone a thorough test series in both water and hydrogen. The test series utilizes both inducer only and a complete pump configuration to assess the inducer interaction to the overall turbopump system. The test goals include verification of suction performance against heritage J-2S data, head production, effects of thermodynamic suppression head (TSH), and evaluation of the inducer dynamic pressure caused by cavitation instabilities. Test facilities at both Pratt & Whitney Rocketdyne (PWR) and NASA s Stennis Space Center (SSC) are employed for the testing. The inducer only water test effort conducted at PWR established performance curves for suction performance, head production, and efficiency over a wide operating range. Because the heritage J-2S suction performance data set is in hydrogen, it is desired to obtain current suction performance data in hydrogen as well, thus avoiding the reliance on a theoretical TSH correlation for direct comparison. This data then provides an empirically based TSH correlation allowing for the comparison of water test suction data to system suction requirements. The FTP testing performed at SSC provides these suction performance relationships as well as inlet duct dynamic pressures during liquid hydrogen operation. The test effort successfully confirms the heritage J-2S suction performance and establishes the amount of TSH between water and hydrogen operation at the design flow coefficient. Correlating data is also obtained for cavitating instability frequency content, illustrating the validity of using the wide flow range water test data to predict hydrogen performance.

  6. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    NASA Astrophysics Data System (ADS)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  7. Sub-Scale Testing and Development of the J-2X Fuel Turbopump Inducer

    NASA Technical Reports Server (NTRS)

    Sargent, Scott R.; Becht, David G.

    2011-01-01

    In the early stages of the J-2X upper stage engine program, various inducer configurations proposed for use in the fuel turbopump (FTP) were tested in water. The primary objectives of this test effort were twofold. First, to obtain a more comprehensive data set than that which existed in the Pratt & Whitney Rocketdyne (PWR) historical archives from the original J-2S program, and second, to supplement that data set with information regarding the cavitation induced vibrations for both the historical J-2S configuration as well as those tested for the J-2X program. The J-2X FTP inducer, which actually consists of an inducer stage mechanically attached to a kicker stage, underwent 4 primary iterations utilizing sub-scaled test articles manufactured and tested in PWR's Engineering Development Laboratory (EDL). The kicker remained unchanged throughout the test series. The four inducer configurations tested retained many of the basic design features of the J-2S inducer, but also included variations on leading edge blade thickness and blade angle distribution, primarily aimed at improving suction performance at higher flow coefficients. From these data sets, the effects of the tested design variables on hydrodynamic performance and cavitation instabilities were discerned. A limited comparison of impact to the inducer efficiency was determined as well.

  8. U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

    SciTech Connect

    Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.; Bowman, Stephen M.

    2016-09-01

    Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

  9. Results from long-term dissolution tests using oxidized spent fuel

    SciTech Connect

    Wilson, C.N.

    1990-11-01

    Two semi-static dissolution tests using oxidized PWR spent fuel specimens are being conducted under ambient hot cell conditions in Nevada Test Site J-13 well water and unsealed fused silica vessels. The test specimens were oxidized at 250{degree}C in air to bulk oxygen-to-metal (O/M) values of 2.21 and 2.33. Following an initial 191-day test cycle, the specimens were restarted in fresh J-13 water for a second long-term test cycle. Results through the first 40 months of Cycle 2 are compared with results from similar tests at 25{degree}C and 85{degree}C using unoxidized spent fuel specimens. Increased concentrations of U, Am, Cm and NP were measured in 0.4-{mu}m filtered samples from the oxidized fuel tests compared to the unoxidized fuel tested at 25{degree}C; Pu concentrations were not affected by the fuel oxidation state. Most of the Am and Cm, and a portion of the Pu, measured in 0.4-{mu}m filtered samples was removed by 2-nm filtration. Fission product release results were normalized to specimen inventories and reported as fractional release. No attempt was made to normalize the data to surface area. Initial {sup 99}Tc release was greatly increased, and prolonged increases in the fractional release rates of {sup 99}Tc and {sup 129}I occurred as a result of fuel oxidation. Fractional release rates for {sup 137}Cs and {sup 90}Sr from oxidized fuel eventually decreased to levels similar to those observed with unoxidized fuel after equivalent testing times, suggesting that matrix dissolution rates normalized to fuel mass were not increased as a result of oxidation. 6 refs., 3 figs., 2 tabs.

  10. Neutronic optimization in high conversion Th-{sup 233}U fuel assembly with simulated annealing

    SciTech Connect

    Kotlyar, D.; Shwageraus, E.

    2012-07-01

    This paper reports on fuel design optimization of a PWR operating in a self sustainable Th-{sup 233}U fuel cycle. Monte Carlo simulated annealing method was used in order to identify the fuel assembly configuration with the most attractive breeding performance. In previous studies, it was shown that breeding may be achieved by employing heterogeneous Seed-Blanket fuel geometry. The arrangement of seed and blanket pins within the assemblies may be determined by varying the designed parameters based on basic reactor physics phenomena which affect breeding. However, the amount of free parameters may still prove to be prohibitively large in order to systematically explore the design space for optimal solution. Therefore, the Monte Carlo annealing algorithm for neutronic optimization is applied in order to identify the most favorable design. The objective of simulated annealing optimization is to find a set of design parameters, which maximizes some given performance function (such as relative period of net breeding) under specified constraints (such as fuel cycle length). The first objective of the study was to demonstrate that the simulated annealing optimization algorithm will lead to the same fuel pins arrangement as was obtained in the previous studies which used only basic physics phenomena as guidance for optimization. In the second part of this work, the simulated annealing method was used to optimize fuel pins arrangement in much larger fuel assembly, where the basic physics intuition does not yield clearly optimal configuration. The simulated annealing method was found to be very efficient in selecting the optimal design in both cases. In the future, this method will be used for optimization of fuel assembly design with larger number of free parameters in order to determine the most favorable trade-off between the breeding performance and core average power density. (authors)

  11. State-of-the-art on instant release of fission products from spent nuclear fuel

    SciTech Connect

    Kienzler, Bernhard; Gonzalez-Robles, Ernesto

    2013-07-01

    Within the EURATOM FP7 Collaborative Project 'Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)', a State-of-the-Art Report was prepared. The fast / instant release fraction (IRF) is defined as a fraction of the inventory of radionuclides that may be rapidly released from the fuel and fuel assembly materials at the time of canister breaching. In the context of safety analysis for a repository, the time span for mobilization of this fraction can be considered instantaneously, even if the process takes some time in experiments. Radionuclides contributing to the fast release are fission gases (Kr and Xe), easily soluble elements such as cesium and iodine, and other elements which are hardly incorporated in the UO{sub 2} crystal lattice. The present contribution summarizes the results obtained from published studies focused on rapid release experiments carried out with different spent nuclear fuel (SNF), samples, sizes, techniques (batch and flow-through), and durations. A total of 80 experiments cover the study of UO{sub 2} SNF from pressure water reactors (PWR) of different initial enrichments and burn-up, while 20 experiments were performed with UO{sub 2} SNF from boiling water reactors (BWR) and 8 with MOX fuel. (authors)

  12. Microstructure of irradiated SBR MOX fuel and its relationship to fission gas release

    NASA Astrophysics Data System (ADS)

    Fisher, S. B.; White, R. J.; Cook, P. M. A.; Bremier, S.; Corcoran, R. C.; Stratton, R.; Walker, C. T.; Ivison, P. K.; Palmer, I. D.

    2002-12-01

    SEM and EPMA examinations of the microstructure and microchemistry of British Nuclear Fuel's quasi-homogeneous SBR MOX fuel following irradiation suggests behaviour which is very similar to that observed in UO 2. Most significantly, a fission gas release of 1% in three-cycle SBR MOX PWR rods is associated with the development of a well-defined intergranular bubble network, which has not been seen previously in the more heterogeneous MOX fuels irradiated under similar conditions. The contrast between the observations is attributed to the relatively low volume fraction and small size of the Pu rich inhomogeneities in the SBR fuel which generate only 4% of the total fission gas and eject most of this into the surrounding mixed oxide matrix. The resulting perturbation in the Xe distribution has a negligible influence on the evolution of the microstructure. A key observation is made from the results of recent post-irradiation annealing experiments performed on SBR MOX and UO 2. These confirm near identical fission gas behaviour in the two fuel types when the influence of thermal conductivity and rod rating are removed.

  13. Halden In-Reactor Test to Exhibit PWR Axial Offset Anomaly

    SciTech Connect

    P.Bennett, B. Beverskog, R.Suther

    2004-12-01

    Many PWRs have encountered the axial offset anomaly (AOA) since the early 1990s, and these experiences have been reported widely. AOA is a phenomenon associated with localized boron hideout in corrosion product deposits (crud) on fuel surfaces. Several mitigation approaches have been developed or are underway to either delay the onset of AOA or avoid it entirely. This study describes the first phase of an experimental program designed to investigate whether the use of enriched boric acid (EBA) in the reactor coolant can mitigate AOA.

  14. Best-estimate LOCA radiation signature for equipment qualification. [PWR; BWR

    SciTech Connect

    Lurie, N.A.; Bonzon, L.L.

    1980-01-01

    The radiation aspect of reactor equipment qualification depends on a knowledge of the appropriate source term. An attempt has been made to define a realistic radiation source corresponding to the loss-of-coolant accident. This best-estimate source is based on available fission product release data from damaged fuel during an unterminated LOCA as described in the Reactor Safety Study (WASH-1400). Energy release rates as a function of time have been calculated for both betas and gamma rays. The results are significantly different from the sources specified in Regulatory Guide 1.89. Spectra corresponding to the best-estimate source have also been computed at selected cooling times.

  15. Fuel Burn Estimation Model

    NASA Technical Reports Server (NTRS)

    Chatterji, Gano

    2011-01-01

    Conclusions: Validated the fuel estimation procedure using flight test data. A good fuel model can be created if weight and fuel data are available. Error in assumed takeoff weight results in similar amount of error in the fuel estimate. Fuel estimation error bounds can be determined.

  16. Aviation fuels outlook

    NASA Technical Reports Server (NTRS)

    Momenthy, A. M.

    1980-01-01

    Options for satisfying the future demand for commercial jet fuels are analyzed. It is concluded that the most effective means to this end are to attract more refiners to the jet fuel market and encourage development of processes to convert oil shale and coal to transportation fuels. Furthermore, changing the U.S. refineries fuel specification would not significantly alter jet fuel availability.

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

    NASA Astrophysics Data System (ADS)

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

    2013-11-01

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

  18. 146. FUEL LINE TO SKID 2 (FUEL LOADER) IN FUEL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    146. FUEL LINE TO SKID 2 (FUEL LOADER) IN FUEL CONTROL ROOM (215), LSB (BLDG. 751). LIQUID NITROGEN/HELIUM HEAT EXCHANGER ON RIGHT. - Vandenberg Air Force Base, Space Launch Complex 3, Launch Pad 3 East, Napa & Alden Roads, Lompoc, Santa Barbara County, CA

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

    SciTech Connect

    Morris, Edgar

    2014-10-01

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

  20. Fuel extender

    SciTech Connect

    Dorn, G.K.; Gilbert, H.A.

    1989-02-21

    An efficient and cost competitive fuel extender liquid is described for blending with lead-free gasoline as an additive thereto in a maximum amount of up to about 35% thereof with 65% by volume of the gasoline in a blended mixture wherein. The content of the extender in the resultant fuel as proportioned on the basis of its thus representative maximum content consists essentially of: naphtha X as represented by C/sub 4/, C/sub 5/ and C/sub 6/ hydrocarbons having a Reid vapor pressure of about 8.5 to 9.6 per ASTM, D323 test procedure and an initial distillation point of about 101/sup 0/F. and an end point of about 280/sup 0/F. within a range of about 10 to 25% by volume, about 3.8 to 6.0% by volume of anhydrous ethanol, a stabilizing amount of a water repellent of the class consisting of ethyl acetate and methyl isotubyl ketone; and about 4 to 10.5% by volume of aromatics benzene and toluene, of benzene and xylene or of benzene with toluene and xylene; the extender having a specific gravity substantially comparable with that of the lead-free gasoline to which it is to be added and having phase stability in the presence of water when mixed with the gasoline.

  1. The J-2X Fuel Turbopump - Design, Development, and Test

    NASA Technical Reports Server (NTRS)

    Tellier, James G.; Hawkins, Lakiesha V.; Shinguchi, Brian H.; Marsh, Matthew W.

    2011-01-01

    Pratt and Whitney Rocketdyne (PWR), a NASA subcontractor, is executing the design, development, test, and evaluation (DDT&E) of a liquid oxygen, liquid hydrogen two hundred ninety four thousand pound thrust rocket engine initially intended for the Upper Stage (US) and Earth Departure Stage (EDS) of the Constellation Program Ares-I Crew Launch Vehicle (CLV). A key element of the design approach was to base the new J-2X engine on the heritage J-2S engine with the intent of uprating the engine and incorporating SSME and RS-68 lessons learned. The J-2S engine was a design upgrade of the flight proven J-2 configuration used to put American astronauts on the moon. The J-2S Fuel Turbopump (FTP) was the first Rocketdyne-designed liquid hydrogen centrifugal pump and provided many of the early lessons learned for the Space Shuttle Main Engine High Pressure Fuel Turbopumps. This paper will discuss the design trades and analyses performed for the current J-2X FTP to increase turbine life; increase structural margins, facilitate component fabrication; expedite turbopump assembly; and increase rotordynamic stability margins. Risk mitigation tests including inducer water tests, whirligig turbine blade tests, turbine air rig tests, and workhorse gas generator tests characterized operating environments, drove design modifications, or identified performance impact. Engineering design, fabrication, analysis, and assembly activities support FTP readiness for the first J-2X engine test scheduled for July 2011.

  2. Fuel cells: A survey

    NASA Technical Reports Server (NTRS)

    Crowe, B. J.

    1973-01-01

    A survey of fuel cell technology and applications is presented. The operating principles, performance capabilities, and limitations of fuel cells are discussed. Diagrams of fuel cell construction and operating characteristics are provided. Photographs of typical installations are included.

  3. Renewable Fuel Standard Program

    EPA Pesticide Factsheets

    Information about regulations, developed by EPA, in collaboration with refiners, renewable fuel producers, and many other stakeholders, that ensure that transportation fuel sold in the United States contains a minimum volume of renewable fuel.

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

    SciTech Connect

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

    2002-10-29

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

  5. Future aviation fuels overview

    NASA Technical Reports Server (NTRS)

    Reck, G. M.

    1980-01-01

    The outlook for aviation fuels through the turn of the century is briefly discussed and the general objectives of the NASA Lewis Alternative Aviation Fuels Research Project are outlined. The NASA program involves the evaluation of potential characteristics of future jet aircraft fuels, the determination of the effects of those fuels on engine and fuel system components, and the development of a component technology to use those fuels.

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

    SciTech Connect

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

    1982-01-01

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

  7. Effect of hydrocarbon fuel type on fuel

    NASA Technical Reports Server (NTRS)

    Wong, E. L.; Bittker, D. A.

    1982-01-01

    A modified jet fuel thermal oxidation tester (JFTOT) procedure was used to evaluate deposit and sediment formation for four pure hydrocarbon fuels over the temperature range 150 to 450 C in 316-stainless-steel heater tubes. Fuel types were a normal alkane, an alkene, a naphthene, and an aromatic. Each fuel exhibited certain distinctive deposit and sediment formation characteristics. The effect of aluminum and 316-stainless-steel heater tube surfaces on deposit formation for the fuel n-decane over the same temperature range was investigated. Results showed that an aluminum surface had lower deposit formation rates at all temperatures investigated. By using a modified JFTOT procedure the thermal stability of four pure hydrocarbon fuels and two practical fuels (Jet A and home heating oil no. 2) was rated on the basis of their breakpoint temperatures. Results indicate that this method could be used to rate thermal stability for a series of fuels.

  8. Fuel cell technology: A sweeter fuel

    NASA Astrophysics Data System (ADS)

    Kendall, Kevin

    2002-12-01

    Eating sugar gives us a boost when we feel tired because our cells use it as fuel to produce energy. Likewise, sugar can now be used to produce power in artificial biological fuel cells that function in a physiological environment.

  9. Fuel processor for fuel cell power system

    DOEpatents

    Vanderborgh, Nicholas E.; Springer, Thomas E.; Huff, James R.

    1987-01-01

    A catalytic organic fuel processing apparatus, which can be used in a fuel cell power system, contains within a housing a catalyst chamber, a variable speed fan, and a combustion chamber. Vaporized organic fuel is circulated by the fan past the combustion chamber with which it is in indirect heat exchange relationship. The heated vaporized organic fuel enters a catalyst bed where it is converted into a desired product such as hydrogen needed to power the fuel cell. During periods of high demand, air is injected upstream of the combustion chamber and organic fuel injection means to burn with some of the organic fuel on the outside of the combustion chamber, and thus be in direct heat exchange relation with the organic fuel going into the catalyst bed.

  10. Fuel Reformer Nozzle Development

    NASA Technical Reports Server (NTRS)

    Lai, Ming-Chia D.

    2003-01-01

    The fellowship work this summer will be in support of the development of a fuel mixer for a liquid fuel reformer that is upstream of a fuel cell. Tasks for the summer shall consist of design of a fuel mixer, setup of the laser diagnostics for determining the degree of fuel mixing, and testing of the fuel mixer. The fuel mixer shall be a venturi section with fuel injected at or near the throat, and an air swirler upstream of the venturi. Data to determine the performance of the mixer shall be taken using a Phase Doppler Particle Analyzer (PDPA).

  11. Transmutation Performance Analysis for Inert Matrix Fuels in Light Water Reactors and Computational Neutronics Methods Capabilities at INL

    SciTech Connect

    Michael A. Pope; Samuel E. Bays; S. Piet; R. Ferrer; Mehdi Asgari; Benoit Forget

    2009-05-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF, thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCmCf, PuNpAm, PuNp, and Pu. The results of this infinite lattice assembly-wise study using SCALE 5.1 indicate that it may be feasible to recycle TRU in this manner using an otherwise typical PWR assembly without violating peaking factor limits.

  12. Internal reforming fuel cell assembly with simplified fuel feed

    DOEpatents

    Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

    2001-01-01

    A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

  13. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    SciTech Connect

    Williamson, D.A.

    1991-12-31

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas & Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States` utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste.

  14. On the oxidation state of UO 2 nuclear fuel at a burn-up of around 100 MWd/kgHM

    NASA Astrophysics Data System (ADS)

    Walker, C. T.; Rondinella, V. V.; Papaioannou, D.; Winckel, S. Van; Goll, W.; Manzel, R.

    2005-10-01

    Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102 MWd/kgHM are presented. The local Δ G bar (O2) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991 ± 0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the Δ G bar (O2) of the fuel increased from -370 kJ mol-1 at r/r0 = 0.1 to -293 kJ mol-1 at r/r0 = 0.975. Thus, the Δ G bar (O2) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

  15. COBRA-SFS thermal analysis of a sealed storage cask for the Monitored Retrievable Storage of spent fuel

    SciTech Connect

    Rector, D.R.; Wheeler, C.L.

    1986-01-01

    The COBRA-SFS (Spent Fuel Storage) computer code was used to predict temperature distributions in a concrete Sealed Storage Cask (SSC). This cask was designed for the Department of Energy in the Monitored Retrievable Storage (MRS) program for storage of spent fuel from commercial power operations. Analytical results were obtained for nominal operation of the SSC with spent fuel from 36 PWR fuel assemblies consolidated in 12 cylindrical canisters. Each canister generates 1650 W of thermal power. A parametric study was performed to assess the effects on cask thermal performance of thermal conductivity of the concrete, the fin material, and the amount of radial reinforcing steel bars (rebar). Seven different cases were modeled. The results of the COBRA-SFS analysis of the current cask design predict that the peak fuel cladding temperature in the SSC will not exceed the 37/sup 0/C design limit for the maximum spent fuel load of 19.8 kW and a maximum expected ambient temperature of 37.8/sup 0/C (100/sup 0/F). The results of the parametric analyses illustrate the importance of material selection and design optimization with regard to the SSC thermal performance.

  16. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

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

    NASA Astrophysics Data System (ADS)

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

    2015-05-01

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

  18. Linking Grain Boundary Microstructure to Stress Corrosion Cracking of Cold Rolled Alloy 690 in PWR Primary Water

    SciTech Connect

    Bruemmer, Stephen M.; Olszta, Matthew J.; Toloczko, Mychailo B.; Thomas, Larry E.

    2012-10-01

    Grain boundary microstructures and microchemistries are examined in cold-rolled alloy 690 tubing and plate materials and comparisons are made to intergranular stress corrosion cracking (IGSCC) behavior in PWR primary water. Chromium carbide precipitation is found to be a key aspect for materials in both the mill annealed and thermally treated conditions. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG carbide distribution. For the same degree of cold rolling, alloys with few IG precipitates exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal that cracked carbides appeared to blunt propagation of IGSCC cracks in many cases. Preliminary results suggest that the localized grain boundary strains and stresses produced during cold rolling promote IGSCC susceptibility and not the cracked carbides and voids.

  19. Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

    SciTech Connect

    Hardin, Ernest; Blink, James; Carter, Joe; Fratoni, Massimiliano; Greenberg, Harris; Sutton, Mark; Howard, Robert

    2012-07-01

    A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR

  20. Fuel dissipater for pressurized fuel cell generators

    DOEpatents

    Basel, Richard A.; King, John E.

    2003-11-04

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a pressurized fuel cell generator (10) when the electrical power output of the fuel cell generator is terminated during transient operation, such as a shutdown; where, two electrically resistive elements (two of 28, 53, 54, 55) at least one of which is connected in parallel, in association with contactors (26, 57, 58, 59), a multi-point settable sensor relay (23) and a circuit breaker (24), are automatically connected across the fuel cell generator terminals (21, 22) at two or more contact points, in order to draw current, thereby depleting the fuel inventory in the generator.

  1. Design of a Prototype Differential Die-Away Instrument Proposed for Swedish Spent Nuclear Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Jansson, Peter; Swinhoe, Martyn T.; Goodsell, Alison V.; Tobin, Stephen J.

    2016-06-01

    As part of the United States (US) Department of Energy's Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project, the traditional Differential Die-Away (DDA) method that was originally developed for waste drum assay has been investigated and modified to provide a novel application to characterize or verify spent nuclear fuel (SNF). Following the promising, yet largely theoretical and simulation based, research of physics aspects of the DDA technique applied to SNF assay during the early stages of the NGSI-SF project, the most recent effort has been focused on the practical aspects of developing the first fully functional and deployable DDA prototype instrument for spent fuel. As a result of the collaboration among US research institutions and Sweden, the opportunity to test the newly proposed instrument's performance with commercial grade SNF at the Swedish Interim Storage Facility (Clab) emerged. Therefore the design of this instrument prototype has to accommodate the requirements of the Swedish regulator as well as specific engineering constrains given by the unique industrial environment. Within this paper, we identify key components of the DDA based instrument and we present methodology for evaluation and the results of a selection of the most relevant design parameters in order to optimize the performance for a given application, i.e. test-deployment, including assay of 50 preselected spent nuclear fuel assemblies of both pressurized (PWR) as well as boiling (BWR) water reactor type.

  2. Analysis of the risk of transporting spent nuclear fuel by train

    SciTech Connect

    Elder, H.K.

    1981-09-01

    This report uses risk analyses to analyze the safety of transporting spent nuclear fuel for commercial rail shipping systems. The rail systems analyzed are those expected to be used in the United States when the total electricity-generating capacity by nuclear reactors is 100 GW in the late 1980s. Risk as used in this report is the product of the probability of a release of material to the environment and the consequences resulting from the release. The analysis includes risks in terms of expected fatalities from release of radioactive materials due to transportation accidents involving PWR spent fuel shipped in rail casks. The expected total risk from such shipments is 1.3 x 10/sup -4/ fatalities per year. Risk spectrums are developed for shipments of spent fuel that are 180 days and 4 years out-of-reactor. The risk from transporting spent fuel by train is much less (by 2 to 4 orders of magnitude) than the risk to society from other man-caused events such as dam failure.

  3. 40 CFR 80.8 - Sampling methods for gasoline, diesel fuel, fuel additives, and renewable fuels.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... fuel, fuel additives, and renewable fuels. 80.8 Section 80.8 Protection of Environment ENVIRONMENTAL... Provisions § 80.8 Sampling methods for gasoline, diesel fuel, fuel additives, and renewable fuels. The..., blendstocks, fuel additives and renewable fuels for purposes of determining compliance with the...

  4. Assessment of TRAC and RELAP5 codes with ORNL POST-CHF tests. [PWR

    SciTech Connect

    Rohatgi, U.S.; Neymotin, L.; Pu, J.

    1984-01-01

    Brookhaven National Laboratory is involved in assessing thermohydraulic models in various advanced codes such as TRAC-PF1 (Version 7), TRAC-BD1 (Version 12) and RELAP5/MOD1/CY=14. These codes have two fluid formulations and detailed descriptions of wall heat transfer regimes. These wall heat transfer models and correlations were developed using results from separate effect tests for specific fluid conditions and should be assessed for conditions which may exist in the reactor at normal and abnormal operation. In this paper the effort is concentrated on evaluating the capabilities of these codes in predicting the critical heat flux situation in the rod bundle geometry. The tests selected for this purpose are Oak Ridge Post-CHF tests. Oak Ridge Post-CHF tests consist of a series of high pressure and high temperature steady-state experiments and were conducted with water flowing upward through an 8 x 8 rod bundle with rod diameter and rod pitch typical of PWRs with 17 x 17 fuel assemblies.

  5. A Monte Carlo based spent fuel analysis safeguards strategy assessment

    SciTech Connect

    Fensin, Michael L; Tobin, Stephen J; Swinhoe, Martyn T; Menlove, Howard O; Sandoval, Nathan P

    2009-01-01

    assessment process, the techniques employed to automate the coupled facets of the assessment process, and the standard burnup/enrichment/cooling time dependent spent fuel assembly library. We also clearly define the diversion scenarios that will be analyzed during the standardized assessments. Though this study is currently limited to generic PWR assemblies, it is expected that the results of the assessment will yield an adequate spent fuel analysis strategy knowledge that will help the down-select process for other reactor types.

  6. FUEL ROD CLUSTERS

    DOEpatents

    Schultz, A.B.

    1959-08-01

    A cluster of nuclear fuel rods and a tubular casing therefor through which a coolant flows in heat-exchange contact with the fuel rods is described. The fuel rcds are held in the casing by virtue of the compressive force exerted between longitudinal ribs of the fuel rcds and internal ribs of the casing or the internal surfaces thereof.

  7. Concentric layer ramjet fuel

    SciTech Connect

    Burdette, G.W.; Francis, J.P.

    1988-03-08

    This patent describes a solid fuel ramjet grain comprising concentric layers of solid ramjet fuel having a perforation therethrough along the center axis of the grain. The performation is connected to a combustion after-chamber. The solid ramjet fuel layers comprises a pure hydroxyl-terminated polybutadiene hydrocarbon fuel or a mixture of a hydroxyl-terminated polybutadiene hydrocarbon fuel and from about 5 to about 60 percent by weight of an additive to increase the fuel regression rate selected from the group consisting of magnesium, boron carbide, aluminum, and zirconium such that, when buried in the operation of the ramjet, each fuel layer produces a different level of thrust.

  8. Fuel transfer system

    DOEpatents

    Townsend, H.E.; Barbanti, G.

    1994-03-01

    A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool. 6 figures.

  9. Fuel transfer system

    DOEpatents

    Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.

  10. Fuel cells seminar

    SciTech Connect

    1996-12-01

    This year`s meeting highlights the fact that fuel cells for both stationary and transportation applications have reached the dawn of commercialization. Sales of stationary fuel cells have grown steadily over the past 2 years. Phosphoric acid fuel cell buses have been demonstrated in urban areas. Proton-exchange membrane fuel cells are on the verge of revolutionizing the transportation industry. These activities and many more are discussed during this seminar, which provides a forum for people from the international fuel cell community engaged in a wide spectrum of fuel cell activities. Discussions addressing R&D of fuel cell technologies, manufacturing and marketing of fuel cells, and experiences of fuel cell users took place through oral and poster presentations. For the first time, the seminar included commercial exhibits, further evidence that commercial fuel cell technology has arrived. A total of 205 papers is included in this volume.

  11. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations; Yucca Mountain Site Characterization Project

    SciTech Connect

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

    1991-12-01

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 {times} 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990.

  12. Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel

    SciTech Connect

    Gauld, I. C.; Ryman, J. C.

    2000-12-11

    This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance. The importance is investigated as a function of increasing burnup to assist in identifying the key changes in spent fuel characteristics between conventional- and extended-burnup regimes. Studies involving both pressurized water-reactor (PWR) fuel assemblies and boiling-water-reactor (BWR) assemblies are included. This study is seen to be a necessary first step in identifying the high-burnup spent fuel characteristics that may adversely affect the accuracy of current computational methods and data, assess the potential impact on previous guidance on isotopic source terms and decay-heat values, and thus help identify areas for methods and data improvement. Finally, several recommendations on the direction of possible future code validation efforts for high-burnup spent fuel predictions are presented.

  13. Application of surface-harmonics code SUHAM-U and Monte-Carlo code UNK-MC for calculations of 2D light water benchmark-experiment VENUS-2 with UO{sub 2} and MOX fuel

    SciTech Connect

    Boyarinov, V. F.; Davidenko, V. D.; Nevinitsa, V. A.; Tsibulsky, V. F.

    2006-07-01

    Verification of the SUHAM-U code has been carried out by the calculation of two-dimensional benchmark-experiment on critical light-water facility VENUS-2. Comparisons with experimental data and calculations by Monte-Carlo code UNK with the same nuclear data library B645 for basic isotopes have been fulfilled. Calculations of two-dimensional facility were carried out with using experimentally measured buckling values. Possibility of SUHAM code application for computations of PWR reactor with uranium and MOX fuel has been demonstrated. (authors)

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

    NASA Astrophysics Data System (ADS)

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

    2015-06-01

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

  15. Development of a water boil-off spent-fuel calorimeter system. [To measure decay heat generation rate

    SciTech Connect

    Creer, J.M.; Shupe, J.W. Jr.

    1981-05-01

    A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW.

  16. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  17. Nonproliferation and safeguards aspects of fuel cycle programs in reduction of excess separated plutonium and high-enriched uranium

    SciTech Connect

    Persiani, P.J.

    1995-06-01

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. Reference annual mass flows and inventories for a representative 1,400 Mwe Pressurized Water Reactor (PWR) fuel cycle have been investigated for three cases: the 100 percent uranium oxide UO{sub 2} fuel loading once through cycle, and the 33 percent mixed oxide MOX loading configuration for a first and second plutonium recycle. The analysis addresses fuel cycle developments; plutonium and uranium inventory and flow balances; nuclear fuel processing operations; UO{sub 2} once-through and MOX first and second recycles; and the economic incentives to draw-down the excess separated plutonium stores. The preliminary analysis explores several options in reducing the excess separated plutonium arisings and HEU, and the consequences of the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials on nonproliferation and safeguards policy assessments.

  18. Alternative aircraft fuels

    NASA Technical Reports Server (NTRS)

    Longwell, J. P.; Grobman, J.

    1978-01-01

    In connection with the anticipated impossibility to provide on a long-term basis liquid fuels derived from petroleum, an investigation has been conducted with the objective to assess the suitability of jet fuels made from oil shale and coal and to develop a data base which will allow optimization of future fuel characteristics, taking energy efficiency of manufacture and the tradeoffs in aircraft and engine design into account. The properties of future aviation fuels are examined and proposed solutions to problems of alternative fuels are discussed. Attention is given to the refining of jet fuel to current specifications, the control of fuel thermal stability, and combustor technology for use of broad specification fuels. The first solution is to continue to develop the necessary technology at the refinery to produce specification jet fuels regardless of the crude source.

  19. Fuel injector system

    DOEpatents

    Hsu, Bertrand D.; Leonard, Gary L.

    1988-01-01

    A fuel injection system particularly adapted for injecting coal slurry fuels at high pressures includes an accumulator-type fuel injector which utilizes high-pressure pilot fuel as a purging fluid to prevent hard particles in the fuel from impeding the opening and closing movement of a needle valve, and as a hydraulic medium to hold the needle valve in its closed position. A fluid passage in the injector delivers an appropriately small amount of the ignition-aiding pilot fuel to an appropriate region of a chamber in the injector's nozzle so that at the beginning of each injection interval the first stratum of fuel to be discharged consists essentially of pilot fuel and thereafter mostly slurry fuel is injected.

  20. Dual Tank Fuel System

    DOEpatents

    Wagner, Richard William; Burkhard, James Frank; Dauer, Kenneth John

    1999-11-16

    A dual tank fuel system has primary and secondary fuel tanks, with the primary tank including a filler pipe to receive fuel and a discharge line to deliver fuel to an engine, and with a balance pipe interconnecting the primary tank and the secondary tank. The balance pipe opens close to the bottom of each tank to direct fuel from the primary tank to the secondary tank as the primary tank is filled, and to direct fuel from the secondary tank to the primary tank as fuel is discharged from the primary tank through the discharge line. A vent line has branches connected to each tank to direct fuel vapor from the tanks as the tanks are filled, and to admit air to the tanks as fuel is delivered to the engine.

  1. DIESEL FUEL LUBRICATION

    SciTech Connect

    Qu, Jun

    2012-01-01

    The diesel fuel injector and pump systems contain many sliding interfaces that rely for lubrication upon the fuels. The combination of the poor fuel lubricity and extremely tight geometric clearance between the plunger and bore makes the diesel fuel injector vulnerable to scuffing damage that severely limits the engine life. In order to meet the upcoming stricter diesel emission regulations and higher engine efficiency requirements, further fuel refinements that will result in even lower fuel lubricity due to the removal of essential lubricating compounds, more stringent operation conditions, and tighter geometric clearances are needed. These are expected to increase the scuffing and wear vulnerability of the diesel fuel injection and pump systems. In this chapter, two approaches are discussed to address this issue: (1) increasing fuel lubricity by introducing effective lubricity additives or alternative fuels, such as biodiesel, and (2) improving the fuel injector scuffing-resistance by using advanced materials and/or surface engineering processes. The developing status of the fuel modification approach is reviewed to cover topics including fuel lubricity origins, lubricity improvers, alternative fuels, and standard fuel lubricity tests. The discussion of the materials approach is focused on the methodology development for detection of the onset of scuffing and evaluation of the material scuffing characteristics.

  2. Fuel economy of hybrid fuel cell vehicles.

    SciTech Connect

    Ahluwalia, R.; Wang, X.; Rousseau, A.; Nuclear Engineering Division

    2004-01-01

    The potential improvement in fuel economy of a mid-size fuel-cell vehicle by combining it with an energy storage system has been assessed. An energy management strategy is developed and used to operate the direct hydrogen, pressurized fuel-cell system in a load-following mode and the energy storage system in a charge-sustaining mode. The strategy places highest priority on maintaining the energy storage system in a state where it can supply unanticipated boost power when the fuel-cell system alone cannot meet the power demand. It is found that downsizing a fuel-cell system decreases its efficiency on a drive cycle which is compensated by partial regenerative capture of braking energy. On a highway cycle with limited braking energy the increase in fuel economy with hybridization is small but on the stop-and-go urban cycle the fuel economy can improve by 27%. On the combined highway and urban drive cycles the fuel economy of the fuel-cell vehicle is estimated to increase by up to 15% by hybridizing it with an energy storage system.

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

    NASA Astrophysics Data System (ADS)

    Porter, Ian Edward

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

  4. Fuel injection nozzle

    SciTech Connect

    Kato, M.; Tojo, S.; Arai, K.

    1986-07-22

    A fuel injection nozzle is described connected to a fuel injection pump to inject fuel into a combustion chamber of an internal combustion engine consisting of: a nozzle housing defining therein a fuel sump chamber, an injection hole communicating with the sump chamber and opened at the outer surface of the nozzle housing, a stepped cylinder bore having a smaller diameter bore section and a larger diameter bore section and a fuel passage communicating at one end with the sump chamber and at the other end with the smaller diameter bore section of the stepped cylinder bore; a stepped plunger fitted in the stepped cylinder bore and having a smaller diameter plunger section fitted into the smaller diameter bore section and a larger diameter plunger section fitted into the larger diameter bore section in which the smaller diameter bore section together with the end face of the smaller diameter plunger section defines a pump chamber communicating with the fuel passage and the larger diameter bore section together with the end face of the larger diameter plunger section defines a main fuel chamber into which a main fuel is supplied from the fuel injection pump; auxiliary fuel supply means for supplying an auxiliary fuel into the sump chamber and pump chamber through the fuel passage; valve means for opening and closing an injection hole; communication means for permitting the main fuel chamber to communicate with the fuel passage when the main fuel is supplied from the injection pump into the main fuel chamber to cause the stepped plunger to be moved a predetermined distance in a direction in which the auxiliary fuel in the pump chamber is pressurized.

  5. Fuel injection system

    SciTech Connect

    Miyaki, M.; Iwanaga, T.; Fujisawa, H.

    1988-10-18

    This patent describes a fuel injection system for a diesel engine comprising: condition detection means for detecting operating conditions of the diesel engine including a rotational position thereof; low pressure fuel supply means for supplying fuel at a low pressure at an output port thereof; high pressure fuel pump means, having a pump chamber communicating with the output port of the low pressure fuel supply means, and plunger means reciprocable within the pump chamber for introducing fuel from the output port of the low pressure fuel supply means into the pump chamber during a movement in a predetermined direction of the plunger means and for pressurizing the introduced fuel during a movement of the plunger means in the opposite direction so that pressurized fuel is delivered from the high pressure fuel pump means at an output port thereof; common rail fuel storage means, connected to the output port of the high pressure fuel pump means, for storing pressurized fuel delivered from the high pressure pump therein at a substantially continuous pressure.

  6. Fuel injection pump

    SciTech Connect

    Miyaki, M.

    1986-01-07

    This patent describes a fuel injection pump for delivering fuel to the cylinders of an internal combustion engine consisting of: a pump housing with a fuel chamber therein to which fuel is supplied from a fuel tank; means for compressing fuel in the pump chamber and delivering the compressed fuel to the engine cylinders with such means including a pump plunger adapted to be reciprocated so as to introduce fuel into the pump chamber and to pressurize the introduced fuel; spill means for spilling to a low-pressure side on a fuel tank side the compressed fuel which was pressurized in the pump chamber to be delivered from the pump chamber to the engine cylinders, the spill mechanism including a spill passage communicating with the pump chamber and including a solenoid valve located in the spill passage for opening and closing the spill passage with predetermined timing; escape for allowing the compressed fuel pressurized in the pump chamber to escape to the low-pressure side of the fuel tank side.

  7. Fuel cells and fuel cell catalysts

    DOEpatents

    Masel, Richard I.; Rice, Cynthia A.; Waszczuk, Piotr; Wieckowski, Andrzej

    2006-11-07

    A direct organic fuel cell includes a formic acid fuel solution having between about 10% and about 95% formic acid. The formic acid is oxidized at an anode. The anode may include a Pt/Pd catalyst that promotes the direct oxidation of the formic acid via a direct reaction path that does not include formation of a CO intermediate.

  8. Microscale Fuel Cells

    SciTech Connect

    Holladay, Jamie D.; Viswanathan, Vish V.

    2005-11-03

    Perhaprs some of the most innovative work on fuel cells has been the research dedicated to applying silicon fabrication techniques to fuel cells technology creating low power microscale fuel cells applicable to microelectro mechanical systems (MEMS), microsensors, cell phones, PDA’s, and other low power (0.001 to 5 We) applications. In this small power range, fuel cells offer the decoupling of the energy converter from the energy storage which may enable longer operating times and instant or near instant charging. To date, most of the microscale fuel cells being developed have been based on proton exchange membrane fuel cell technology (PEMFC) or direct methanol fuel cell (DMFC) technology. This section will discuss requirements and considerations that need to be addressed in the development of microscale fuel cells, as well as some proposed designs and fabrication strategies.

  9. FUEL CELL ELECTRODE MATERIALS

    DTIC Science & Technology

    FUEL CELL ELECTRODE MATERIALS. RAW MATERIAL SELECTION INFLUENCES POLARIZATION BUT IS NOT A SINGLE CONTROLLING FACTOR. AVAILABLE...DATA INDICATES THAT AN INTERRELATIONSHIP OF POROSITY, AVERAGE PORE VOLUME, AND PERMEABILITY CONTRIBUTES TO ELECTRODE FUEL CELL BEHAVIOR.

  10. Integrated fuel management system

    SciTech Connect

    Barbeau, D.E.

    1987-09-29

    An aircraft fuel management system to regulate fuel from an airframe reservoir is described. The system comprises: an aircraft turbine engine having a combustor providing propulsion for the aircraft; a fuel pump receiving fuel from the reservoir and supplying fuel to the turbine engine; a motor controlling the pump so as to provide fuel to the turbine engine; means for sensing at least one engine condition; means responsive to the sensing means for controlling fuel flow to the turbine engine, and wherein the pump and the motor are of the constant speed type and further comprising valve means for controlling the fuel flow rate to the turbine engine and wherein the controlling means modulates the position of the valve means.

  11. Alternative aircraft fuels technology

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1976-01-01

    NASA is studying the characteristics of future aircraft fuels produced from either petroleum or nonpetroleum sources such as oil shale or coal. These future hydrocarbon based fuels may have chemical and physical properties that are different from present aviation turbine fuels. This research is aimed at determining what those characteristics may be, how present aircraft and engine components and materials would be affected by fuel specification changes, and what changes in both aircraft and engine design would be required to utilize these future fuels without sacrificing performance, reliability, or safety. This fuels technology program was organized to include both in-house and contract research on the synthesis and characterization of fuels, component evaluations of combustors, turbines, and fuel systems, and, eventually, full-scale engine demonstrations. A review of the various elements of the program and significant results obtained so far are presented.

  12. Fuel quality combustion analysis

    NASA Technical Reports Server (NTRS)

    Naegeli, D. W.; Moses, C. A.

    1979-01-01

    A high pressure research combustor operating over a wide range of burner inlet conditions was used to determine the effects of fuel molecular structure on soot formation. Six test fuels with equal hydrogen content (12.8%) were blended to stress different molecular components and final boiling points. The fuels containing high concentrations (20%) of polycyclic aromatics and partially saturated polycyclic structures such as tetralin, produced more soot than would be expected from a hydrogen content correlation for typical petroleum based fuels. Fuels containing naphthenes such as decalin agreed with the hydrogen content correlation. The contribution of polycyclic aromatics to soot formation was equivalent to a reduction in fuel hydrogen content of about one percent. The fuel sensitivity to soot formation due to the polycyclic aromatic contribution decreased as burner inlet pressure and fuel/air ratio increased.

  13. Nuclear fuel element

    DOEpatents

    Zocher, Roy W.

    1991-01-01

    A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

  14. Reformulated diesel fuel

    DOEpatents

    McAdams, Hiramie T [Carrollton, IL; Crawford, Robert W [Tucson, AZ; Hadder, Gerald R [Oak Ridge, TN; McNutt, Barry D [Arlington, VA

    2006-03-28

    Reformulated diesel fuels for automotive diesel engines which meet the requirements of ASTM 975-02 and provide significantly reduced emissions of nitrogen oxides (NO.sub.x) and particulate matter (PM) relative to commercially available diesel fuels.

  15. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  16. Evaluation of effective-stress-function algorithm for nuclear fuel simulation

    SciTech Connect

    Kim, H. C.; Yang, Y. S.; Koo, Y. H.

    2013-07-01

    In a pressurized water reactor (PWR), the mechanical integrity of nuclear fuel is the most critical issue as it is an important barrier for fission products released into the environment. The integrity of zirconium cladding that surrounds uranium oxide can be threatened during off-normal operation owing to a pellet-cladding mechanical interaction (PCMI). To analyze the fuel and cladding behavior during off-operation, the fuel performance code should calculate an inelastic analysis in two - or three-dimensional calculations. In this paper, the effective stress function (ESF) algorithm based on a two-dimensional FE module has been implemented to simulate the inelastic behavior of the cladding with stability and accuracy. The ESF algorithm solves the governing equations of the inelastic constitutive behavior by calculating the zero of the appropriate effective-stress-function. To verify the accuracy of the ESF algorithm for an inelastic analysis, a code-to-code benchmark was performed using the commercial FE code, ANSYS 13.0. To demonstrate the stability and convergence of the implemented algorithm, the number of iterations in the ESF algorithm was compared with that in a sequential algorithm in the case of an inelastic problem. Consequently, the evaluation results demonstrate that the implemented ESF algorithm improves the efficiency of the computation without a loss of accuracy for an inelastic analysis. (authors)

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

    DOE PAGES

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; ...

    2015-03-04

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

  18. Fuel collecting and recycling system

    SciTech Connect

    Cole, E.F.

    1980-06-10

    This system serves to collect and recycle fuel leftover in the fuel manifold and fuel distribution system of a gas turbine power plant when it is shutdown and operates in conjunction with the power plant's existing fuel control.

  19. COMPOSITE FUEL ELEMENT

    DOEpatents

    Hurford, W.J.; Gordon, R.B.; Johnson, W.A.

    1962-12-25

    A sandwich-type fuel element for a reactor is described. This fuel element has the shape of an elongated flat plate and includes a filler plate having a plurality of compartments therein in which the fuel material is located. The filler plate is clad on both sides with a thin cladding material which is secured to the filler plate only to completely enclose the fuel material in each compartment. (AEC)

  20. Direct hydrocarbon fuel cells

    DOEpatents

    Barnett, Scott A.; Lai, Tammy; Liu, Jiang

    2010-05-04

    The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

  1. Navy Fuel Specification Standardization

    DTIC Science & Technology

    1992-04-01

    surfaced periodically to convert further to a single-fuel operation, i.e., one fuel for both aircraft and ship propulsion /power systems. This study...lead to the development of a single distillate fuel for ship propulsion , resulting eventually in the MIL-F-16884 Naval Distillate Fuel (NDF) used today...for both aircraft and ship propulsion /power systems. This report summarizes a study to consider this problem in light of current systems and

  2. Fakir fuel pump

    NASA Technical Reports Server (NTRS)

    1922-01-01

    In designing the Fakir fuel pump, the fundamental idea was to obtain a simple and reliable method of conveying the fuel from a low tank to the carburetor, with the avoidance of the faults of all former methods and the simultaneous warming of the fuel by means of the heat of compression generated. The principle of the Fakir fuel pump rests on the well-known principle of the diaphragm pump, which must be suitably adapted to the present purpose.

  3. Miniature ceramic fuel cell

    DOEpatents

    Lessing, Paul A.; Zuppero, Anthony C.

    1997-06-24

    A miniature power source assembly capable of providing portable electricity is provided. A preferred embodiment of the power source assembly employing a fuel tank, fuel pump and control, air pump, heat management system, power chamber, power conditioning and power storage. The power chamber utilizes a ceramic fuel cell to produce the electricity. Incoming hydro carbon fuel is automatically reformed within the power chamber. Electrochemical combustion of hydrogen then produces electricity.

  4. Transportation fuels from wood

    SciTech Connect

    Baker, E.G.; Elliott, D.C.; Stevens, D.J.

    1980-01-01

    The various methods of producing transportation fuels from wood are evaluated in this paper. These methods include direct liquefaction schemes such as hydrolysis/fermentation, pyrolysis, and thermochemical liquefaction. Indirect liquefaction techniques involve gasification followed by liquid fuels synthesis such as methanol synthesis or the Fischer-Tropsch synthesis. The cost of transportation fuels produced by the various methods are compared. In addition, three ongoing programs at Pacific Northwest Laboratory dealing with liquid fuels from wood are described.

  5. Jet fuel instability mechanisms

    NASA Technical Reports Server (NTRS)

    Daniel, S. R.

    1985-01-01

    The mechanisms of the formation of fuel-insoluble deposits were studied in several real fuels and in a model fuel consisting of tetralin in dodecane solution. The influence of addition to the fuels of small concentrations of various compounds on the quantities of deposits formed and on the formation and disappearance of oxygenated species in solution was assessed. The effect of temperature on deposit formation was also investigated over the range of 308-453 K.

  6. FUEL ROD ASSEMBLY

    DOEpatents

    Hutter, E.

    1959-09-01

    A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

  7. Vented nuclear fuel element

    DOEpatents

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  8. Metallic fuel development

    SciTech Connect

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising.

  9. Fuels from Recycling Systems

    ERIC Educational Resources Information Center

    Tillman, David A.

    1975-01-01

    Three systems, operating at sufficient scale, produce fuels that may be alternatives to oil and gas. These three recycling systems are: Black Clawson Fiberclaim, Franklin, Ohio; Union Carbide, South Charleston, West Virginia; and Union Electric, St. Louis, Missouri. These produce a wet fuel, a pyrolytic gas, and a dry fuel, respectively. (BT)

  10. Fireplaces and Fireplace Fuels.

    ERIC Educational Resources Information Center

    Metz, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fireplaces and fuels. Its objective is for the student to be able to discuss the structural design, operation, and efficiency of fireplaces and characteristics of different fireplace fuels. Some topics covered are fuels, elements…

  11. Alternative Fuels Data Center

    SciTech Connect

    2013-06-01

    Fact sheet describes the Alternative Fuels Data Center, which provides information, data, and tools to help fleets and other transportation decision makers find ways to reduce petroleum consumption through the use of alternative and renewable fuels, advanced vehicles, and other fuel-saving measures.

  12. Fuel cell generator

    DOEpatents

    Isenberg, Arnold O.

    1983-01-01

    High temperature solid oxide electrolyte fuel cell generators which allow controlled leakage among plural chambers in a sealed housing. Depleted oxidant and fuel are directly reacted in one chamber to combust remaining fuel and preheat incoming reactants. The cells are preferably electrically arranged in a series-parallel configuration.

  13. Vehicle fuel system

    DOEpatents

    Risse, John T.; Taggart, James C.

    1976-01-01

    A vehicle fuel system comprising a plurality of tanks, each tank having a feed and a return conduit extending into a lower portion thereof, the several feed conduits joined to form one supply conduit feeding fuel to a supply pump and using means, unused fuel being returned via a return conduit which branches off to the several return conduits.

  14. Comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe PWR vessel

    SciTech Connect

    Fabry, A.; Chaouadi, R.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rosinski, S.T.; Carter, R.G.

    1999-10-01

    The sister pressure vessels at the BR3 and Yankee Rowe PWR plants were operated at lower-than-usual temperature ({approx}260 C) and their plates were austenitized at higher-than-usual temperature ({approx}970 C) -- a heat treatment leading to a coarser microstructure than is typical for the fine grain plates considered in development of USNRC Regulatory Guide 1.99. The surveillance programs provided by Westinghouse for the two plants were limited to the same A302-B plate representative of the Rowe vessel upper shell plate; this material displayed outlier behavior characterized by a 41J. Charpy-V Notch shift significantly larger than predicted by Regulatory Guide 1.99. Because lower irradiation temperature and nickel alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements embodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: (1) The accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively; (2) The BR3 surveillance and vessel testing program; this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, and was trepanned in early 1995; (3) The accelerated irradiations in the Belgian test reactor BR2 of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is contended that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the

  15. The OSMOSE program for the qualification of integral cross sections of actinides: Preliminary results in a PWR-UOx spectrum

    SciTech Connect

    Hudelot, J. P.; Antony, M.; Bernard, D.; Fougeras, P.

    2006-07-01

    -worth of the individual samples. The first experimental results were obtained with a very good reproducibility in 2005 and 2006 in the R1-UO{sub 2} core configuration representative of a PWR UOx standard spectrum. The preliminary results of measurements and comparison to calculational models are reported. (authors)

  16. Neutronic fuel element fabrication

    DOEpatents

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure

  17. CHF Performance of Hybrid Mixing Vane Grid for a Nuclear Fuel Bundle

    SciTech Connect

    Shin, Chang-Hwan; Chun, Tae-Hyun; Choo, Yeon-Jun; Moon, Sang-Ki; Chun, Se-Young

    2007-07-01

    Numerous studies have shown that the mixing vanes of the spacer grids in a nuclear fuel rod bundle increase the Critical Heat Flux (CHF) significantly. The amount of the CHF enhancement depends strongly on the design of the mixing vanes such as the vane shape and vane bending angle. Recently a new mixing vane design was developed for an advanced spacer grid. It is called a Hybrid Mixing Vane. The main objective of this work is to evaluate the CHF performance of the hybrid vane grid and to compare it with that of a split vane grid. Three kinds of rod bundles were tested for the above objectives: no mixing vane grids, the hybrid mixing vane grids, and the split mixing vane grids. To measure the CHF data, 5x5 rod bundle experiments were conducted in the FTHEL (Freon Thermal Hydraulic Experiment Loop). Each experiment was performed by maintaining the following system conditions as constant: inlet pressure, inlet temperature, and mass flow rate. The experiments were performed in ranges of the inlet pressure, P{sub in} = 2000{approx}3000 kPa, mass flux, G = 1000{approx}3000 kg/m{sup 2}s, and inlet subcooling, {delta}h{sub in}= 10{approx}55 kJ/kg, which simulates the PWR operating conditions for a water equivalence through a fluid-to-fluid modeling. The CHF performances were compared with the data belonging to a PWR's operating conditions; a pressure of 2000{approx}3000 kPa and a mass flux of 1500{approx}3000 kg/m{sup 2}s. The average of the CHF increase for the hybrid mixing grids for 20 data sets is 18.2% higher than that for the no vane grids. While the average of the CHF increase for the split mixing vane grids for 20 data sets is 14.5% higher than that for the no vane grids. Consequently, the CHF performance of the hybrid mixing vane grid is superior by about 4% to that of the split mixing vane grid near the normal PWR operating conditions even under a longer grid span than usual. (authors)

  18. Fuel characteristics pertinent to the design of aircraft fuel systems

    NASA Technical Reports Server (NTRS)

    Barnett, Henry C; Hibbard, R R

    1953-01-01

    Because of the importance of fuel properties in design of aircraft fuel systems the present report has been prepared to provide information on the characteristics of current jet fuels. In addition to information on fuel properties, discussions are presented on fuel specifications, the variations among fuels supplied under a given specification, fuel composition, and the pertinence of fuel composition and physical properties to fuel system design. In some instances the influence of variables such as pressure and temperature on physical properties is indicated. References are cited to provide fuel system designers with sources of information containing more detail than is practicable in the present report.

  19. Fuel Cell Handbook update

    SciTech Connect

    Owens, W.R.; Hirschenhofer, J.H.; Engleman, R.R. Jr.; Stauffer, D.B.

    1993-11-01

    The objective of this work was to update the 1988 version of DOE`s Fuel Cell Handbook. Significant developments in the various fuel cell technologies required revisions to reflect state-of-the-art configurations and performance. The theoretical presentation was refined in order to make the handbook more useful to both the casual reader and fuel cell or systems analyst. In order to further emphasize the practical application of fuel cell technologies, the system integration information was expanded. In addition, practical elements, such as suggestions and guidelines to approximate fuel cell performance, were provided.

  20. Alternative aircraft fuels

    NASA Technical Reports Server (NTRS)

    Longwell, J. P.; Grobman, J. S.

    1977-01-01

    The efficient utilization of fossil fuels by future jet aircraft may necessitate the broadening of current aviation turbine fuel specifications. The most significant changes in specifications would be an increased aromatics content and a higher final boiling point in order to minimize refinery energy consumption and costs. These changes would increase the freezing point and might lower the thermal stability of the fuel, and could cause increased pollutant emissions, increased combustor liner temperatures, and poorer ignition characteristics. The effects that broadened specification fuels may have on present-day jet aircraft and engine components and the technology required to use fuels with broadened specifications are discussed.

  1. Alternative aviation turbine fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1977-01-01

    The efficient utilization of fossil fuels by future jet aircraft may necessitate the broadening of current aviation turbine fuel specifications. The most significant changes in specifications would be an increased aromatics content and a higher final boiling point in order to minimize refinery energy consumption and costs. These changes would increase the freezing point and might lower the thermal stability of the fuel and could cause increased pollutant emissions, increased smoke and carbon formation, increased combustor liner temperatures, and poorer ignition characteristics. This paper discusses the effects that broadened specification fuels may have on present-day jet aircraft and engine components and the technology required to use fuels with broadened specifications.

  2. A Demonstration of Advanced Safety Analysis Tools and Methods Applied to Large Break LOCA and Fuel Analysis for PWRs

    SciTech Connect

    Szilard, Ronaldo Henriques; Smith, Curtis Lee; Martineau, Richard Charles

    2016-03-01

    The U.S. Nuclear Regulatory Commission (NRC) is currently proposing a rulemaking designated as 10 CFR 50.46c to revise the loss-of-coolant accident (LOCA)/emergency core cooling system acceptance criteria to include the effects of higher burnup on fuel/cladding performance. We propose a demonstration problem of a representative four-loop PWR plant to study the impact of this new rule in the US nuclear fleet. Within the scope of evaluation for the 10 CFR 50.46c rule, aspects of safety, operations, and economics are considered in the industry application demonstration presented in this paper. An advanced safety analysis approach is used, by integrating the probabilistic element with deterministic methods for LOCA analysis, a novel approach to solving these types of multi-physics, multi-scale problems.

  3. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    SciTech Connect

    Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  4. Analyzing Nuclear Fuel Cycles from Isotopic Ratios of Waste Products Applicable to Measurement by Accelerator Mass Spectrometry

    SciTech Connect

    Biegalski, S R; Whitney, S M; Buchholz, B

    2005-08-24

    An extensive study was conducted to determine isotopic ratios of nuclides in spent fuel that may be utilized to reveal historical characteristics of a nuclear reactor cycle. This forensic information is important to determine the origin of unknown nuclear waste. The distribution of isotopes in waste products provides information about a nuclear fuel cycle, even when the isotopes of uranium and plutonium are removed through chemical processing. Several different reactor cycles of the PWR, BWR, CANDU, and LMFBR were simulated for this work with the ORIGEN-ARP and ORIGEN 2.2 codes. The spent fuel nuclide concentrations of these reactors were analyzed to find the most informative isotopic ratios indicative of irradiation cycle length and reactor design. Special focus was given to long-lived and stable fission products that would be present many years after their creation. For such nuclides, mass spectrometry analysis methods often have better detection limits than classic gamma-ray spectroscopy. The isotopic ratios {sup 151}Sm/{sup 146}Sm, {sup 149}Sm/{sup 146}Sm, and {sup 244}Cm/{sup 246}Cm were found to be good indicators of fuel cycle length and are well suited for analysis by accelerator mass spectroscopy.

  5. REVIEW OF RESULTS FOR THE OECD/NEA PHASE VII BENCHMARK: STUDY OF SPENT FUEL COMPOSITIONS FOR LONG TERM DISPOSAL

    SciTech Connect

    Radulescu, Georgeta; Wagner, John C

    2011-01-01

    This paper summarizes the problem specification and compares participants results for the OECD/NEA/WPNCS Expert Group on Burn-up Credit Criticality Safety Phase VII Benchmark Study of Spent Fuel Compositions for Long-Term Disposal. The Phase VII benchmark was developed to study the ability of relevant computer codes and associated nuclear data to predict spent fuel isotopic compositions and corresponding keff values in a cask configuration over the time duration relevant to spent nuclear fuel (SNF) disposal. The benchmark was divided into two sets of calculations: (1) decay calculations out to 1,000,000 years for provided pressurized-water-reactor (PWR) UO2 discharged fuel compositions and (2) burnup credit criticality calculations for a representative cask model at selected time steps. Contributions from 15 organizations and companies in 10 countries were submitted to the Phase VII benchmark exercise. This paper provides a description of the Phase VII benchmark and detailed comparisons of the participants isotopic compositions and keff values that were calculated with a diversity of computer codes and nuclear data sets. Differences observed in the calculated time-dependent nuclide densities are attributed to different decay data or code-specific numerical approximations. The variability of the keff results is consistent with the evaluated uncertainty associated with cross-section data.

  6. Non-intrusive Experimental Study on Nuclear Fuel Assembly Response to Seismic Loads

    NASA Astrophysics Data System (ADS)

    Weichselbaum, Noah A.

    Experimental measurements of nuclear fuel bundle response to seismic loads have primarily been focused on the response of the structure. Forcing methods have included use of shake tables, however, the majority of work has used hydraulic actuators rigidly connected to a single spacer grid to force the fuel bundle. Structural measurements utilize such instruments as linear variable displacement transducers (LVDT) that are mounted on the structure. From these measurements it has been shown that fuel bundles in prototypical conditions, with an axial flow of 6 m/s, behave markedly different from fuel bundles in still water when there is external forcing on the core from an earthquake. It has also been shown that the structure and fluid are fully coupled. Thus more recently attention has been focused on fluid measurements in the bypass region around fuel bundles with external forcing with laser doppler velocimetry (LDV), which is a point wise fluid velocity measurement technique. This work describes a unique facility that has garnered a large experimental database of fully coupled fluid and structure measurements with time resolved particle image velocimetry (PIV) and digital image correlation (DIC) within a full height 6x6 fuel bundle exposed to seismic forcing on a large 6 degree of freedom shake table. A refractive index matched (RIM) vertical liquid tunnel is mounted on the shake table and houses the fuel bundle which is based on the geometry of a prototypical fuel bundle in a pressurized water reactor (PWR). PIV is obtained with high spatial resolution by rigidly mounting all optical equipment to the test section on the shake table, where the laser light is delivered through high power multi-mode step index fiber optics from a high powered Nd:YLF laser located 10 meters away from the test section. High temporal resolution for the PIV measurements is obtained with state of the art high speed CMOS cameras that record straight to hard drive allowing for increased

  7. Fuel nozzle assembly

    DOEpatents

    Johnson, Thomas Edward; Ziminsky, Willy Steve; Lacey, Benjamin Paul; York, William David; Stevenson, Christian Xavier

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  8. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  9. Fuel injection apparatus

    SciTech Connect

    Omori, T.; Sakakibara, S.; Kato, M.; Masuda, A.

    1986-06-10

    A fuel injection apparatus is described for supplying fuel from a fuel tank to at least one combustion chamber of an internal combustion engine, comprising: an injector housing defining therein a pressure cylinder bore, an injection cylinder bore and a timing cylinder bore connected in series with the pressure cylinder bore; a pressure plunger, fitted in the pressure cylinder bore and having one end defining, in the pressure cylinder bore, a pressure pump chamber communicating with the injection cylinder bore and the timing cylinder bore; first fuel supply means for supplying fuel from the fuel tank to the pressure pump chamber; pressurizing means for reciprocating the pressure plunger in the pressure cylinder bore in synchronism with the engine operation so as to pressurize fuel in the pressure pump chamber; an injection plunger, having a diameter smaller than the pressure plunger, fitted in the injection cylinder bore and having two ends defining, in the injection cylinder bore, an injection pump chamber and a first communicating chamber communicating with the pressure pump chamber; second fuel supply means for adjusting the quantity of the fuel from the fuel tank in accordance with engine operation conditions so as to supply the adjusted fuel quantity to the injection pump chamber; injection valve means for injecting the fuel into the fuel pump chamber toward a combustion chamber of the engine when the pressure in the pressure pump chamber is applied to the injection plunger through the first communicating chamber, and the fuel pressure in the injection pump chamber reaches a predetermined injection pressure.

  10. A High Temperature, non-TRISO Fuel and Clad Design with Commercial-Grade Enrichment for the Prismatic Block Very High Temperature Reactor

    SciTech Connect

    James W. Sterbentz

    2005-11-01

    The prismatic block Very High Temperature Reactor (VHTR) is a leading Generation IV reactor concept. This reactor with its relatively low core power density and large graphite mass currently satisfies the fundamental goals of the Generation IV charter. However, modifications can be made to the fuel and clad design, such that (1) VHTR uranium enrichment can be lowered to near commercial-grade pressurized water reactor (PWR) enrichments, (2) fuel burnups are extended, and (3) the thermal safety margin under transient conditions is increased. This paper outlines a possible fuel and clad design concept for use in a VHTR prismatic block core which could lead to substantial improvements in overall VHTR economics and sustainability. The results of depletion calculations here will demonstrate comparable burnup between the new fuel and clad design with only 4-6 wt% enriched uranium and the current higher enriched 10-20 wt% VHTR fuel design. In addition, the new fuel and clad design concept uses high-temperature ceramic fuel and clad materials that have the potential to significantly increase the thermal margin under VHTR transient conditions. The current fuel block design for the VHTR is the hexagonal Fort Saint Vrain (FSV) fuel block with 108 coolant channels, 210 fuel rods, and six burnable poison holes drilled axially in the block. This basic FSV block is also part of the new design concept here. The basic hexagonal block dimensions remain fixed with only the fuel pellet and clad materials and radii changed. Further optimizations of the fuel block are in progress. Currently, the proposed nuclear fuel for the prismatic VHTR is the well-known TRISO-coated particle fuel. The TRISO-coated particle offers a nice spherical, high-integrity pressure vessel containment for the fission gases (SiC layer). However, due to the multiple particle coating layers, the fuel kernel represents only 9.4% of the total particle volume (350 {micro}m kernel diameter particle) and together with

  11. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    SciTech Connect

    Hursin, M.; Perret, G.

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  12. Fuel cell market applications

    SciTech Connect

    Williams, M.C.

    1995-12-31

    This is a review of the US (and international) fuel cell development for the stationary power generation market. Besides DOE, GRI, and EPRI sponsorship, the US fuel cell program has over 40% cost-sharing from the private sector. Support is provided by user groups with over 75 utility and other end-user members. Objectives are to develop and demonstrate cost-effective fuel cell power generation which can initially be commercialized into various market applications using natural gas fuel by the year 2000. Types of fuel cells being developed include PAFC (phosphoric acid), MCFC (molten carbonate), and SOFC (solid oxide); status of each is reported. Potential international applications are reviewed also. Fuel cells are viewed as a force in dispersed power generation, distributed power, cogeneration, and deregulated industry. Specific fuel cell attributes are discussed: Fuel cells promise to be one of the most reliable power sources; they are now being used in critical uninterruptible power systems. They need hydrogen which can be generated internally from natural gas, coal gas, methanol landfill gas, or other fuels containing hydrocarbons. Finally, fuel cell development and market applications in Japan are reviewed briefly.

  13. Diesel fuel injection system

    SciTech Connect

    Schechter, M.M.; Simko, A.O.

    1986-04-22

    A fuel injection pump is described of the multiple plunger spill port type for an automotive type internal combustion engine, the pump including at least four axially spaced engine camshaft driven pump plungers grouped in pairs and sequentially and in succession moved in one direction through a fuel pumping stroke and oppositely through a fuel intake stroke. A fuel pressurization/supply chamber is contiquous to the end of each plunger for pressurization of the fuel therein or supply of fuel thereto from a supply passage upon coordinate movement of the plunger, fill/spill passage means connected to a single fuel return spill port and in parallel flow relationship to each of the plunger bores as a function of the position of the plungers, each plunger having a pair of internal passages connected at all times to its chamber and alternately alignable with the supply or fill/spill passage means as a function of the position of the plunger. A fuel discharge passage is operatively connecting each of the chambers to an individual engine cylinder, a single spill port control valve movable to block or permit the spill of fuel through the spill port to a return line to control the pressurization of fuel in all of the fuel chambers and associated discharge passages, a single solenoid connected to the spill control valve for moving it to block or unblock the spill port, and a single shuttle valve operatively associated with all of the fill/spill passage means and spill port reciprocably movable between positions to sequentially connect the plunger chambers one at a time in succession to the spill port during the pumping pressurization stroke of its plunger for the injection of fuel to an individual cylinder while the other chambers are in various stages of being refilled with fuel and preparing for pressurization upon successive actuation of the plungers by the camshaft.

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

    SciTech Connect

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

    2006-02-28

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

  15. Alternative fuels for road vehicles

    SciTech Connect

    Poulton, M.L.

    1994-12-31

    The finite nature of global fossil fuel resources underscores the need to develop alternative vehicular fuels. Increased use of renewable and alternative fuels can extend fossil fuel supplies and help resolve air pollution problems inherent in automotive use of conventional fuels. Fuel characteristics, safety implications, feedstocks, infrastructure, fuel production costs, emissions performance, required vehicle modifications, and outlook are described for LPG, reformulated gasoline, natural gas, hydrogen, electricity, biofuels, ethanol, and methanol. 26 fig., 288 refs., 29 tabs.

  16. Fuel cell generator with fuel electrodes that control on-cell fuel reformation

    DOEpatents

    Ruka, Roswell J.; Basel, Richard A.; Zhang, Gong

    2011-10-25

    A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

  17. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  18. Spiral cooled fuel nozzle

    DOEpatents

    Fox, Timothy; Schilp, Reinhard

    2012-09-25

    A fuel nozzle for delivery of fuel to a gas turbine engine. The fuel nozzle includes an outer nozzle wall and a center body located centrally within the nozzle wall. A gap is defined between an inner wall surface of the nozzle wall and an outer body surface of the center body for providing fuel flow in a longitudinal direction from an inlet end to an outlet end of the fuel nozzle. A turbulating feature is defined on at least one of the central body and the inner wall for causing at least a portion of the fuel flow in the gap to flow transverse to the longitudinal direction. The gap is effective to provide a substantially uniform temperature distribution along the nozzle wall in the circumferential direction.

  19. Alcohol fuels for aviation

    SciTech Connect

    Schauffler, P.

    1982-06-01

    The ten-fold increase in aviation fuel prices in eight years has caused a reassessment of alcohol fuels. In a recent test, methanol fuel-flow rate was high at takeoff, and levelled off at 10,000 feet, but above 18,000 fell 30% below avgas use. Because methanol sells thirty cents below avgas per gallon it is already an attractive fuel for piston-engine aircraft. But as 95% of aviation fuel is burned as kerosene in turbines a test program has been set up to look at the performance of small shaft turbine engines with various combinations of alcohols and water, and of straight methanol, and to look at major thrust engine at optimum fuel as well. These tests should determine the overall alcohol potentials for aviation. The tests will also tell if the breakthrough will be modest or major.

  20. Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRS

    SciTech Connect

    Hyun-Jong Joe; Jones, Barclay G.

    2002-07-01

    Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H{sub 2} and O{sub 2}, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H{sub 3}BO{sub 3}), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities. (authors)