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Sample records for cask thermal evaluation

  1. Thermal evaluation of alternative shipping cask for irradiated experiments

    SciTech Connect

    Guillen, Donna Post

    2015-06-01

    Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and post-irradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. Furthermore, from a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.

  2. THERMAL EVALUATION OF ALTERNATE SHIPPING CASK FOR GTRI EXPERIMENTS

    SciTech Connect

    Donna Post Guillen

    2014-06-01

    The Global Threat Reduction Initiative (GTRI) has many experiments yet to be irradiated in support of the High Performance Research Reactor fuels development program. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for post irradiation examination. To date, the General Electric (GE)-2000 cask has been used to transport GTRI experiments between these facilities. However, the availability of the GE-2000 cask to support future GTRI experiments is at risk. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger GTRI experiments. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled experiments. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask.

  3. Thermal evaluation of alternative shipping cask for irradiated experiments

    DOE PAGESBeta

    Guillen, Donna Post

    2015-06-01

    Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and post-irradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavitymore » of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. Furthermore, from a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.« less

  4. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D. R.; Cuta, J. M.; Enderlin, C. W.

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  5. Nondestructive Evaluation of the VSC-17 Cask

    SciTech Connect

    Sheryl Morton; Al Carlson; Cecilia Hoffman; James Rivera; Phil Winston; Koji Shirai; Shin Takahashi; Masaharo Tanaka

    2006-01-01

    In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC 17) spent nuclear fuel storage cask, originally located at the INL Test Area North, as a candidate to study cask performance because it had been used to store fuel as part of a dry cask storage demonstration project for over 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. The INL team met with the CRIEPI representatives in December of 2004 to discuss the next steps. As a result of that meeting, CRIEPI requested that in the summer 2005 INL perform additional surveys on the VSC 17 cask with participation of CRIEPI scientists. This document summarizes the evaluation methods used on the VSC 17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.

  6. 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.

  7. Thermal Hydraulic Analysis of Spent Fuel Casks

    Energy Science and Technology Software Center (ESTSC)

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  8. THERMAL MODELING ANALYSIS OF SRS 70 TON CASK

    SciTech Connect

    Lee, S.; Jordan, J.; Hensel, S.

    2011-03-08

    The primary objective of this work was to perform the thermal calculations to evaluate the Material Test Reactor (MTR) fuel assembly temperatures inside the SRS 70-Ton Cask loaded with various bundle powers. MTR fuel consists of HFBR, MURR, MIT, and NIST. The MURR fuel was used to develop a bounding case since it is the fuel with the highest heat load. The results will be provided for technical input for the SRS 70 Ton Cask Onsite Safety Assessment. The calculation results show that for the SRS 70 ton dry cask with 2750 watts total heat source with a maximum bundle heat of 670 watts and 9 bundles of MURR bounding fuel, the highest fuel assembly temperatures are below about 263 C. Maximum top surface temperature of the plastic cover is about 112 C, much lower than its melting temperature 260 C. For 12 bundles of MURR bounding fuel with 2750 watts total heat and a maximum fuel bundle of 482 watts, the highest fuel assembly temperatures are bounded by the 9 bundle case. The component temperatures of the cask were calculated by a three-dimensional computational fluid dynamics approach. The modeling calculations were performed by considering daily-averaged solar heat flux.

  9. Safety evaluation for packaging (onsite) SERF cask

    SciTech Connect

    Edwards, W.S.

    1997-10-24

    This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  10. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    SciTech Connect

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  11. Evaluation of improvement potential for spent fuel cask handling

    SciTech Connect

    Franklin, A.L.

    1981-02-01

    This report describes the quantitative analysis of opportunities to improve the loading/unloading operations for spent fuel shipping casks. The improvement potential is defined as a reduction in the time for completion or worker exposure for the complete handling operations. Two casks have been chosen as representative of presently available shipping casks. These are the NAC-1/NFS-4 legal weight truck cask and the IF-300 rail cask. The handling operations for each of these casks are broken down into a series of sequential steps. The time for completion and worker exposure is described by a probability density function for each step. These step descriptions are then combined to form a base case description of the total loading/unloading operation. Potential improvement opportunities are evaluated by modifying the appropriate probability density function descriptors then recombining the steps to form a probabilistic description of the modified operation.

  12. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    SciTech Connect

    THIELGES, J.R.; CHASTAIN, S.A.

    2007-06-21

    The Disposable Solid Waste Cask (DSWC) is a shielded cask capable of transporting, storing, and disposing of six non-fuel core components or approximately 27 cubic feet of radioactive solid waste. Five existing DSWCs are candidates for use in storing and disposing of non-fuel core components and radioactive solid waste from the Interim Examination and Maintenance Cell, ultimately shipping them to the 200 West Area disposal site for burial. A series of inspections, studies, analyses, and modifications were performed to ensure that these casks can be used to safely ship solid waste. These inspections, studies, analyses, and modifications are summarized and attached in this report. Visual inspection of the casks interiors provided information with respect to condition of the casks inner liners. Because water was allowed to enter the casks for varying lengths of time, condition of the cask liner pipe to bottom plate weld was of concern. Based on the visual inspection and a corrosion study, it was concluded that four of the five casks can be used from a corrosion standpoint. Only DSWC S/N-004 would need additional inspection and analysis to determine its usefulness. The five remaining DSWCs underwent some modification to prepare them for use. The existing cask lifting inserts were found to be corroded and deemed unusable. New lifting anchor bolts were installed to replace the existing anchors. Alternate lift lugs were fabricated for use with the new lifting anchor bolts. The cask tiedown frame was modified to facilitate adjustment of the cask tiedowns. As a result of the above mentioned inspections, studies, analysis, and modifications, four of the five existing casks can be used to store and transport waste from the Interim Examination and Maintenance Cell to the disposal site for burial. The fifth cask, DSWC S/N-004, would require further inspections before it could be used.

  13. PRELIMINARY REPORT: EFFECTS OF IRRADIATION AND THERMAL EXPOSURE ON ELASTOMERIC SEALS FOR CASK TRANSPORTATION AND STORAGE

    SciTech Connect

    Verst, C.; Skidmore, E.; Daugherty, W.

    2014-05-30

    A testing and analysis approach to predict the sealing behavior of elastomeric seal materials in dry storage casks and evaluate their ability to maintain a seal under thermal and radiation exposure conditions of extended storage and beyond was developed, and initial tests have been conducted. The initial tests evaluate the aging response of EPDM elastomer O-ring seals. The thermal and radiation exposure conditions of the CASTOR® V/21 casks were selected for testing as this cask design is of interest due to its widespread use, and close proximity of the seals to the fuel compared to other cask designs leading to a relatively high temperature and dose under storage conditions. A novel test fixture was developed to enable compression stress relaxation measurements for the seal material at the thermal and radiation exposure conditions. A loss of compression stress of 90% is suggested as the threshold at which sealing ability of an elastomeric seal would be lost. Previous studies have shown this value to be conservative to actual leakage failure for most aging conditions. These initial results indicate that the seal would be expected to retain sealing ability throughout extended storage at the cask design conditions, though longer exposure times are needed to validate this assumption. The high constant dose rate used in the testing is not prototypic of the decreasingly low dose rate that would occur under extended storage. The primary degradation mechanism of oxidation of polymeric compounds is highly dependent on temperature and time of exposure, and with radiation expected to exacerbate the oxidation.

  14. Evaluation of gamma radiation shielding for nuclear waste shipping casks

    SciTech Connect

    Liu, Y.Y.; Carlson, R.D.; Primeau, S.J.; Wangler, M.E.

    1998-05-01

    A method has been developed for evaluating gamma radiation shielding of shipping casks that are used to transport nuclear waste with ill-defined radionuclide contents. The method is based on calculations that establish individual limits for a comprehensive list of radionuclides in the waste, assuming that each radionuclide is uniformly distributed in a volumetric source in the cask. For multiple radionuclide mixtures, a linear fraction rule is used to restrict the total amount of radionuclides such that the sum of the fractions does not exceed 1. As long as the radionuclide limits and the linear fraction rule are followed, it can be shown that the regulatory dose rate requirements for a cask will be satisfied under normal conditions of transport and in a hypothetical accident during which the shielding thickness of the cask has been reduced by 40%.

  15. CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

    SciTech Connect

    Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

    2005-11-07

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L{sub R}, for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A{sub 2}, are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate.

  16. Safety evaluation for packaging (onsite) disposable solid waste cask

    SciTech Connect

    Flanagan, B.D., Westinghouse Hanford

    1996-12-20

    This safety evaluation for packaging (SEP) evaluates and documents the ability of the Disposable Solid Waste Cask (DSWC) to meet the packaging requirements of HNF-CM-2-14, Hazardous Material Packaging and Shipping, for the onsite transfer of special form, highway route controlled quantity, Type B fissile radioactive material. This SEP evaluates five shipments of DSWCs used for the transport and storage of Fast Flux Test Facility unirradiated fuel to the Plutonium Finishing Plant Protected Area.

  17. Evaluation of three partially volatile neutron shields for high-performance shipping casks. [Thermal stability, mechanical properties of borated silicone rubbers and wood

    SciTech Connect

    Rack, H.J.; Pearson, H.S.

    1981-02-01

    The thermal stability and mechanical behavior of three partially volatile candidate neutron shield materials have been evaluated. The results indicate that silicone based rubbers, impregnated with elemental boron or boron carbide, Boro-silicone 236 and Bisco NS-I respectively are more thermally stable than are borated beechwoods, e.g., Permali JN. Mechanical property measurements indicated however that the compressive strength of the borated beechwood is 10 to 48 times higher than that of the silicone-based rubbers. The compressive strengths of the borated beechwood and boron carbide impregnated silicone rubber were substantially more sensitive to test temperature than was the compressive strength of the boron impregnated silicone rubber. Finally the compressive strengths and energy absorbing capability of the boron impregnated silicone rubber is not affected by prior thermal exposure at 425/sup 0/K for 1000h.

  18. STACE: Source Term Analyses for Containment Evaluations of transport casks

    SciTech Connect

    Seager, K. D.; Gianoulakis, S. E.; Barrett, P. R.; Rashid, Y. R.; Reardon, P. C.

    1992-01-01

    Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source term has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volatile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking (e.g., the quantity and size distribution of fuel rod breaches) in which experimental validation is planned. The CRUD spallation fraction is the major area where no quantitative data has been found; therefore, this also requires experimental validation. In the interim, STACE conservatively assumes a 100% spallation fraction for computing the releasable activity. The source term methodology also conservatively assumes that there is 1 Ci of residual contamination available for release in the transport cask. However, residual contamination is still by far the smallest contributor to the source term activity.

  19. STABILITY EVALUATION OF METAL CASK ATTACHED TO A TRANSFER PALLET DURING LONG-PERIOD SEISMIC MOTIONS

    NASA Astrophysics Data System (ADS)

    Kawaguchi, Shohei; Shirai, Koji; Kanazawa, Kenji

    Rocking behavior of unfixed body is affected by center of mass, material coefficient of restitution and so on. 2/5 scale metal cask model considering these parameter was used for seismic test to evaluate stability of grounding metal cask attached to a transfer pallet under the influence of long-period earthquake motion. The newest knowledge from seismic test indicates seismic motion with high velocity over 100 kine not always cause the raise of response velocity of metal cask because of energy consumption by cask sliding and impact deformation of concrete. And new estimation method (called "Window energy spectrum method") of earthquake response spectrum gives suitable evaluation of response energy.

  20. Nuclear Criticality Safety Evaluation of the 9965, 9968, 9972, 9973, 9974, and 9975 Shipping Casks

    SciTech Connect

    Frost, R.L.

    1999-02-26

    A Nuclear Criticality Safety Evaluation (NCSE) has been performed for the 9965, 9968, 9972, 9973, 9974, and 9975 SRS-designed shipping casks. This was done in support of the recertification effort for the 9965 and 9968, and the certification of the newly designed 9972-9975 series. The analysis supports the use of these packages as Fissile Class I for shipment of fissionable material from the SRS FB-Line, HB-Line, and from Lawrence Livermore national Laboratory. six different types of material were analyzed with varying Isotopic composition, of both oxide and metallic form. The mass limits required to support the fissile Class I rating for each of the envelopes are given in the Table below. These mass limits apply if DOE approves an exception as described in 10 CFR 71.55(c), such that water leakage into the primary containment vessel does not need to be considered in the criticality analysis. If this exception is not granted, the mass limits are lower than those shown below. this issue is discussed in detail in sections 5 and 6 of the report.One finding from this work is important enough to highlight in the abstract. The fire tests performed for this family of shipping casks indicates only minimal charring of the Celotex thermal insulation. Analysis of the casks with no Celotex insulation (assuming it has all burned away), results in values of k-eff that exceed 1.0. Therefore, the Celotex insulation must remain intact in order to guarantee sub criticality of the 9972-9975 family of shipping casks.

  1. COBRA-SFS CYCLE 3. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  2. Evaluation of Cask Drop Criticality Issues at K Basin

    SciTech Connect

    GOLDMANN, L.H.

    2000-01-24

    An analysis of ability of Multi-canister Overpack (MCO) to withstand drops at K Basin without exceeding the criticality design requirements. Report concludes the MCO will function acceptably. The spent fuel currently residing in the 105 KE and 105 KW storage basins will be placed in fuel storage baskets which will be loaded into the MCO cask assembly. During the basket loading operations the MCO cask assembly will be positioned near the bottom of the south load out pit (SLOP). The loaded MCO cask will be lifted from the SLOP transferred to the transport trailer and delivered to the Cold Vacuum Drying Facility (CVDF). In the wet condition there is a potential for criticality problems if significant changes in the designed fuel configurations occur. The purpose of this report is to address structural issues associated with criticality design features for MCO cask drop accidents in the 105 KE and 105 KW facilities.

  3. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    SciTech Connect

    Not Available

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  4. COBRA-SFS. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form the latest release of the code, Cycle 2.

  5. Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

    SciTech Connect

    Carbajo, J.J.; Lindner, C.N. )

    1992-01-01

    The design of spent-fuel casks for storage and transport requires modeling the cask for criticality, shielding, thermal, and structural analyses. While some parts of the cask are homogeneous, other regions are heterogeneous with different materials intermixed. For simplicity, some of the heterogeneous regions may be modeled as homogeneous. This paper evaluates the effect of homogenizing some regions of a cask on calculating radiation dose rates outside the cask. The dose rate calculations were performed with the one-dimensional discrete ordinates shielding XSDRNPM code coupled with the XSDOSE code and with the three-dimensional QAD-CGGP code. Dose rates were calculated radially at the midplane of the cask at two locations, cask surface and 2.3 m from the radial surface. The last location corresponds to a point 2 m from the lateral sides of a transport railroad car.

  6. COBRA-SFS CYCLE 3: Code System for Thermal Hydraulic Analysis of Spent Fuel Casks

    Energy Science and Technology Software Center (ESTSC)

    2003-11-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  7. 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.

  8. Casks (computer analysis of storage casks): A microcomputer based analysis system for storage cask review

    SciTech Connect

    Chen, T.F.; Mok, G.C.; Carlson, R.W.

    1995-08-01

    CASKS is a microcomputer based computer system developed by LLNL to assist the Nuclear Regulatory Commission in performing confirmatory analyses for licensing review of radioactive-material storage cask designs. The analysis programs of the CASKS computer system consist of four modules: the impact analysis module, the thermal analysis module, the thermally-induced stress analysis module, and the pressure-induced stress analysis module. CASKS uses a series of menus to coordinate input programs, cask analysis programs, output programs, data archive programs and databases, so the user is able to run the system in an interactive environment. This paper outlines the theoretical background on the impact analysis module and the yielding surface formulation. The close agreement between the CASKS analytical predictions and the results obtained form the two storage casks drop tests performed by SNL and by BNFL at Winfrith serves as the validation of the CASKS impact analysis module.

  9. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    SciTech Connect

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  10. A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

    SciTech Connect

    Andress, D. and Associates, Inc., Kensington, MD ); Joy, D.S. ); McLeod, N.B. Associates, Inc., Oakton, VA ); Peterson, R.W. and Associates, Inc., Alexandria, VA ); Rahimi, M. )

    1991-01-01

    The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elements as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.

  11. Scientific Ecology Group, Inc., 3-82B cask safety evaluation for packaging

    SciTech Connect

    Smith, R.J.

    1996-01-22

    This safety evaluation for packaging (SEP) provides the analysis and authorization to transport high-activity waste from the 324 Facility to PUREX, using the SEG 3-82B Type B cask. For the proposed campaign, the payload has larger quantities of radioactive material, is not fissile-exempt, and has higher decay heat loads than that specified by the 3-82B cask certificate of compliance. No changes will be made to the current design of the packaging. Onsite transport of the package with the higher source term will be authorized by this SEP to demonstrate equivalent safety of the package, as specified in PNL-MA-81, Hazardous Material Shipping Manual.

  12. Status update of the BWR cask simulator

    SciTech Connect

    Lindgren, Eric R.; Durbin, Samuel G.

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of

  13. Discussion of Available Methods to Support Reviews of Spent Fuel Storage Installation Cask Drop Evaluations

    SciTech Connect

    Witte, M.

    2000-03-28

    Applicants seeking a Certificate of Compliance for an Independent Spent Fuel Storage Installation (ISFSI) cask must evaluate the consequences of a handling accident resulting in a drop or tip-over of the cask onto a concrete storage pad. As a result, analytical modeling approaches that might be used to evaluate the impact of cylindrical containers onto concrete pads are needed. One such approach, described and benchmarked in NUREG/CR-6608,{sup 1} consists of a dynamic finite element analysis using a concrete material model available in DYNA3D{sup 2} and in LS-DYNA,{sup 3} together with a method for post-processing the analysis results to calculate the deceleration of a solid steel billet when subjected to a drop or tip-over onto a concrete storage pad. The analysis approach described in NUREG/CR-6608 gives a good correlation of analysis and test results. The material model used for the concrete in the analyses in NUREG/CR-6608 is, however, somewhat troublesome to use, requiring a number of material constants which are difficult to obtain. Because of this a simpler approach, which adequately evaluates the impact of cylindrical containers onto concrete pads, is sought. Since finite element modeling of metals, and in particular carbon and stainless steel, is routinely and accurately accomplished with a number of finite element codes, the current task involves a literature search for and a discussion of available concrete models used in finite element codes. The goal is to find a balance between a concrete material model with a limited number of required material parameters which are readily obtainable, and a more complex model which is capable of accurately representing the complex behavior of the concrete storage pad under impact conditions. The purpose of this effort is to find the simplest possible way to analytically represent the storage cask deceleration during a cask tip-over or a cask drop onto a concrete storage pad. This report is divided into three sections

  14. Shipping Cask Design Review Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1998-01-04

    Version 01 SCANS (Shipping Cask ANalysis System) is a microcomputer based system of computer programs and databases for evaluating safety analysis reports on spent fuel shipping casks. SCANS calculates the global response to impact loads, pressure loads, and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. Analysis options are based on regulatory cases described in the Code of Federal Regulations (1983) and Regulatory Guides published by the NRC in 1977more » and 1978. The system is composed of a series of menus and input entry cask analysis, and output display programs. An analysis is performed by preparing the necessary input data and then selecting the appropriate analysis: impact, thermal (heat transfer), thermally-induced stress, or pressure-induced stress. All data are entered through input screens with descriptive data requests, and, where possible, default values are provided. Output (i.e., impact force, moment and sheer time histories; impact animation; thermal/stress geometry and thermal/stress element outlines; temperature distributions as isocontours or profiles; and temperature time histories) is displayed graphically and can also be printed.« less

  15. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2002-06-26

    Document presents analysis performed to demonstrate criticality safety of packaging spent PWR fuel assemblies currently located at the 324 Building into a NAC-1 cask. Interim storage of the cask is also documented.

  16. CASKS (Computer Analysis of Storage casKS): A microcomputer based analysis system for storage cask design review. User`s manual to Version 1b (including program reference)

    SciTech Connect

    Chen, T.F.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L.; Mok, G.C.

    1995-02-01

    CASKS (Computer Analysis of Storage casKS) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent-fuel storage casks. The bulk of the complete program and this user`s manual are based upon the SCANS (Shipping Cask ANalysis System) program previously developed at LLNL. A number of enhancements and improvements were added to the original SCANS program to meet requirements unique to storage casks. CASKS is an easy-to-use system that calculates global response of storage casks to impact loads, pressure loads and thermal conditions. This provides reviewers with a tool for an independent check on analyses submitted by licensees. CASKS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests.

  17. Safety evaluation for packaging for the transport of K Basin sludge samples in the PAS-1 cask

    SciTech Connect

    SMITH, R.J.

    1998-11-17

    This safety evaluation for packaging authorizes the shipment of up to two 4-L sludge samples to and from the 325 Lab or 222-S Lab for characterization. The safety of this shipment is based on the current U.S. Department of Energy Certification of Compliance (CoC) for the PAS-1 cask, USA/9184/B(U) (DOE).

  18. Shielding and Containment Evaluations of the NAC-LWT Cask with Tritium Burnable Poison Rods

    SciTech Connect

    Holger Pfeifer; Norman Meinert

    2000-06-04

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the U.S. Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for a shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload.

  19. Shielding and containment evaluations of the NAC-LWT cask with tritium burnable poison rods

    SciTech Connect

    Pfeifer, H.; Meinert, N.

    2000-07-01

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the US Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for the shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload.

  20. Evaluation of the Cask Transportation Facility Modifications (CTFM) compliance to DOE order 6430.1A Project A.5 and A.6

    SciTech Connect

    ARD, K.E.

    2000-04-24

    This report was prepared to evaluate the compliance of CTFM to DOE Order 6430.1A. This document presents the results of an evaluation that was performed to assess compliance of the K West (KW) Cask Transportation Facility Modifications (CTFM) designs against applicable requirements of Department of Energy (DOE) Order 6430.1 A, General Design Criteria. This evaluation was grouped under two categories described as Cask Loadout System (CLS) and Cranes/Other Modifications.

  1. SCANS (Shipping Cask ANalysis System) a microcomputer based analysis system for shipping cask design review: Volume 4--Theory manual: Thermal analysis

    SciTech Connect

    Johnson, G.L.; Shapiro, A.B.

    1989-02-01

    TOPAZ is the two-dimensional, implicit, finite-element computer code included in the SCANS cask analysis system for heat conduction calculations. TOPAZ, a code developed on LLNL mainframes, has been implemented on IBM PC computers. This report provides documentation of TOPAZ controls and variables and a description of the numerical algorithms used. Sample problems with analytical solutions are presented. 10 refs., 32 figs., 11 tabs.

  2. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect

    Durbin, Samuel; Lindgren, Eric R.

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  3. DESIGN EVALUATION OF A LARGE CONCRETE CASK TO MEET IP-2 REQUIREMENTS

    SciTech Connect

    Shappert, L.B.

    2001-08-30

    Oak Ridge National Laboratory (ORNL) has a large quantity of low-level waste, in the form of concrete monoliths, that are stored in large concrete vaults in ORNL's Melton Valley Storage Tanks (MVST). During FY 2000, a number of the monoliths were transferred from the concrete vaults to a Nuclear Regulatory Commission (NRC)-certified lead-shielded cask and shipped to the Nevada Test Site (NTS) for disposal. This activity has resulted in (1) increased radiation exposure both when the monoliths were transferred to the lead-shielded cask and when they were unloaded and buried at the NTS and (2) high cask rental and shipping costs for the program, and (3) the accumulation of empty vaults at ORNL which will also have to be disposed of at NTS, adding a significant additional transportation cost. As a result, Department of Energy (DOE)--Oak Ridge has been exploring ways to ship the MVST cask with its monolith to the NTS for disposal as a unit. To do this, the MVST cask would have to be self-certified as meeting IP-2 package requirements.

  4. Adapting Dry Cask Storage for Aging at a Geologic Repository

    SciTech Connect

    C. Sanders; D. Kimball

    2005-08-02

    A Spent Nuclear Fuel (SNF) Aging System is a crucial part of operations at the proposed Yucca Mountain repository in the United States. Incoming commercial SNF that does not meet thermal limits for emplacement will be aged on outdoor pads. U.S. Department of Energy SNF will also be managed using the Aging System. Proposed site-specific designs for the Aging System are closely based upon designs for existing dry cask storage (DCS) systems. This paper evaluates the applicability of existing DCS systems for use in the SNF Aging System at Yucca Mountain. The most important difference between existing DCS facilities and the Yucca Mountain facility is the required capacity. Existing DCS facilities typically have less than 50 casks. The current design for the aging pad at Yucca Mountain calls for a capacity of over 2,000 casks (20,000 MTHM) [1]. This unprecedented number of casks poses some unique problems. The response of DCS systems to off-normal and accident conditions needs to be re-evaluated for multiple storage casks. Dose calculations become more complicated, since doses from multiple or very long arrays of casks can dramatically increase the total boundary dose. For occupational doses, the geometry of the cask arrays and the order of loading casks must be carefully considered in order to meet ALARA goals during cask retrieval. Due to the large area of the aging pad, skyshine must also be included when calculating public and worker doses. The expected length of aging will also necessitate some design adjustments. Under 10 CFR 72.236, DCS systems are initially certified for a period of 20 years [2]. Although the Yucca Mountain facility is not intended to be a storage facility under 10 CFR 72, the operational life of the SNF Aging System is 50 years [1]. Any cask system selected for use in aging will have to be qualified to this design lifetime. These considerations are examined, and a summary is provided of the adaptations that must be made in order to use DCS

  5. DYNAMIC NON LINEAR IMPACT ANALYSIS OF FUEL CASK CONTAINMENT VESSELS

    SciTech Connect

    Leduc, D

    2008-06-10

    Large fuel casks present challenges when evaluating their performance in the accident sequence specified in 10CFR 71. Testing is often limited because of cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing using simplified analytical methods. This paper details the use of dynamic non-linear analysis of large fuel casks using advanced computational techniques. Results from the dynamic analysis of two casks, the T-3 Spent Fuel Cask and the Hanford Un-irradiated Fuel Package are examined in detail. These analyses are used to fully evaluate containment vessel stresses and strains resulting from complex loads experienced by cask components during impacts. Importantly, these advanced analytical analyses are capable of examining stresses in key regions of the cask including the cask closure. This paper compares these advanced analytical results with the results of simplified cask analyses like those detailed in NUREG 3966.

  6. Spent Fuel Transportation Cask Response to the Caldecott Tunnel Fire Scenario

    SciTech Connect

    Adkins, Harold E.; Koeppel, Brian J.; Cuta, Judith M.

    2007-01-01

    On April 7, 1982, a tank truck and trailer carrying 8,800 gallons of gasoline was involved in an accident in the Caldecott tunnel on State Route 24 near Oakland, California. The tank trailer overturned and subsequently caught fire. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook analyses to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by truck. The Fire Dynamics Simulator (FDS) code developed by National Institute of Standards and Technology (NIST) was used to determine the thermal environment in the Caldecott tunnel during the fire. The FDS results were used to define boundary conditions for a thermal transient model of a truck transport cask containing spent nuclear fuel. The Nuclear Assurance Corporation (NAC) Legal Weight Truck (LWT) transportation cask was selected for this evaluation, as it represents a typical truck (over-the-road) cask, and can be used to transport a wide variety of spent nuclear fuels. Detailed analysis of the cask response to the fire was performed using the ANSYS® computer code to evaluate the thermal performance of the cask design in this fire scenario. This report describes the methods and approach used to assess the thermal response of the selected cask design to the conditions predicted in the Caldecott tunnel fire. The results of the analysis are presented in detail, with an evaluation of the cask response to the fire. The staff concluded that some components of smaller transportation casks resembling the NAC LWT, despite placement within an ISO container, could degrade significantly. Small transportation casks similar to the NAC LWT would probably experience failure of seals in this severe accident scenario. USNRC staff evaluated the radiological consequences of the cask response to the Caldecott tunnel fire. Although some

  7. Documentation for initial testing and inspections of Beneficial Uses Shipping System (BUSS) Cask

    SciTech Connect

    Lundeen, J.E.

    1994-08-25

    The purpose of this report is to compile data generated during the initial tests and inspections of the Beneficial Uses Shipping System (BUSS) Cask. In addition, this report will verify that the testing criteria identified in section 8.1 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The BUSS Cask body and lid are each one-piece forgings fabricated from ASTM A473, Type 304 stainless steel. The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Chapter 8 of the BUSS Cask SARP requires several acceptance tests and inspections, each intended to evaluate the performance of different components of the BUSS Cask system, to be performed before its first use. The results of the tests and inspections required are included in this document.

  8. Radiant heat transfer from storage casks to the environment

    SciTech Connect

    Carlson, R W; Hovingh, J; Thomas, G R

    1999-05-10

    A spent fuel storage cask must efficiently transfer the heat released by the fuel assemblies through the cask walls to the environment. This heat must be transferred through passive means, limiting the energy transfer mechanisms from the cask to natural convection and radiation heat transfer.. Natural convection is essentially independent of the characteristics of the array of casks, provided there is space between casks to permit a convection loop. Radiation heat transfer, however, depends on the geometric arrangement of the array of casks because the peripheral casks will shadow the interior casks and restrict radiant heat transfer from all casks to the environment. The shadowing of one cask by its neighbors is determined by a view factor that represents the fraction of radiant energy that leaves the surface of a cask and reaches the environment. This paper addresses the evaluation of the view factor between a centrally located spent fuel storage cask and the environment. By combining analytic expressions for the view factor of (1) infinitely long cylinders and (2) finite cylinders with a length-to-diameter ratio of 2 to represent spent fuel storage casks, the view factor can be evaluated for any practical array of spent fuel storage casks.

  9. COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.

    1986-12-01

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.

  10. A method for determining the spent-fuel contribution to transport cask containment requirements

    SciTech Connect

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Duffey, T.A.; Sutherland, S.H.; Reardon, P.C.

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  11. COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

    SciTech Connect

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Dodge, R.E.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS is a general thermal-hydraulic analysis computer code for prediction of material temperatures and fluid conditions in a wide variety of systems. The code has been validated for analysis of spent fuel storage systems, as part of the Commercial Spent Fuel Management Program of the US Department of Energy. The code solves finite volume equations representing the conservation equations for mass, moment, and energy for an incompressible single-phase heat transfer fluid. The fluid solution is coupled to a finite volume solution of the conduction equation in the solid structure of the system. This document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part 1 describes the conservation equations, constitutive models, and solution methods used in the code. Part 2 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part 3 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available.

  12. Fire resistant nuclear fuel cask

    DOEpatents

    Heckman, Richard C.; Moss, Marvin

    1979-01-01

    The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked.

  13. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect

    Pope, R B; Diggs, J M

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  14. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    SciTech Connect

    Richard, R.F.

    1995-05-11

    It has been postulated that a degradation phenomenon, referred to as ``hot cell rot``, may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ``Hot cell rot`` refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ``hot cell rot`` phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical.

  15. Solar Thermal Concept Evaluation

    NASA Technical Reports Server (NTRS)

    Hawk, Clark W.; Bonometti, Joseph A.

    1995-01-01

    Concentrated solar thermal energy can be utilized in a variety of high temperature applications for both terrestrial and space environments. In each application, knowledge of the collector and absorber's heat exchange interaction is required. To understand this coupled mechanism, various concentrator types and geometries, as well as, their relationship to the physical absorber mechanics were investigated. To conduct experimental tests various parts of a 5,000 watt, thermal concentrator, facility were made and evaluated. This was in anticipation at a larger NASA facility proposed for construction. Although much of the work centered on solar thermal propulsion for an upper stage (less than one pound thrust range), the information generated and the facility's capabilities are applicable to material processing, power generation and similar uses. The numerical calculations used to design the laboratory mirror and the procedure for evaluating other solar collectors are presented here. The mirror design is based on a hexagonal faceted system, which uses a spherical approximation to the parabolic surface. The work began with a few two dimensional estimates and continued with a full, three dimensional, numerical algorithm written in FORTRAN code. This was compared to a full geometry, ray trace program, BEAM 4, which optimizes the curvatures, based on purely optical considerations. Founded on numerical results, the characteristics of a faceted concentrator were construed. The numerical methodologies themselves were evaluated and categorized. As a result, the three-dimensional FORTRAN code was the method chosen to construct the mirrors, due to its overall accuracy and superior results to the ray trace program. This information is being used to fabricate and subsequently, laser map the actual mirror surfaces. Evaluation of concentrator mirrors, thermal applications and scaling the results of the 10 foot diameter mirror to a much larger concentrator, were studied. Evaluations

  16. A validated methodology for evaluating burnup credit in spent fuel casks

    SciTech Connect

    Brady, M.C. ); Sanders, T.L. )

    1991-01-01

    The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in k{sub eff}. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs.

  17. SCANS (Shipping Cask ANalysis System) a microcomputer-based analysis system for shipping cask design review: User`s manual to Version 3a. Volume 1, Revision 2

    SciTech Connect

    Mok, G.C.; Thomas, G.R.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L.

    1998-03-01

    SCANS (Shipping Cask ANalysis System) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent fuel shipping casks. SCANS is an easy-to-use system that calculates the global response to impact loads, pressure loads and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. SCANS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests. Analysis options are based on regulatory cases described in the Code of Federal Regulations 10 CFR 71 and Regulatory Guides published by the US Nuclear Regulatory Commission in 1977 and 1978.

  18. Evaluation of Helium Purge & Vent Process to Reduce Hydrogen Concentrations in the Large Diameter Container & Cask Void Volumes at T Plant

    SciTech Connect

    PACKER, M.J.

    2002-10-15

    The purpose of this document is to provide calculations for two primary activities: (1) Model a Helium Purge/Vent Cycle Process to reduce hydrogen gas concentration (i.e., H{sub 2} mole fraction) to a required limit in the Cask and Large Diameter Container (LDC) void volumes prior to T-Plant Operations activities. (2) Predict a hydrogen generation rate within each sludge-contained LDC, after the T-Plant helium purge/vent process (aka Post Purge/Vent Cycle Duration) to determine the transient hydrogen concentration. The calculations will evaluate a helium purge process to reduce the hydrogen concentration in the void spaces of the LDC after receipt at T-Plant. During transport from K-Basins to T-Plant, the hydrogen concentration will increase but the low or absent oxygen concentration from the K-Basin helium purge/vent process will ensure a non-flammable event. Upon receipt at T-Plant, the increased hydrogen concentration will require a process reduction (i.e., helium purge/vent cycling) prior to removing the Cask lid, otherwise, the removed lid permits air ingress and associated oxygen with the assumed high hydrogen concentration. In addition, once the Cask lid is removed at T-Plant, and the LDC is moved to the process cell, two threaded caps must be removed from the LDC to allow the escape of hydrogen during long-term storage. It is essential that the T-Plant helium purge/vent system reduces the hydrogen in both the Cask and LDC void volumes below the required limit. The calculations will also aide in predicting actual hydrogen generation rates and concentrations in each of the void volumes after the helium purge/vent cycle process is completed. Transient hydrogen plots or figures will be provided to help achieve this objective.

  19. A comparison of spent fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    SciTech Connect

    Manson, S.J.; Gianoulakis, S.E.

    1994-04-01

    An examination of the effect of a realistic (though conservative) hot day environment on the thermal transient behavior of spent fuel shipping casks is made. These results are compared to those that develop under the prescribed normal thermal condition of 10 CFR 71. Of specific concern are the characteristics of propagating thermal waves, which are set up by diurnal variations of temperature and insolation in the outdoor environment. In order to arrive at a realistic approximation of these variations on a conservative hot day, actual temperature and insolation measurements have been obtained from the National Climatic Data Center (NCDC) for representatively hot and high heat flux days. Thus, the use of authentic meteorological data ensures the realistic approach sought. Further supporting the desired realism of the modeling effort is the use of realistic cask configurations in which multiple laminations of structural, shielding, and other materials are expected to attenuate the propagating thermal waves. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by enforcement of the regulatory environmental conditions of 10 CFR 71. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the prescribed regulatory conditions. However, the temperature differences are small enough that the normal conservative assumptions that are made in the course of typical cask evaluations should correct for any potential violations. The analysis demonstrates that diurnal temperature variations that penetrate the cask wall all have maxima substantially less than the corresponding regulatory solutions. Therefore it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the conditions of 10 CFR 71.

  20. Evaluation of low-velocity impact tests of solid steel billet onto concrete pads, and application to generic ISFSI storage cask for tipover and side drop

    SciTech Connect

    Witte, M. C.; Chen, T.F.; Mok, G.C.; Murty, S.S.; Fischer, L.E.

    1997-03-01

    Spent Fuel Storage Casks intended for use at Independent Spent Fuel Storage Installations (ISFSIS) typically are evaluated during the application and review process for low-energy impacts representative of possible handling accidents including tipover events. In the past, the analyses involved in these evaluations have assumed that the casks dropped or tipped onto an unyielding surface, a conservative and simplifying assumption. Since 10 CFR Part 72`, the regulation imposed by the Nuclear Regulatory Commission (NRC), does not require this assumption, applicants are currently seeking a more realistic model for the analyses and are using analytical models which predict the effect of a cask dropping onto a reinforced concrete pad, including energy absorbing aspects such as cracking and flexure. In order to develop data suitable for benchmarking these analyses, the NRC has conducted several series of drop-test studies. The tests described in this report were primarily intended to determine the response characteristics of concrete pads during tipover and side impacts of a solid steel billet onto the pads. This series of tests is fourth in a program of tests funded by the NRC; all four series of tests address issues of impact involving spent fuel storage casks. The first series was performed in March 1993 by Sandia National Laboratories (SNL) and involved five end-drops of a billet, nearly identical to the one used in the present series, onto a variety of surfaces from a height of 18 inches. The second series of tests was performed between July and October 1993, and involved four end- drops of a near-full-scale empty Excellox 3A cask onto a full-scale concrete pad and foundation, or onto an essentially unyielding surface, from heights ranging from 18 inches to 60 inches, and was conducted by the British Nuclear Fuels Limited in Winfrith, England. (Two of the drops in the second series were sponsored by Electric Power Research Institute.) The third test series was

  1. Performance of bolted closure joint elastomers under cask aging conditions

    SciTech Connect

    Verst, C.; Sindelar, R.; Skidmore, E.; Daugherty, W.

    2015-07-23

    The bolted closure joint of a bare spent fuel cask is susceptible to age-related degradation and potential loss of confinement function under long-term storage conditions. Elastomeric seals, a component of the joint typically used to facilitate leak testing of the primary seal that includes the metallic seal and bolting, is susceptible to degradation over time by several mechanisms, principally via thermo-oxidation, stress-relaxation, and radiolytic degradation under time and temperature condition. Irradiation and thermal exposure testing and evaluation of an ethylene-propylene diene monomer (EPDM) elastomeric seal material similar to that used in the CASTOR® V/21 cask for a matrix of temperature and radiation exposure conditions relevant to the cask extended storage conditions, and development of semiempirical predictive models for loss of sealing force is in progress. A special insert was developed to allow Compressive Stress Relaxation (CSR) measurements before and after the irradiation and/or thermal exposure without unloading the elastomer. A condition of the loss of sealing force for the onset of leakage was suggested. The experimentation and modeling being performed could enable acquisition of extensive coupled aging data as well as an estimation of the timeframe when loss of sealing function under aging (temperature/radiation) conditions may occur.

  2. The Performance of Spent Fuel Casks in Severe Tunnel Fires

    SciTech Connect

    Bajwa, C.S.; Easton, E.P.; Hansen, A.

    2006-07-01

    The Nuclear Regulatory Commission (NRC), working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of various spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. The thermal performance of three different spent fuel cask designs (HOLTEC HI-STAR 100, TransNuclear TN-68, and NAC-LWT) was evaluated with the ANSYS{sup R} and COBRA-SFS analysis codes, utilizing boundary conditions for the tunnel fire obtained using NIST's Fire Dynamics Simulator (FDS) code. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of these analyses are described in detail in Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario, NUREG/CR-6886, published in draft for comment in November 2005. Comments received by the NRC on NUREG/CR-6886 will be addressed in the final version of the report. (authors)

  3. Cask Processing Enclosure Specification/Operation - 12231

    SciTech Connect

    Gentry, Ronald

    2012-07-01

    Following an evaluation of throughput rates in the Hot Cell at the Transuranic Waste Processing Center and considering the variability in the waste with respect to actual dose rates a new approach to processing transuranic waste was necessary. Compounding the issue was the remote equipment poor reliability and high down-time. After considering all the factors, the evaluation resulted in the design and construction of a new waste processing area for handling the concrete casks that predominately contain contact handled transuranic (TRU) waste. The area is called the Cask Processing Enclosure and essentially the Cask Processing Enclosure mimics the projects current process techniques used for processing Contact Handled -TRU waste in the existing Box Breakdown Area and Glovebox. The Cask Processing Enclosure approach was developed based on a review of the RH processing throughput rates in the Hot Cell. As the process was reviewed consideration was given to the variability in the waste with respect to actual dose rates and the lack of equipment reliability and high wear in the Hot Cell. Based on that review, a new contact handled processing area for handling the concrete casks is being constructed and startup is expected shortly following WM2012. The Cask Processing Enclosure essentially mimics the projects current process techniques used for processing Contact Handled waste in the existing Box Breakdown Area and Glovebox and the design takes into consideration six years of operational experience. (authors)

  4. European experience in transport/storage cask for vitrified residues

    SciTech Connect

    Otton, Camille; Sicard, Damien

    2007-07-01

    Available in abstract form only. Full text of publication follows: Because of the evolution of burnup of spent fuel to be reprocessed, the high activity vitrified residues would not be transported in the existing cask designs. Therefore, TN International has decided in the late nineties to develop a brand new design of casks with optimized capacity able to store and transport the most active and hottest canisters: the TN{sup TM}81 casks currently in use in Switzerland and the TN{sup TM}85 cask which shall permit in the near future in Germany the storage and the transport of the most active vitrified residues defining a thermal power of 56 kW (kilowatts). The challenges for the TN{sup TM}81 and TN{sup TM}85 cask designs were that the geometry entry data were very restrictive and were combined with a fairly wide range set by the AREVA NC Specification relative to vitrified residue canister. The TN{sup TM}81 and the TN{sup TM}85 casks have been designed to fully anticipate shipment constraints of the present vitrified residue production. It also used the feedback of current shipments and the operational constraints and experience of receiving and shipping facilities. The casks had to fit as much as possible in the existing procedures for the already existing flasks such as the TN{sup TM}28 cask and TS 28 V cask, all along the logistics chain of loading, unloading, transport and maintenance. (authors)

  5. Alternative Cask Maintenance Facility concepts, an update and reassessment

    SciTech Connect

    Attaway, C.R.; Medley, L.B.; Williamson, A.; Pope, R.B.; Shappert, L.B.

    1992-02-01

    The results of three trade-off studies of alternative concepts for performing cask maintenance for Civilian Radioactive Waste Management System casks are presented. An earlier study resulted in a recommendation that a submerged pool concept for cask internal component removal be used in the design of a Cask Maintenance Facility. The first trade-off study resulted in confirming the previous recommendation that a submerged pool concept be used rather than an isolation cell; the basis for this continued recommendation is discussed. The second study provides an evaluation of the previously proposed facility for the capability of handling an increased quantity of OCRWM casks. This third study provides a preliminary concept for adding the capability to repaint the exterior cylindrical portions of casks.

  6. The Feasibility of Cask "Fingerprinting" as a Spent-Fuel, Dry-Storage Cask Safeguards Technique

    SciTech Connect

    Ziock, K P; Vanier, P; Forman, L; Caffrey, G; Wharton, J; Lebrun, A

    2005-07-27

    This report documents a week-long measurement campaign conducted on six, dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. A gamma-ray imager, a thermal-neutron imager and a germanium spectrometer were used to collect data on the casks. The campaign was conducted to examine the feasibility of using the cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the germanium spectrometer and differences between thermal neutron images and neutron-dose meters, this result is thought to be due to the limitations of the extant imagers used, rather than of the basic concept. Results indicate that measurements with improved imagers could contain significantly more information. Follow-on measurements with new imagers either currently available as laboratory prototypes or under development are recommended.

  7. A cask fleet operations study

    SciTech Connect

    Not Available

    1988-03-01

    This document describes the cask fleet currently available to transport spent nuclear fuels. The report describes the proposed operational procedures for these casks and the vehicles intended to transport them. Included are techniques for loading the cask, lifting it onto the transport vehicle, preparing the invoices, and unloading the cask at the destination. The document concludes with a discussion on the maintenance and repair of the casks. (tem) 29 figs.

  8. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2001-11-01

    This analysis references a previous analysis (Larson, 1999) for a qualitative acceptability argument, and an appropriate ANS standard (ANS, 1998), and a reference (Clark 1966) for safe cylinder diameters as a function of {sup 235}U enrichment in UO{sub 2} fuel rods (pins) in water, for a quantitative acceptability argument. The previous analysis established the criticality safety of PWR assemblies without broken pins, pin segments or powdered fuel. This addendum extends the previous range of applicability in accordance with the controlling procedure (FH 2001b). The operation, when conducted according to the established limits stated in this document, complies with the incredibility principle. The evaluations demonstrated criticality safety for PWR assemblies with broken pins and pin segments of UO{sub 2} fuel.

  9. Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

    SciTech Connect

    Powell, F.P.

    1995-04-01

    This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask`s primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives.

  10. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect

    Forsberg, Charles W.; Dole, Leslie R.

    2007-07-01

    Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can out-perform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal conductivity cask. A multi-year, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports. (authors)

  11. 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.

  12. Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

    SciTech Connect

    Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa; Koji Shirai; Akihiro Sasahara; Takatoshi Hattori

    2006-04-01

    In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstration project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.

  13. FACSIM/MRS-1: Cask receiving and consolidation performance assessment

    SciTech Connect

    Lotz, T.L.; Shay, M.R.

    1987-06-01

    A simulation analysis was completed to assess the performance of the shipping cask receiving and spent-fuel handling, consolidation and canistering operations of the Monitored Retrievable Storage (MRS) facility. One purpose of this evaluation was to estimate the limits of MRS operational capabilities and factors leading to those limitations. The model used to obtain the performance assessment, FACSIM/MRS-1, is one of two components of the FACSIM model developed by PNL's simulation effort for the nuclear waste-handling facility. FACSIM/MRS-1 provides the user with information about lag-storage requirements, machine use, cask queues, welder queues, and cask process and cask turnaround times. The model can help determine the effect that the following activities have on operating efficiency: (1) receiving multiple cask shipments, when rail-cask or truck-cask shipments arrive at the facility in groups of two or more, and (2) operating the facility five days per week, three shifts per day or seven days per week, three shifts per day for any conditions. In addition, sensitivity to equipment failure frequency and the time needed for equipment repair can be studied. Information on the above operating characteristics may be obtained for any spent-fuel rate, any split of shipments between truck and rail transport, or any split of boiling water reactor/pressurized water reactor fuel.

  14. Viability of Existing INL Facilities for Dry Storage Cask Handling

    SciTech Connect

    Randy Bohachek; Charles Park; Bruce Wallace; Phil Winston; Steve Marschman

    2013-04-01

    This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

  15. Viability of Existing INL Facilities for Dry Storage Cask Handling

    SciTech Connect

    Bohachek, Randy; Wallace, Bruce; Winston, Phil; Marschman, Steve

    2013-04-30

    This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

  16. SLI Thermal Imaging Requirements Evaluation

    NASA Astrophysics Data System (ADS)

    Hoffman, E. H.; Woody, L. M.; Wirth, S. M.; Smith, D. S.

    2015-12-01

    The Landsat program has provided a continuous record of global terrestrial imagery since 1972. This data record is an invaluable resource for determining long term trends and monitoring rates of change in land usage, forest health, water quality, and glacier retreat. In 2014, the National Aeronautics and Space Administration (NASA), supported by the United States Geological Survey (USGS), initiated the sustainable land imaging (SLI) architecture study to develop an affordable system design for acquiring future terrestrial imagery compatible with the existing Landsat data record. The principal objective has been to leverage recent advances in focal plane technologies to enable smaller, lower-cost instruments and launch options. We present an evaluation of the trade space implied by the SLI thermal imaging requirements as well as the performance potential of enabling technologies. Multiple approaches, each incorporating measured performance data for state-of-the-art detectors, are investigated to simultaneously optimize instrument mass and volume, spatial response, radiometric sensitivity, and radiometric uncertainty.

  17. Design analysis report for the TN-WHC cask and transportation system

    SciTech Connect

    Brisbin, S.A., Fluor Daniel Hanford

    1997-02-13

    This document presents the evaluation of the Spent Nuclear Fuel Cask and Transportation System. The system design was developed by Transnuclear, Inc. and its team members NAC International, Nelson Manufacturing, Precision Components Corporation, and Numatec, Inc. The cask is designated the TN-WHC cask. This report describes the design features and presents preliminary analyses performed to size critical dimensions of the system while meeting the requirements of the performance specification.

  18. Thermal barrier coating evaluation needs

    NASA Technical Reports Server (NTRS)

    Brindley, William J.; Miller, Robert A.

    1990-01-01

    A 0.025 cm (0.010 in) thick thermal barrier coating (TBC) applied to turbine airfoils in a research gas turbine engine provided component temperature reductions of up to 190 C. These impressive temperature reductions can allow increased engine operating temperatures and reduced component cooling to achieve greater engine performance without sacrificing component durability. The significant benefits of TBCs are well established in aircraft gas turbine engine applications and their use is increasing. TBCs are also under intense development for use in the Low Heat Rejection (LHR) diesel engine currently being developed and are under consideration for use in utility and marine gas turbines. However, to fully utilize the benefits of TBCs it is necessary to accurately characterize coating attributes that affect the insulation and coating durability. The purpose there is to discuss areas in which nondestructive evaluation can make significant contributions to the further development and full utilization of TBCs for aircraft gas turbine engines and low heat rejection diesel engines.

  19. Compton Dry-Cask Imaging System

    ScienceCinema

    None

    2013-05-28

    The Compton-Dry Cask Imaging Scanner is a system that verifies and documents the presence of spent nuclear fuel rods in dry-cask storage and determines their isotopic composition without moving or opening the cask. For more information about this project, visit http://www.inl.gov/rd100/2011/compton-dry-cask-imaging-system/

  20. Compton Dry-Cask Imaging System

    SciTech Connect

    2011-01-01

    The Compton-Dry Cask Imaging Scanner is a system that verifies and documents the presence of spent nuclear fuel rods in dry-cask storage and determines their isotopic composition without moving or opening the cask. For more information about this project, visit http://www.inl.gov/rd100/2011/compton-dry-cask-imaging-system/

  1. Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

    SciTech Connect

    Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

    2002-07-01

    Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis

  2. A new type-B cask design for transporting {sub 252}Cf

    SciTech Connect

    Simmons, C.M.

    2000-07-01

    A project to design, certify, and build a new US Department of Energy (DOE) Type B container for transporting >5 mg of {sup 252}Cf is more than halfway to completion. This project was necessitated by the fact that the existing Oak Ridge National Laboratory (ORNL) Type B containers were designed and built many years ago and thus do not have the records and supporting data that current regulations require. Once the new cask is available, it will replace the existing Type B containers. The cask design is driven by the unique properties of {sup 252}Cf, which is a very intense spontaneous fission neutron source and necessitates a large amount of neutron shielding. The cask is designed to contain up to 60 mg of {sup 252}Cf in the form of californium oxide or californium oxysulfate, in pellet, wire, or sintered material forms that are sealed inside small special-form capsules. The new cask will be capable of all modes of transport (land, sea, and air). The ORNL team, composed of technical and purchasing personnel and using rigorous selection criteria, chose NAC, International (NAC), as the subcontractor for the project. In January 1997, NAC started work on developing the conceptual design and performing the analyses. The original design concept was for a tungsten alloy gamma shield surrounded by two concentric shells of NS-4-FR neutron shield material. A visit to US Nuclear Regulatory Commission (NRC) regulators in November 1997 to present the conceptual design for their comments resulted in a design modification when the question of potential straight-line cracking in the NS-4-FR neutron shield material arose. NAC's modified design includes offset, wedgelike segments of the neutron shield material. The new geometry eliminates concerns about straight-line cracking but increases the weight of the packaging and makes the fabrication more complex. NAC has now completed the cask design and performed the analyses (shielding, structural, thermal, etc.) necessary to certify the

  3. Parametric study of radiation dose rates from rail and truck spent fuel transport casks

    SciTech Connect

    Parks, C.V.; Hermann, O.W.; Knight, J.R.

    1985-08-01

    Neutron and gamma dose rates from typical rail and truck spent fuel transport casks are reported for a variety of spent PWR fuel sources and cask conditions. The IF 300 rail cask and NLI 1/2 truck cask were selected for use as appropriate cask models. All calculations (cross section preparation, generation of spent fuel source terms, radiation transport calculations, and dose evaluation) were performed using various modules of the SCALE computational system. Conditions or parameters for which there were variations between cases include: detector distance from cask, spent fuel cooling time, the setting of fuel or neutron shielding cavities to either wet or dry, the cobalt content of assembly materials, normal fuel assemblies and consolidated cannisters, the geometry mesh interval size, and the order of the angular quadrature set. 13 refs., 6 figs., 9 tabs.

  4. Assessment of spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    SciTech Connect

    Daling, P.M.

    1985-08-01

    Realistic truck/rail modal fractions are specifically needed to support the Monitored Retrievable Storage (MRS) and repository facility designs and envirionmental assessment activities. The objective of this study was to evaluate the spent fuel shipping cask handling capabilities at operating and planned commercial LWRs and use this information to estimate realistic truck/rail modal fractions. The cask handling parameter data collected in this study includes cask handling crane capabilities, dimensions of loading pools, structural limits, availability of rail service, past experience with spent fuel shipments (i.e., which cask was used.), and any other conditions which could impede or preclude use of a particular shipping cask. The results of this evaluation are presented for each reactor. A summary of the results which indicates the number of plants that are capable of handling each transport mode is presented. Note that two types of highway shipments are considered; legal-weight truck (LWT) and overweight truck (OWT). The primary differences between these two types of highway shipments are the size and cargo capacity of the spent fuel shipping casks. The OWT cask is roughly 50% heavier, 50% larger in diameter, and has a 300% larger cargo capacity. As a result of this size differential, some plants are capable of handling LWT casks but not OWT casks.

  5. CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect

    E.F. Loros

    2000-06-23

    The Carrier/Cask Handling System receives casks on railcars and legal-weight trucks (LWTs) (transporters) that transport loaded casks and empty overpacks to the Monitored Geologic Repository (MGR) from the Carrier/Cask Transport System. Casks that come to the MGR on heavy-haul trucks (HHTs) are transferred onto railcars before being brought into the Carrier/Cask Handling System. The system is the interfacing system between the railcars and LWTs and the Assembly Transfer System (ATS) and Canister Transfer System (CTS). The Carrier/Cask Handling System removes loaded casks from the cask transporters and transfers the casks to a transfer cart for either the ATS or CTS, as appropriate, based on cask contents. The Carrier/Cask Handling System receives the returned empty casks from the ATS and CTS and mounts the casks back onto the transporters for reshipment. If necessary, the Carrier/Cask Handling System can also mount loaded casks back onto the transporters and remove empty casks from the transporters. The Carrier/Cask Handling System receives overpacks from the ATS loaded with canisters that have been cut open and emptied and mounts the overpacks back onto the transporters for disposal. If necessary, the Carrier/Cask Handling System can also mount empty overpacks back onto the transporters and remove loaded overpacks from them. The Carrier/Cask Handling System is located within the Carrier Bay of the Waste Handling Building System. The system consists of cranes, hoists, manipulators, and supporting equipment. The Carrier/Cask Handling System is designed with the tooling and fixtures necessary for handling a variety of casks. The Carrier/Cask Handling System performance and reliability are sufficient to support the shipping and emplacement schedules for the MGR. The Carrier/Cask Handling System interfaces with the Carrier/Cask Transport System, ATS, and CTS as noted above. The Carrier/Cask Handling System interfaces with the Waste Handling Building System for building

  6. Inspection of Used Fuel Dry Storage Casks

    SciTech Connect

    Dennis C. Kunerth; Tim McJunkin; Mark McKay; Sasan Bakhtiari

    2012-09-01

    ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) regulates the storage of used nuclear fuel, which is now and will be increasingly placed in dry storage systems. Since a final disposition pathway is not defined, the fuel is expected to be maintained in dry storage well beyond the time frame originally intended. Due to knowledge gaps regarding the viability of current dry storage systems for long term use, efforts are underway to acquire the technical knowledge and tools required to understand the issues and verify the integrity of the dry storage system components. This report summarizes the initial efforts performed by researchers at Idaho National Laboratory and Argonne National Laboratory to identify and evaluate approaches to in-situ inspection dry storage casks. This task is complicated by the design of the current storage systems that severely restrict access to the casks.

  7. Evaluation of impact limiter performance during end-on and slapdown drop tests of a one-third scale model storage/transport cask system

    SciTech Connect

    Yoshimura, H.R.; Bronowski, D.R.; Uncapher, W.L.; Attaway, S.W.; Bateman, V.I.; Carne, T.G.; Gregory, D.L. ); Huerta, M. )

    1990-12-01

    This report describes drop testing of a one-third scale model shipping cask system. Two casks were designed and fabricated by Transnuclear, Inc., to ship spent fuel from the former Nuclear Fuel Services West Valley reprocessing facility in New York to the Idaho National Engineering Laboratory for a long-term spent fuel dry storage demonstration project. As part of the NRC's regulatory certification process, one-third scale model tests were performed to obtain experimental data on impact limiter performance during impact testing. The objectives of the testing program were to (1) obtain deceleration and displacement information for the cask and impact limiter system, (2) obtain dynamic force-displacement data for the impact limiters, (3) verify the integrity of the impact limiter retention system, and (4) examine the crush behavior of the limiters. Two 30-ft (9-m) drop tests were conducted on a mass model of the cask body and scaled balsa and redwood-filled impact limiters. This report describes the results of both tests in terms of measured decelerations, posttest deformation measurements, and the general structural response of the system. 3 refs., 32 figs.

  8. JWST ISIM Harness Thermal Evaluation

    NASA Technical Reports Server (NTRS)

    Kobel, Mark; Glazer, Stuart; Tuttle, Jim; Martins, Mario; Ruppel, Sean

    2008-01-01

    The James Webb Space Telescope (JWST) will be a large infrared telescope with a 6.5-meter primary mirror. Launch is planned for 2013. JWST wl1 be the premier observatory of the next decade serving thousands of astronomers worldwide. The Integrated Science Instrument Module (ISIM) is the unit that will house thc four main JWST instruments. The ISIM enclosure passively cooled to 37 Kelvin and has a tightly managed thermal budget. A significant portion of the ISIM heat load is due to parasitic heat gains from the instrument harnesses. These harnesses provide a thermal path from the Instrument Electronics Control (IEC) to the ISIM. Because of the impact of this load to the ISIM thermal design, understanding the harness parasitic heat gains is critical. To this effect, a thermal test program has been conducted in order to characterize these parasitic loads and verify harness thermal models. Recent parasitic heat loads tests resulted in the addition of a dedicated multiple stage harness radiator. In order for the radiator to efficiently reject heat from the harness, effective thermal contact conductance values for multiple harnesses had to be determined. This presentation will describe the details and the results of this test program.

  9. Routine methods for post-transportation accident recovery of spent fuel casks

    SciTech Connect

    Shappert, L.B.; Pope, R.B. ); Best, R.E. ); Jones, R.H. , Los Gatos, CA )

    1991-01-01

    Spent fuel casks and other large radioactive material packages have been examined to determine whether the designs are adequate to allow the casks to be recovered using conventional recovery methods following a transportation accident. Casks and similar packages are typically designed with, and handled by, trunnions that support the package during transport. These trunnions are considered the best cask feature with which to grapple the cask once it is no longer in its usual shipping mode. Following a transport accident, the trunnions may be buried or entangled so that they are not readily accessible to initiate the recovery process. To evaluate the effectiveness of applying traditional recovery methods to spent fuel casks, a workshop was held in which a series of accidents involving casks were postulated; the modes of transportation considered included truck, rail, and barge. These participants knowledgeable in transport, handling, and, in some cases, recovery of large, heavy containers attended. Participants concluded that the physical recovery of a cask involved in an accident, irrespective of where the accident occurs, would be a straightforward rigging operation and that the addition of specific recovery features (e.g., additional trunnions) to the cask appears unnecessary.

  10. Routine methods for post-transportation accident recovery of spent fuel casks

    SciTech Connect

    Shappert, L.B.; Pope, R.B.; Best, R.E.; Jones, R.H.

    1991-12-31

    Spent fuel casks and other large radioactive material packages have been examined to determine whether the designs are adequate to allow the casks to be recovered using conventional recovery methods following a transportation accident. Casks and similar packages are typically designed with, and handled by, trunnions that support the package during transport. These trunnions are considered the best cask feature with which to grapple the cask once it is no longer in its usual shipping mode. Following a transport accident, the trunnions may be buried or entangled so that they are not readily accessible to initiate the recovery process. To evaluate the effectiveness of applying traditional recovery methods to spent fuel casks, a workshop was held in which a series of accidents involving casks were postulated; the modes of transportation considered included truck, rail, and barge. These participants knowledgeable in transport, handling, and, in some cases, recovery of large, heavy containers attended. Participants concluded that the physical recovery of a cask involved in an accident, irrespective of where the accident occurs, would be a straightforward rigging operation and that the addition of specific recovery features (e.g., additional trunnions) to the cask appears unnecessary.

  11. Cask fleet operations study

    SciTech Connect

    Not Available

    1988-01-01

    The Nuclear Waste Policy Act of 1982 assigned to the Department of Energy's (DOE) Office of Civilian Waste Management the responsibility for disposing of high-level waste and spent fuel. A significant part of that responsibility involves transporting nuclear waste materials within the federal waste management system; that is, from the waste generator to the repository. The lead responsibility for transportation operations has been assigned to Oak Ridge Operations, with Oak Ridge National Laboratory (ORNL) providing technical support through the Transportation Operations Support Task Group. One of the ORNL support activities involves assessing what facilities, equipment and services are required to assure that an acceptable, cost-effective and safe transportation operations system can be designed, operated and maintained. This study reviews, surveys and assesses the experience of Nuclear Assurance Corporation (NAC) in operating a fleet of spent-fuel shipping casks to aid in developing the spent-fuel transportation system.

  12. Implications of the Baltimore Rail Tunnel Fire for Full-Scale Testing of Shipping Casks

    SciTech Connect

    Halstead, R. J.; Dilger, F.

    2003-02-25

    The U.S. Nuclear Regulatory Commission (NRC) does not currently require full-scale physical testing of shipping casks as part of its certification process. Stakeholders have long urged NRC to require full-scale testing as part of certification. NRC is currently preparing a full-scale casktesting proposal as part of the Package Performance Study (PPS) that grew out of the NRC reexamination of the Modal Study. The State of Nevada and Clark County remain committed to the position that demonstration testing would not be an acceptable substitute for a combination of full-scale testing, scale-model tests, and computer simulation of each new cask design prior to certification. Based on previous analyses of cask testing issues, and on preliminary findings regarding the July 2001 Baltimore rail tunnel fire, the authors recommend that NRC prioritize extra-regulatory thermal testing of a large rail cask and the GA-4 truck cask under the PPS. The specific fire conditions and other aspects of the full-scale extra-regulatory tests recommended for the PPS are yet to be determined. NRC, in consultation with stakeholders, must consider past real-world accidents and computer simulations to establish temperature failure thresholds for cask containment and fuel cladding. The cost of extra-regulatory thermal testing is yet to be determined. The minimum cost for regulatory thermal testing of a legal-weight truck cask would likely be $3.3-3.8 million.

  13. FUEL CASK IMPACT LIMITER VULNERABILITIES

    SciTech Connect

    Leduc, D; Jeffery England, J; Roy Rothermel, R

    2009-02-09

    Cylindrical fuel casks often have impact limiters surrounding just the ends of the cask shaft in a typical 'dumbbell' arrangement. The primary purpose of these impact limiters is to absorb energy to reduce loads on the cask structure during impacts associated with a severe accident. Impact limiters are also credited in many packages with protecting closure seals and maintaining lower peak temperatures during fire events. For this credit to be taken in safety analyses, the impact limiter attachment system must be shown to retain the impact limiter following Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) impacts. Large casks are often certified by analysis only because of the costs associated with testing. Therefore, some cask impact limiter attachment systems have not been tested in real impacts. A recent structural analysis of the T-3 Spent Fuel Containment Cask found problems with the design of the impact limiter attachment system. Assumptions in the original Safety Analysis for Packaging (SARP) concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.

  14. Human Thermal Model Evaluation Using the JSC Human Thermal Database

    NASA Technical Reports Server (NTRS)

    Bue, Grant; Makinen, Janice; Cognata, Thomas

    2012-01-01

    Human thermal modeling has considerable long term utility to human space flight. Such models provide a tool to predict crew survivability in support of vehicle design and to evaluate crew response in untested space environments. It is to the benefit of any such model not only to collect relevant experimental data to correlate it against, but also to maintain an experimental standard or benchmark for future development in a readily and rapidly searchable and software accessible format. The Human thermal database project is intended to do just so; to collect relevant data from literature and experimentation and to store the data in a database structure for immediate and future use as a benchmark to judge human thermal models against, in identifying model strengths and weakness, to support model development and improve correlation, and to statistically quantify a model s predictive quality. The human thermal database developed at the Johnson Space Center (JSC) is intended to evaluate a set of widely used human thermal models. This set includes the Wissler human thermal model, a model that has been widely used to predict the human thermoregulatory response to a variety of cold and hot environments. These models are statistically compared to the current database, which contains experiments of human subjects primarily in air from a literature survey ranging between 1953 and 2004 and from a suited experiment recently performed by the authors, for a quantitative study of relative strength and predictive quality of the models.

  15. A robotic system to conduct radiation and contamination surveys on nuclear waste transport casks

    SciTech Connect

    Harrigan, R.W.; Sanders, T.L.

    1990-06-01

    The feasibility of performing, numerous spent fuel cask operations using fully integrated robotic systems is under evaluation. Using existing technology, operational and descriptive software and hardware in the form of robotic end effectors are being designed in conjunction with interfacing cask components. A robotic radiation and contamination survey system has been developed and used on mock-up cask hardware to evaluate the impact of such fully automated operations on cask design features and productivity. Based on experience gained from the survey system, numerous health physics operations can be reliably performed with little human intervention using a fully automated system. Such operations can also significantly reduce time requirements for cask-receiving operations. 7 refs., 51 figs., 6 tabs.

  16. Human Thermal Model Evaluation Using the JSC Human Thermal Database

    NASA Technical Reports Server (NTRS)

    Cognata, T.; Bue, G.; Makinen, J.

    2011-01-01

    The human thermal database developed at the Johnson Space Center (JSC) is used to evaluate a set of widely used human thermal models. This database will facilitate a more accurate evaluation of human thermoregulatory response using in a variety of situations, including those situations that might otherwise prove too dangerous for actual testing--such as extreme hot or cold splashdown conditions. This set includes the Wissler human thermal model, a model that has been widely used to predict the human thermoregulatory response to a variety of cold and hot environments. These models are statistically compared to the current database, which contains experiments of human subjects primarily in air from a literature survey ranging between 1953 and 2004 and from a suited experiment recently performed by the authors, for a quantitative study of relative strength and predictive quality of the models. Human thermal modeling has considerable long term utility to human space flight. Such models provide a tool to predict crew survivability in support of vehicle design and to evaluate crew response in untested environments. It is to the benefit of any such model not only to collect relevant experimental data to correlate it against, but also to maintain an experimental standard or benchmark for future development in a readily and rapidly searchable and software accessible format. The Human thermal database project is intended to do just so; to collect relevant data from literature and experimentation and to store the data in a database structure for immediate and future use as a benchmark to judge human thermal models against, in identifying model strengths and weakness, to support model development and improve correlation, and to statistically quantify a model s predictive quality.

  17. Histopathological evaluation of tissue undergoing thermal insult

    PubMed Central

    Chaudhary, Minal; Bonde, Dushyant; Patil, Swati; Gawande, Madhuri; Hande, Alka; Jain, Deepali

    2016-01-01

    Context: Thermal insult is the major cause of thermal injury or death and in case of death due to thermal injury the body often has to be recovered from the site. Histologically, one can predict whether the victim was alive or dead when the fire was on going. However, determination of probable cause of thermal insult to which victim subjected to be difficult when the victim's body is found somewhere else from the crime scene or accident site or found alone. Hence, histopathological evaluation of the tissue which has undergone thermal insult in such conditions could help to place evidence in front of law officials, regarding probable condition, or scenario at time of burn of victim. Aims: Keeping this as a criteria in this study we aim to evaluate burnt tissue histopathologically, that undergone various degree of thermal insult, which simulates various real life scenario for mortality in burn cases. Settings and Design: We evaluate the changes in hematoxylin and eosin staining pattern of tissue which has undergone thermal insult compared to normal tissue and also the progressive changes in staining pattern, architectural, and cellular details. Materials and Methods: Samples were taken from the patients, in various surgical procedures. Each sample was cut into five parts with close margins so that each burnt tissue is evaluated for same field or region. The tissue that obtained was immediately subjected to varying degree of temperature over a specific period so as to simulate the various real-life condition. Then the tissues were fixed, processed, and stained with routine H and E staining. The processed slides of tissue were examined under the microscope, and the staining, and architectural changes were evaluated and described. Results: Results show that there was a progressive changes in the architectural pattern of the epithelium and connective tissue showing cleft formation and vacuolization, staining pattern also shows mixing of stains progressively as the

  18. Evaluation of New Thermally Conductive Geopolymer in Thermal Energy Storage

    NASA Astrophysics Data System (ADS)

    Černý, Matěj; Uhlík, Jan; Nosek, Jaroslav; Lachman, Vladimír; Hladký, Radim; Franěk, Jan; Brož, Milan

    This paper describes an evaluation of a newly developed thermally conductive geopolymer (TCG), consisting of a mixture of sodium silicate and carbon micro-particles. The TCG is intended to be used as a component of high temperature energy storage (HTTES) to improve its thermal diffusivity. Energy storage is crucial for both ecological and economical sustainability. HTTES plays a vital role in solar energy technologies and in waste heat recovery. The most advanced HTTES technologies are based on phase change materials or molten salts, but suffer with economic and technological limitations. Rock or concrete HTTES are cheaper, but they have low thermal conductivity without incorporation of TCG. It was observed that TCG is stable up to 400 °C. The thermal conductivity was measured in range of 20-23 W m-1 K-1. The effect of TCG was tested by heating a granite block with an artificial fissure. One half of the fissure was filled with TCG and the other with ballotini. 28 thermometers, 5 dilatometers and strain sensors were installed on the block. The heat transport experiment was evaluated with COMSOL Multiphysics software.

  19. Automated shielding analysis sequences for spent fuel casks

    SciTech Connect

    Tang, J.S.; Parks, C.V.; Hermann, O.W.

    1987-01-01

    Two important Shielding Analysis Sequences (SAS) have recently been developed within the SCALE computational system. These sequences significantly enhance the existing SCALE system capabilities for evaluating radiation doses exterior to spent fuel casks. These new control module sequences (SAS1 and SAS4) and their capabilities are discussed and demonstrated, together with the existing SAS2 sequence that is used to generate radiation sources for spent fuel. Particular attention is given to the new SAS4 sequence which provides an automated scheme for generating and using biasing parameters in a subsequent Monte Carlo analysis of a cask.

  20. GNS spent fuel cask experience

    SciTech Connect

    Weh, R. )

    1993-05-01

    The Gesellschaft fuer Nuklear-Service mbH (GNS), which is owned by German utilities, is responsible for the management of spent fuel and nuclear waste on behalf of the German utilities operating nuclear power plants. This paper describes the spent reactor fuel and waste shipping and/or storage casks that GNS manufacturers for nuclear facilities in Germany, and worldwide. So far more than 30 different casks have been produced in quantities ranging from one to several hundred of each type. GNS participates in the German Support Program to assist the International Atomic Energy Agency (IAEA) in developing verification procedures for dry storage casks containing spent fuel. This activity is also summarized.

  1. Computational Fluid Dynamics Best Practice Guidelines in the Analysis of Storage Dry Cask

    SciTech Connect

    Zigh, A.; Solis, J.

    2008-07-01

    Computational fluid dynamics (CFD) methods are used to evaluate the thermal performance of a dry cask under long term storage conditions in accordance with NUREG-1536 [NUREG-1536, 1997]. A three-dimensional CFD model was developed and validated using data for a ventilated storage cask (VSC-17) collected by Idaho National Laboratory (INL). The developed Fluent CFD model was validated to minimize the modeling and application uncertainties. To address modeling uncertainties, the paper focused on turbulence modeling of buoyancy driven air flow. Similarly, in the application uncertainties, the pressure boundary conditions used to model the air inlet and outlet vents were investigated and validated. Different turbulence models were used to reduce the modeling uncertainty in the CFD simulation of the air flow through the annular gap between the overpack and the multi-assembly sealed basket (MSB). Among the chosen turbulence models, the validation showed that the low Reynolds k-{epsilon} and the transitional k-{omega} turbulence models predicted the measured temperatures closely. To assess the impact of pressure boundary conditions used at the air inlet and outlet channels on the application uncertainties, a sensitivity analysis of operating density was undertaken. For convergence purposes, all available commercial CFD codes include the operating density in the pressure gradient term of the momentum equation. The validation showed that the correct operating density corresponds to the density evaluated at the air inlet condition of pressure and temperature. Next, the validated CFD method was used to predict the thermal performance of an existing dry cask storage system. The evaluation uses two distinct models: a three-dimensional and an axisymmetrical representation of the cask. In the 3-D model, porous media was used to model only the volume occupied by the rodded region that is surrounded by the BWR channel box. In the axisymmetric model, porous media was used to model

  2. Source storage and transfer cask: Users Guide

    SciTech Connect

    Eccleston, G.W.; Speir, L.G.; Garcia, D.C.

    1985-04-01

    The storage and shield cask for the dual californium source is designed to shield and transport up to 3.7 mg (2 Ci) of /sup 252/Cf. the cask meets Department of Transportation (DOT) license requirements for Type A materials (DOT-7A). The cask is designed to transfer sources to and from the Flourinel and Fuel Storage (FAST) facility delayed-neutron interrogator. Californium sources placed in the cask must be encapsulated in the SR-CF-100 package and attached to Teleflex cables. The cask contains two source locations. Each location contains a gear box that allows a Teleflex cable to be remotely moved by a hand crank into and out of the cask. This transfer procedure permits sources to be easily removed and inserted into the delayed-neutron interrogator and reduces personnel radiation exposure during transfer. The radiation dose rate with the maximum allowable quantity of californium (3.7 mg) in the cask is 30 mR/h at the surface and less than 2 mR/h 1 m from the cask surface. This manual contains information about the cask, californium sources, describes the method to ship the cask, and how to insert and remove sources from the cask. 28 figs.

  3. Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    SciTech Connect

    Daling, P.M.; Konzek, G.J.; Lezberg, A.J.; Votaw, E.F.; Collingham, M.I.

    1985-04-01

    This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%.

  4. Status of spent-fuel shipping cask development

    SciTech Connect

    Hall, I.K.; Hinschberger, T.S.

    1989-01-01

    The purpose of the Cask Systems Development Program is to develop a variety of cask systems that can safely and economically transport commercial spent fuel and high-level waste from the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. This paper is limited to a discussion of the status of from-reactor spent-fuel cask development; future cask development plans include MRS-to-repository casks, specialty casks for nonstandard spent fuel and nonfuel materials, and defense high-level waste casks. Spent-fuel casks must be available in the late 1990s to support the U.S. Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) shipments from utilities. DOE-Idaho, with the support of EG G Idaho, Inc., Sandia National Laboratories, and selected cask developing contractors, has been assigned the responsibility for developing a new generation of cask systems. Four categories of spent fuel casks were initially proposed: (1) legal weight truck (LWT) casks (2) overweight truck (OWT) casks (3) rail/barge (R/B) casks (4) dual purpose (DP) storage/transport casks. Casks are being designed for reduced occupational radiation exposure at the receiving facility by facilitating the use of remote handling equipment. Automation of remote handling systems may be used to reduce cask turnaround time. Reducing turnaround time promotes reduced radiation exposure to occupational workers and improves cask utilization efficiency.

  5. Transportation capabilities of the existing cask fleet

    SciTech Connect

    Johnson, P.E.; Joy, D.S.; Wankerl, M.W.

    1991-01-01

    This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 46 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the Department of Energy (DOE) consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated. 5 refs., 4 tabs.

  6. Used Fuel Cask Identification through Neutron Profile

    SciTech Connect

    Rauch, Eric Benton

    2015-11-20

    Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

  7. Parametric thermal evaluations of waste package emplacement

    SciTech Connect

    Bahney, R.H. III; Doering, T.W.

    1996-02-01

    Parametric thermal evaluations of spent nuclear fuel (SNF) waste packages (WPs) emplaced in the potential repository were performed to determine the impact of thermal loading, WP spacing, drift diameter, SNF aging, backfill, and relocation on the design of the Engineered Barrier System. Temperatures in the WP and near-field host rock are key to radionuclide containment, as they directly affect oxidation rates of the metal barriers and the ability of the rock to impede particle movement which must be demonstrated for a safe and licensable repository. Maximum allowable temperatures are based on material performance criteria and are specified as the following design goals for the WP/EBS design: SNF cladding 350{degrees}C, drift wall 200{degrees}C, and TSw3 rock 115{degrees}C.

  8. Evaluation of thermal overload in boiler operators.

    PubMed

    Braga, Camila Soares; Rodrigues, Valéria Antônia Justino; Campos, Julio César Costa; de Souza, Amaury Paulo; Minette, Luciano José; de Moraes, Angêlo Casali; Sensato, Guilherme Luciano

    2012-01-01

    The Brazilians educational institutions need a large energy demand for the operation of laundries, restaurants and accommodation of students. Much of that energy comes from steam generated in boilers with wood fuel. The laboral activity in boiler may present problems for the operator's health due to exposure to excessive heat, and its operation has a high degree of risk. This paper describes an analysis made the conditions of thermal environment in the operation of a B category boiler, located at a Higher Education Institution, located in the Zona da Mata Mineira The equipments used to collect data were Meter WBGT of the Heat Index; Meter of Wet Bulb Index and Globe Thermometer (WBGT); Politeste Instruments, an anemometer and an Infrared Thermometer. By the application of questionnaires, the second phase consisted of collecting data on environmental factors (temperature natural environment, globe temperature, relative humidity and air velocity). The study concluded that during the period evaluated, the activity had thermal overload. PMID:22316768

  9. Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

    DOE PAGESBeta

    Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; Scaglione, John M.

    2016-06-15

    We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

  10. THERMAL EVALUATION OF DIFFERENT DRIFT DIAMETER SIZES

    SciTech Connect

    H.M. Wade

    1999-01-04

    The purpose of this calculation is to estimate the thermal response of a repository-emplaced waste package and its corresponding drift wall surface temperature with different drift diameters. The case examined is that of a 21 pressurized water reactor (PWR) uncanistered fuel (UCF) waste package loaded with design basis spent nuclear fuel assemblies. This calculation evaluates a 3.5 meter to 6.5 meter drift diameter range in increments of 1.0 meters. The time-dependent temperatures of interest, as determined by this calculation, are the spent nuclear fuel cladding temperature, the waste package surface temperature, and the drift wall surface temperature.

  11. Results of Evaluation of Solar Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Woodcock, Gordon; Byers, Dave

    2003-01-01

    The solar thermal propulsion evaluation reported here relied on prior research for all information on solar thermal propulsion technology and performance. Sources included personal contacts with experts in the field in addition to published reports and papers. Mission performance models were created based on this information in order to estimate performance and mass characteristics of solar thermal propulsion systems. Mission analysis was performed for a set of reference missions to assess the capabilities and benefits of solar thermal propulsion in comparison with alternative in-space propulsion systems such as chemical and electric propulsion. Mission analysis included estimation of delta V requirements as well as payload capabilities for a range of missions. Launch requirements and costs, and integration into launch vehicles, were also considered. The mission set included representative robotic scientific missions, and potential future NASA human missions beyond low Earth orbit. Commercial communications satellite delivery missions were also included, because if STP technology were selected for that application, frequent use is implied and this would help amortize costs for technology advancement and systems development. A C3 Topper mission was defined, calling for a relatively small STP. The application is to augment the launch energy (C3) available from launch vehicles with their built-in upper stages. Payload masses were obtained from references where available. The communications satellite masses represent the range of payload capabilities for the Delta IV Medium and/or Atlas launch vehicle family. Results indicated that STP could improve payload capability over current systems, but that this advantage cannot be realized except in a few cases because of payload fairing volume limitations on current launch vehicles. It was also found that acquiring a more capable (existing) launch vehicle, rather than adding an STP stage, is the most economical in most cases.

  12. Shielding Analysis of the 5320 Shipping Cask

    SciTech Connect

    Blanchard, A.; Nathan, S.

    1998-05-01

    The purpose of this work is to demonstrate that the 5320 shipping cask meets Federal regulations for maximum radiation dose rates when loaded with the intended plutonium oxide cargo. It should be emphasized that the 5320 is an existing cask, and therefore this work represents confirmatory analysis rather than design analysis.

  13. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    SciTech Connect

    J. Bisset

    2005-02-14

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known.

  14. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  15. Design review report FFTF interim storage cask

    SciTech Connect

    Scott, P.L.

    1995-01-03

    Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location.

  16. Followup audit of the cask development program

    SciTech Connect

    Not Available

    1994-03-15

    The Department of Energy is responsible for developing a system for the transportation and storage of spent nuclear fuel generated by utility companies. To carry out this responsibility, the Department of Energy established the Office of Civilian Radioactive Waste Management (Waste Management Office). The Waste Management office began development of a series of new shipping casks to transport the spent fuel. The purpose of this audit was to review the current development status of the cask designs; compare the original milestone dates to current milestone dates; and review the program funds that have been used to date on the development of these casks. The Office of Inspector General audited the cask development program in 1987. The audit report (DOE/IG-0244), recommended that program management establish minimum criteria that each cask must meet to qualify for further development funding. Our followup audit found that this recommendation had not been adequately implemented. As a result, the Waste Management office will spend an estimated $143 million on the cask development program and receive only two cask designs that were originally scheduled to cost $26 million. Moreover, it is not certain, at this time, whether those two cask designs will eventually receive the Nuclear Regulatory Commission certification. Historically, the program has experienced slippage in milestone dates and steady increases in total cost. Management generally agreed with our current recommendations to establish formal contingency plans to counter further delays, develop current baselines and schedules in sufficient detail to adequately control cask development schedules and costs, and reevaluate the current status of the casks under development for the purpose of justifying further development. Management has proposed actions to correct the milestone date slippages and continued growth in the total cost of the program.

  17. 40 CFR 90.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 20 2011-07-01 2011-07-01 false Catalyst thermal stress resistance... Gaseous Exhaust Test Procedures § 90.427 Catalyst thermal stress resistance evaluation. (a) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress...

  18. 40 CFR 91.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 20 2011-07-01 2011-07-01 false Catalyst thermal stress resistance... Procedures § 91.427 Catalyst thermal stress resistance evaluation. (a)(1) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress on catalyst...

  19. 40 CFR 90.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Catalyst thermal stress resistance... Gaseous Exhaust Test Procedures § 90.427 Catalyst thermal stress resistance evaluation. (a) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress...

  20. 40 CFR 91.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Catalyst thermal stress resistance... Procedures § 91.427 Catalyst thermal stress resistance evaluation. (a)(1) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress on catalyst...

  1. 40 CFR 91.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 21 2012-07-01 2012-07-01 false Catalyst thermal stress resistance... Procedures § 91.427 Catalyst thermal stress resistance evaluation. (a)(1) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress on catalyst...

  2. 40 CFR 90.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 21 2013-07-01 2013-07-01 false Catalyst thermal stress resistance... Gaseous Exhaust Test Procedures § 90.427 Catalyst thermal stress resistance evaluation. (a) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress...

  3. 40 CFR 90.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 21 2012-07-01 2012-07-01 false Catalyst thermal stress resistance... Gaseous Exhaust Test Procedures § 90.427 Catalyst thermal stress resistance evaluation. (a) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress...

  4. 40 CFR 90.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 20 2014-07-01 2013-07-01 true Catalyst thermal stress resistance... Gaseous Exhaust Test Procedures § 90.427 Catalyst thermal stress resistance evaluation. (a) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress...

  5. 40 CFR 91.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 21 2013-07-01 2013-07-01 false Catalyst thermal stress resistance... Procedures § 91.427 Catalyst thermal stress resistance evaluation. (a)(1) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress on catalyst...

  6. 40 CFR 91.427 - Catalyst thermal stress resistance evaluation.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 20 2014-07-01 2013-07-01 true Catalyst thermal stress resistance... Procedures § 91.427 Catalyst thermal stress resistance evaluation. (a)(1) The purpose of the evaluation procedure specified in this section is to determine the effect of thermal stress on catalyst...

  7. FFTF disposable solid waste cask

    SciTech Connect

    Thomson, J. D.; Goetsch, S. D.

    1983-01-01

    Disposal of radioactive waste from the Fast Flux Test Facility (FFTF) will utilize a Disposable Solid Waste Cask (DSWC) for the transport and burial of irradiated stainless steel and inconel materials. Retrievability coupled with the desire for minimal facilities and labor costs at the disposal site identified the need for the DSWC. Design requirements for this system were patterned after Type B packages as outlined in 10 CFR 71 with a few exceptions based on site and payload requirements. A summary of the design basis, supporting analytical methods and fabrication practices developed to deploy the DSWC is provided in this paper.

  8. Thermal analysis of the IDENT 1578 fuel pin shipping container

    SciTech Connect

    Ingham, J.G.

    1980-01-01

    The IDENT 1578 container, which is a 110-in. long 5.5-in. OD tube, is designed for shipping FFTF fuel elements in T-3 casks between HEDL, HFEF, and other laboratories. The thermal analysis was conducted to evaluate whether or not the container satisfies its thermal design criteria (handle a decay heat load of 600 watts, max fuel pin cladding temperature not exceeding 800/sup 0/F).

  9. Thermal Protection Materials: Development, Characterization and Evaluation

    NASA Technical Reports Server (NTRS)

    Johnson, Silvia M.

    2012-01-01

    Thermal protection materials and systems (TPS) are used to protect space vehicles from the heat experienced during entry into an atmosphere. The application for these materials is very specialized as are the materials. They must have specific properties to withstand conditions during specific entries. There is no one-size-fits-all TPS as the conditions experienced by a material are very dependent upon the atmosphere, the entry speed, the size and shape of the vehicle, and the location on the vehicle. However, all TPS must be reliable and efficient to ensure mission safety, that is to protect the vehicle while ensuring that payload is maximized. Types of TPS will be reviewed in relation to types of missions and applications. Both reusable and ablative materials will be discussed. Approaches to characterizing and evaluating these materials will be presented. The role of heritage versus new materials will be described.

  10. 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).

  11. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    SciTech Connect

    Dilger, Fred; Halstead, Robert J.; Ballard, James D.

    2012-07-01

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's

  12. Use of depleted uranium metal as cask shielding in high-level waste storage, transport, and disposal systems

    SciTech Connect

    Yoshimura, H.R.; Ludwigsen, J.S.; McAllaster, M.E.

    1996-09-01

    The US DOE has amassed over 555,000 metric tons of depleted uranium from its uranium enrichment operations. Rather than dispose of this depleted uranium as waste, this study explores a beneficial use of depleted uranium as metal shielding in casks designed to contain canisters of vitrified high-level waste. Two high-level waste storage, transport, and disposal shielded cask systems are analyzed. The first system employs a shielded storage and disposal cask having a separate reusable transportation overpack. The second system employs a shielded combined storage, transport, and disposal cask. Conceptual cask designs that hold 1, 3, 4 and 7 high-level waste canisters are described for both systems. In all cases, cask design feasibility was established and analyses indicate that these casks meet applicable thermal, structural, shielding, and contact-handled requirements. Depleted uranium metal casting, fabrication, environmental, and radiation compatibility considerations are discussed and found to pose no serious implementation problems. About one-fourth of the depleted uranium inventory would be used to produce the casks required to store and dispose of the nearly 15,400 high-level waste canisters that would be produced. This study estimates the total-system cost for the preferred 7-canister storage and disposal configuration having a separate transportation overpack would be $6.3 billion. When credits are taken for depleted uranium disposal cost, a cost that would be avoided if depleted uranium were used as cask shielding material rather than disposed of as waste, total system net costs are between $3.8 billion and $5.5 billion.

  13. Drop tests and numerical impact analyses of new cask designs for High Activity Waste (Haw) and spent fuel - updated BAM design testing experiences

    SciTech Connect

    Volzke, H.; Zencker, U.; Qiao, L.; Feutlinske, K.; Musolff, A.

    2007-07-01

    In Germany, several new cask designs by international vendors (Gesellschaft fuer Nuklear Service mbH (GNS), TN International (TNI), Mitsubishi Heavy Industries (MHI)) are under design testing and within official licensing procedures for transport and storage casks for spent fuel and high activity waste (HAW). BAM (the German Federal Institute for Materials Research and Testing) has been performing several extensive drop test series with prototype casks to evaluate the safety margins against mechanical test conditions. An important project is the new GNS cask design for HAW, the CASTOR{sup R} HAW 28M. Sixteen drop tests have been performed under transport conditions with a 1:2 scale cask model equipped with impact limiters and extensively instrumented with strain gauges and accelerometers. Additionally, the accident scenario inside a storage facility has been investigated by a cask drop without impact limiters onto a nearly unyielding target. This scenario is dominated by highly dynamic effects and interactions between the test object and the target. Complete safety assessments for such mechanical accident scenarios and highly loaded cask structures require additional numerical investigations. They are done by complex finite element (FE) calculations that provide detailed dynamic stress and strain analyses all over the cask structure and at such points where sensors can't be applied. In addition, differences between the material property quantities of the prototype cask and the minimum material property requirements for the cask series production can be investigated as well as dimensional tolerances. By example, the safety assessment method and some of its special aspects are illustrated by the cask drop without an impact limiter onto a hard foundation. The main aspects and challenges are to develop a sufficient computer model of the cask and foundation and to provide detailed interpretation of the large amount of measurement data for achieving good correlation

  14. Evaluation of thermal data for geologic applications

    NASA Technical Reports Server (NTRS)

    Kahle, A. B.; Palluconi, F. D.; Levine, C. J.; Abrams, M. J.; Nash, D. B.; Alley, R. E.; Schieldge, J. P.

    1982-01-01

    Sensitivity studies using thermal models indicated sources of errors in the determination of thermal inertia from HCMM data. Apparent thermal inertia, with only simple atmospheric radiance corrections to the measured surface temperature, would be sufficient for most operational requirements for surface thermal inertia. Thermal data does have additional information about the nature of surface material that is not available in visible and near infrared reflectance data. Color composites of daytime temperature, nighttime temperature, and albedo were often more useful than thermal inertia images alone for discrimination of lithologic boundaries. A modeling study, using the annual heating cycle, indicated the feasibility of looking for geologic features buried under as much as a meter of alluvial material. The spatial resolution of HCMM data is a major limiting factor in the usefulness of the data for geologic applications. Future thermal infrared satellite sensors should provide spatial resolution comparable to that of the LANDSAT data.

  15. Ageing of a neutron shielding used in transport/storage casks

    SciTech Connect

    Nizeyiman, Fidele; Alami, Aatif; Issard, Herve; Bellenger, Veronique

    2012-07-11

    In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

  16. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    SciTech Connect

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs.

  17. Documentation for fiscal year 1995 annual BUSS cask SARP testing and inspections

    SciTech Connect

    Saueressig, P.T.

    1994-11-08

    The purpose of this report is to compile the data generated during the Fiscal Year (FY) 1995 annual tests and inspections performed on the Beneficial Uses Shipping System (BUSS) cask. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Section 8.2 ``Maintenance and Periodic Inspection Program`` of the BUSS Cask SARP requires that the following tests and inspections be performed on an annual basis: hydrostatic pressure test; helium leak test; dye penetrant test on the trunnions and life lugs; torque test on all permanent bolts; and impact limiter inspection and weight test. In addition to compiling the generated data, this report will verify that the testing criteria identified in section 8.2 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met.

  18. Cask system maintenance in the Federal Waste Management System

    SciTech Connect

    Pope, R.B.; Rennich, M.J.; Medley, L.G.; Attaway, C.R.

    1991-01-01

    In early 1988, in support of the development of the transportation system for the Office of Civilian Radioactive Waste Management System (OCRWM), a feasibility study was undertaken to define a the concept for a stand-alone, green-field'' facility for maintaining the Federal Waste Management System (FWMS) casks. This study provided and initial layout facility design, an estimate of the construction costs, and an acquisition schedule for a Cask Maintenance Facility (CMF). It also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrievable Storage (MRS) facility. The data, design, and estimated costs derived from the study have been organized for use in the total transportation system decision-making process. Most importantly, they also provide a foundation for continuing design and planning efforts. The feasibility study was based on an assumed stand-alone, green-field'' configuration. This design approach provides a comprehensive design evaluation, to guide the development of a cost estimate and to permit flexibility in locating the facility. The following sections provide background information on cask system maintenance, briefly summarizes some of the functional requirements that a CMF must satisfy, provides a physical description of the CMF, briefly discusses the cost and schedule estimates and then reviews the findings of the efforts undertaken since the feasibility study was completed. 15 refs., 3 figs.

  19. Radioactive materials shipping cask anticontamination enclosure

    DOEpatents

    Belmonte, Mark S.; Davis, James H.; Williams, David A.

    1982-01-01

    An anticontamination device for use in storing shipping casks for radioactive materials comprising (1) a seal plate assembly; (2) a double-layer plastic bag; and (3) a water management system or means for water management.

  20. 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.

  1. Objective evaluation of cutaneous thermal sensivity

    NASA Technical Reports Server (NTRS)

    Vanbeaumont, W.

    1972-01-01

    The possibility of obtaining reliable and objective quantitative responses was investigated under conditions where only temperature changes in localized cutaneous areas evoked measurable changes in remote sudomotor activity. Both male and female subjects were studied to evaluate sex difference in thermal sensitivity. The results discussed include: sweat rate responses to contralateral cooling, comparison of sweat rate responses between men and women to contralateral cooling, influence of the menstrual cycle on the sweat rate responses to contralateral cooling, comparison of threshold of sweating responses between men and women, and correlation of latency to threshold for whole body sweating. It is concluded that the quantitative aspects of the reflex response is affected by both the density and activation of receptors as well as the rate of heat loss; men responded 8-10% more frequently than women to thermode cooling, the magnitude of responses being greater for men; and women responded 7-9% more frequently to thermode cooling on day 1 of menstruation, as compared to day 15.

  2. Thermal fatigue of composites: Ultrasonic and SEM evaluations

    SciTech Connect

    Forsyth, D.S.; Kasap, S.O. . Dept. of Electrical Engineering); Wacker, I.; Yannacopoulos, S. . Dept. of Mechanical Engineering)

    1994-01-01

    Results are presented on the evaluation of thermal fatigue in three fiber reinforced polymer composites, using ultrasonic techniques and scanning electron microscopy. The composites examined were (a) continuous carbon fibers in a vinylester matrix (b) continuous aramid fibers in a vinylester matrix and (c) randomly oriented aramid fibers in a polyphenylene matrix. Specimens of these composites were subjected to thermal fatigue by thermal cycling from [minus]25 C to 75 C. Changes in ultrasonic attenuation and velocity were monitored during thermal cycling, and scanning electron microscopy was used to qualitatively evaluate any damage. It was observed that ultrasonic attenuation is sensitive to thermal fatigue, increasing with increasing number of thermal cycles. SEM evaluations showed that the primary damage due to thermal fatigue is due to fiber-matrix debonding.

  3. Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination

    SciTech Connect

    Bare, Walter Claude; Ebner, Matthias Anthony; Torgerson, Laurence Dale

    2001-08-01

    This report documents visual examination and testing conducted in 1999 and early 2000 at the Idaho National Engineering and Environmental Laboratory (INEEL) on a Gesellschaft für Nuklear Service (GNS) CASTOR V/21 pressurized water reactor (PWR) spent fuel dry storage cask. The purpose of the examination and testing is to develop a technical basis for renewal of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level waste at independent spent fuel storage installation sites. The examination and testing was conducted to assess the condition of the cask internal and external surfaces, cask contents consisting of 21 Westinghouse PWR spent fuel assemblies from Dominion’s (formerly named Virginia Power) Surry Power Station and cask concrete storage pad. The assemblies have been continuously stored in the CASTOR cask since 1985. Cask exterior surface and selected fuel assembly temperatures, and cask surface gamma and neutron dose rates were measured. Cask external/internal surfaces, fuel basket components including accessible weldments, fuel assembly exteriors, and primary lid seals were visually examined. Selected fuel rods were removed from one fuel assembly, visually examined, and then shipped to Argonne National Laboratory for nondestructive, destructive, and mechanical examination. Cask interior crud samples and helium cover gas samples were collected and analyzed. The results of the examination and testing indicate the concrete storage pad, CASTOR V/21 cask, and cask contents exhibited sound structural and seal integrity and that long-term storage has not caused detectable degradation of the spent fuel cladding or the release of gaseous fission products between 1985 and 1999.

  4. Development, fabrication and evaluation of composite thermal engine insulation

    NASA Technical Reports Server (NTRS)

    1973-01-01

    Foil enclosure configurations of 10 variations were fabricated and evaluated. A discussion of the thermal protection system panel design includes: (1) description of 3DSX/foil concept, (2) design environment, (3) material selection, (4) fabrication enclosure, (5) structural design, (6) thermal sizing, and (7) weight analysis. The structural design study includes foil evaluation, venting pressure loads, thermomechanical behavior, and enclosure venting (burst) pressure tests. Results of experimental demonstrations of performance and reuse capabilities are given for both thermal and acoustic testing.

  5. DRACS thermal performance evaluation for FHR

    SciTech Connect

    Lv, Q.; Lin, H. C.; Kim, I. H.; Sun, X.; Christensen, R. N.; Blue, T. E.; Yoder, G. L.; Wilson, D. F.; Sabharwall, P.

    2015-03-01

    Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger and the Natural Draft Heat Exchanger. In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, and to keep the DRACS ready for activation, if needed, during accidents. To help with the design and thermal performance evaluation of the DRACS, a computer code using MATLAB has been developed. This code is based on a one-dimensional formulation and its principle is to solve the energy balance and integral momentum equations. By discretizing the DRACS system in the axial direction, a bulk mean temperature is assumed for each mesh cell. The temperatures of all the cells, as well as the mass flow rates in the DRACS loops, are predicted by solving the governing equations that are obtained by integrating the energy conservation equation over each cell and integrating the momentum conservation equation over each of the DRACS loops. In addition, an intermediate heat transfer loop equipped with a pump has also been modeled in the code. This enables the study of flow reversal phenomenon in the DRACS primary loop, associated with the pump trip process. Experimental data from a High-Temperature DRACS Test Facility (HTDF) are not available yet to benchmark the code. A preliminary code validation is performed by using natural circulation experimental data available in the literature, which are as closely relevant as possible. The code is subsequently applied to the HTDF that is under

  6. Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

    SciTech Connect

    Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

    1986-12-01

    This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions.

  7. Systems evaluation of thermal bus concepts

    NASA Technical Reports Server (NTRS)

    Stalmach, D. D.

    1982-01-01

    Thermal bus concepts, to provide a centralized thermal utility for large, multihundred kilowatt space platforms, were studied and the results are summarized. Concepts were generated, defined, and screened for inclusion in system level thermal bus trades. Parametric trade studies were conducted in order to define the operational envelope, performance, and physical characteristics of each. Two concepts were selected as offering the most promise for thermal bus development. All of four concepts involved two phase flow in order to meet the required isothermal nature of the thermal bus. Two of the concepts employ a mechanical means to circulate the working fluid, a liquid pump in one case and a vapor compressor in another. Another concept utilizes direct osmosis as the driving force of the thermal bus. The fourth concept was a high capacity monogroove heat pipe. After preliminary sizing and screening, three of these concepts were selected to carry into the trade studies. The monogroove heat pipe concept was deemed unsuitable for further consideration because of its heat transport limitations. One additional concept utilizing capillary forces to drive the working fluid was added. Parametric system level trade studies were performed. Sizing and weight calculations were performed for thermal bus sizes ranging from 5 to 350 kW and operating temperatures in the range of 4 to 120 C. System level considerations such as heat rejection and electrical power penalties and interface temperature losses were included in the weight calculations.

  8. Interim storage cask (ISC), a concrete and steel dry storage cask

    SciTech Connect

    Grenier, R.M.; Koploy, M.A.

    1995-12-31

    General Atomics (GA) has designed and is currently fabricating the Interim Storage Cask (ISC) for Westinghouse Hanford Company (WHC). The ISC is a dry storage cask that will safely store a Core Component Container (CCC) with Fast Flux Test Facility (FFTF) spent fuel assemblies or fuel pin containers for a period of up to 50 years at the US Department of Energy (DOE) Hanford site. The cask may also be used to transfer the fuel to different areas within the Hanford site. The ISC is designed to stringent criteria from both 10CFR71 and 10CFR72 for safe storage and on-site transportation of FFTF spent fuel and fuel pin containers. The cask design uses a combination of steel and concrete materials to achieve a cost-effective means of storing spent fuel. The casks will be extensively tested before use to verify that the design and construction meet the design requirements.

  9. The impact of using reduced capacity baskets on cask fleet size and cask fleet mix

    SciTech Connect

    Joy, D.S.; Johnson, P.E.; Andress, D.A.

    1993-06-01

    The Civilian Radioactive Waste Management System transportation system will encounter a wide range of spent fuel characteristics. Since the Initiative I casks are being designed to transport 10-year-old fuel with a burnup of 35,000 MWd/MTU, there is a good likelihood that a number of the cask shipments will need to be derated in order to meet the Nuclear Regulatory Commission radiation guidelines. This report discusses the impact of cask derating by using reduced-capacity baskets. Cask derating, while enhancing the ability to move spent fuel with a wider range of age and burnup characteristics, increases the number of shipments; the amount of equipment (cask bodies, baskets, etc.); and the number of visits to both shipping and receiving sites required to transport a specific amount of spent fuel.

  10. Full analytical evaluation of thermal transport properties of nanomaterials

    NASA Astrophysics Data System (ADS)

    Mamedov, B. A.

    2016-02-01

    New approaches for the analytical evaluation of the heat capacities and thermal conductivity of nanowires are presented. The most significant result of our calculation is an explicit closed form in terms of elementary functions. This allows the specific heat and thermal conductivity of nanowires to be easily evaluated within the arbitrary values of parameters. The proposed method is applied successfully to the evaluation of the heat capacities and thermal conductivity of Ni nanowire and can be used as a universal heat capacity evaluation scheme for all nanowires and other nanostructures. The theoretical model has been verified by comparing the predicted results with those obtained from the available analytical and literature data.

  11. Performance evaluation of floor thermal storage system

    SciTech Connect

    Shinkai, Koichiro; Kasuya, Atsushi; Kato, Masahiro

    2000-07-01

    Environmental issues were seriously addressed when a new building was designed with district heating and cooling for the Osaka gas company. As a result, the building was officially recognized as Environmentally Conscious Building No. 1 by the Construction Ministry. In order to reduce cost by peak shaving, adoption of a floor thermal storage system was planned. This paper describes results regarding the peak shaving by floor thermal storage system in designing the air-conditioning system.

  12. Genetics Home Reference: CASK-related intellectual disability

    MedlinePlus

    ... XL-ID with or without nystagmus (rapid, involuntary eye movements) is a milder form of CASK -related intellectual ... to promote development of the nerves that control eye movement (the oculomotor neural network). Mutations in the CASK ...

  13. Truncated CASK does not alter skeletal muscle or protein interactors.

    PubMed

    Sanford, Jamie L; Mays, Tessily A; Varian, Kenneth D; Wilson, Joanna B; Janssen, Paul M L; Rafael-Fortney, Jill A

    2008-09-01

    CASK (Ca2+, calmodulin-associated serine/threonine kinase) is an essential mammalian cell junction protein and is also crucial at Drosophila neuromuscular synapses. We have shown that CASK is present in mammalian skeletal muscle at the postsynaptic membrane of the neuromuscular junction. CASK interacts biochemically with channels at central synapses, and studies in cultured cells have led to proposed functions for CASK. However, in vivo functions of CASK in skeletal muscle remain unknown. To test hypotheses of CASK functions, we generated two lines of transgenic mice, which overexpress full-length and truncated CASK protein in skeletal muscle. Extensive analyses showed that overexpression of CASK protein did not affect the morphology or physiology of skeletal muscle, the morphology of the neuromuscular junction, or the levels or distribution of protein interactors. These results contrast with previous cell culture experiments and emphasize the importance of in vivo analysis of protein function. PMID:18642383

  14. Safety analysis report vitrified high level waste type B shipping cask

    SciTech Connect

    1995-03-01

    This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

  15. MODELING HEAT TRANSFER IN SPENT FUEL TRANSFER CASK NEUTRON SHIELDS – A CHALLENGING PROBLEM IN NATURAL CONVECTION

    SciTech Connect

    Fort, James A.; Cuta, Judith M.; Bajwa, C.; Baglietto, E.

    2010-07-18

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10-15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not

  16. Test facilities for evaluating nuclear thermal propulsion systems

    SciTech Connect

    Beck, D.F.; Allen, G.C.; Shipers, L.R.; Dobranich, D.; Ottinger, C.A.; Harmon, C.D.; Fan, W.C. ); Todosow, M. )

    1992-09-22

    Interagency panels evaluating nuclear thermal propulsion (NTP) development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and baseline performance of some of the major subsystems designed to support a proposed ground test complex for evaluating nuclear thermal propulsion fuel elements and engines being developed for the Space Nuclear Thermal Propulsion (SNTP) program. Some preliminary results of evaluating this facility for use in testing other NTP concepts are also summarized.

  17. SUMOylation of the MAGUK protein CASK regulates dendritic spinogenesis.

    PubMed

    Chao, Hsu-Wen; Hong, Chen-Jei; Huang, Tzyy-Nan; Lin, Yi-Ling; Hsueh, Yi-Ping

    2008-07-14

    Membrane-associated guanylate kinase (MAGUK) proteins interact with several synaptogenesis-triggering adhesion molecules. However, direct evidence for the involvement of MAGUK proteins in synapse formation is lacking. In this study, we investigate the function of calcium/calmodulin-dependent serine protein kinase (CASK), a MAGUK protein, in dendritic spine formation by RNA interference. Knockdown of CASK in cultured hippocampal neurons reduces spine density and shrinks dendritic spines. Our analysis of the time course of RNA interference and CASK overexpression experiments further suggests that CASK stabilizes or maintains spine morphology. Experiments using only the CASK PDZ domain or a mutant lacking the protein 4.1-binding site indicate an involvement of CASK in linking transmembrane adhesion molecules and the actin cytoskeleton. We also find that CASK is SUMOylated. Conjugation of small ubiquitin-like modifier 1 (SUMO1) to CASK reduces the interaction between CASK and protein 4.1. Overexpression of a CASK-SUMO1 fusion construct, which mimicks CASK SUMOylation, impairs spine formation. Our study suggests that CASK contributes to spinogenesis and that this is controlled by SUMOylation. PMID:18606847

  18. Evaluation of thermal cameras for non-destructive thermal testing applications

    NASA Astrophysics Data System (ADS)

    Chrzanowski, K.; Park, S. N.

    2001-04-01

    Thermal cameras are nowadays often used in industry and science for non-destructive thermal testing (NDTT). There have been published, by the American Society for Testing of Materials, two standards that present detailed measurement procedures of the minimum resolvable temperature difference (MRTD) and the minimum detectable temperature difference (MDTD) of commercial thermal cameras for NDTT applications. However, the standards provide only very general guidelines about the use of the measured MRTD and MDTD values for evaluation of thermal cameras for NDTT applications. Precise methods that enable evaluation of a thermal imager for NDTT application on the basis of measurement results of the MRTD and the MDTD are presented in this paper. The methods enable estimation of probabilities of detection, orientation, recognition and identification of thermal anomalies generated by flaws in the materials imaged.

  19. Structural design of concrete storage pads for spent-fuel casks

    SciTech Connect

    Rashid, Y.R.; Nickell, R.E.; James, R.J. )

    1993-04-01

    The loading experienced by spent fuel dry storage casks and storage pads due to potential drop or tip-over accidents is evaluated using state-of-the-art concrete structural analysis methodology. The purpose of this analysis is to provide simple design charts and formulas so that design adequacy of storage pads and dry storage casks can be demonstrated. The analysis covers a wide range of slab-design parameters, e.g., reinforcement ratio, slab thickness, concrete compressive strength, and sub-base soil compaction, as well as variations in drop orientation and drop height. The results are presented in the form of curves, giving the force on the cask as a function of storage pad hardness for various drop heights. In addition, force-displacement curves, deformed shapes, crack patterns, stresses and strains are given for various slab-design conditions and drop events. The utility of the results in design are illustrated through examples.

  20. Evaluation of Students' Understanding of Thermal Concepts in Everyday Contexts

    ERIC Educational Resources Information Center

    Chu, Hye-Eun; Treagust, David F.; Yeo, Shelley; Zadnik, Marjan

    2012-01-01

    The aims of this study were to determine the underlying conceptual structure of the thermal concept evaluation (TCE) questionnaire, a pencil-and-paper instrument about everyday contexts of heat, temperature, and heat transfer, to investigate students' conceptual understanding of thermal concepts in everyday contexts across several school years and…

  1. MSFC Skylab thermal and environmental control system mission evaluation

    NASA Technical Reports Server (NTRS)

    Hopson, G. D.; Littles, J. W.; Patterson, W. C.

    1974-01-01

    An evaluation of the performance of the Skylab thermal and environmental control system is presented. Actual performance is compared to design and functional requirements and anomalies and discrepancies and their resolution are discussed. The thermal and environmental control systems performed their intended role. Based on the experience gained in design, development and flight, recommendations are provided which may be beneficial to future system designs.

  2. Safety analysis report for packaging (onsite) multicanister overpack cask

    SciTech Connect

    Edwards, W.S.

    1997-07-14

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  3. An Approach of Uncertainty Evaluation for Thermal-Hydraulic Analysis

    SciTech Connect

    Katsunori Ogura; Hisashi Ninokata

    2002-07-01

    An approach to evaluate uncertainty systematically for thermal-hydraulic analysis programs is demonstrated. The approach is applied to the Peach Bottom Unit 2 Turbine Trip 2 Benchmark and is validated. (authors)

  4. A GAMMA RAY SCANNING APPROACH TO QUANTIFY SPENT FUEL CASK RADIONUCLIDE CONTENTS

    SciTech Connect

    Branney, S.

    2011-07-01

    The International Atomic Energy Agency (IAEA) has outlined a need to develop methods of allowing re-verification of LWR spent fuel stored in dry storage casks without the need of a reference baseline measurement. Some scanning methods have been developed, but improvements can be made to readily provide required data for spent fuel cask verification. The scanning process should be conditioned to both confirm the contents and detect any changes due to container/contents degradation or unauthorized removal or tampering. Savannah River National Laboratory and The University of Tennessee are exploring a new method of engineering a high efficiency, cost effective detection system, capable of meeting the above defined requirements in a variety of environmental situations. An array of NaI(Tl) detectors, arranged to form a 'line scan' along with a matching array of 'honeycomb' collimators provide a precisely defined field of view with minimal degradation of intrinsic detection efficiency and with significant scatter rejection. Scanning methods are adapted to net optimum detection efficiency of the combined system. In this work, and with differing detectors, a series of experimental demonstrations are performed that map system spatial performance and counting capability before actual spent fuel cask scans are performed. The data are evaluated to demonstrate the prompt ability to identify missing fuel rods or other content abnormalities. To also record and assess cask tampering, the cask is externally examined utilizing FTIR hyper spectral and other imaging/sensing approaches. This provides dated records and indications of external abnormalities (surface deposits, smears, contaminants, corrosion) attributable to normal degradation or to tampering. This paper will describe the actual gathering of data in both an experimental climate and from an actual spent fuel dry storage cask, and how an evaluation may be performed by an IAEA facility inspector attempting to draw an

  5. Thermal decomposition hazard evaluation of hydroxylamine nitrate.

    PubMed

    Wei, Chunyang; Rogers, William J; Mannan, M Sam

    2006-03-17

    Hydroxylamine nitrate (HAN) is an important member of the hydroxylamine family and it is a liquid propellant when combined with alkylammonium nitrate fuel in an aqueous solution. Low concentrations of HAN are used primarily in the nuclear industry as a reductant in nuclear material processing and for decontamination of equipment. Also, HAN has been involved in several incidents because of its instability and autocatalytic decomposition behavior. This paper presents calorimetric measurement for the thermal decomposition of 24 mass% HAN/water. Gas phase enthalpy of formation of HAN is calculated using both semi-empirical methods with MOPAC and high-level quantum chemical methods of Gaussian 03. CHETAH is used to estimate the energy release potential of HAN. A Reactive System Screening Tool (RSST) and an Automatic Pressure Tracking Adiabatic Calorimeter (APTAC) are used to characterize thermal decomposition of HAN and to provide guidance about safe conditions for handling and storing of HAN. PMID:16154263

  6. Evaluation of thermal-storage concepts for solar cooling applications

    NASA Astrophysics Data System (ADS)

    Hughes, P. J.; Morehouse, J. H.; Choi, M. K.; White, N. M.; Scholten, W. B.

    1981-10-01

    Various configuration concepts for utilizing thermal energy storage to improve the thermal and economic performance of solar cooling systems for buildings were analyzed. The storge concepts evaluated provide short-term thermal storge via the bulk containment of water or salt hydrates. The evaluations were made for both residential-size cooling systems (3-ton) and small commercial-size cooling systems (25-ton). The residential analysis considers energy requirements for space heating, space cooling and water heating, while the commercial building analysis is based only on energy requirements for space cooling. The commercial building analysis considered a total of 10 different thermal storage/solar systems, 5 each for absorption and Rankine chiller concepts. The residential analysis considered 4 thermal storage/solar systems, all utilizing an absorption chiller. The trade-offs considered include: cold-side versus hot-side storage, single vs multiple stage storage, and phase-change vs sensible heat storage.

  7. SUMOylation of the MAGUK protein CASK regulates dendritic spinogenesis

    PubMed Central

    Chao, Hsu-Wen; Hong, Chen-Jei; Huang, Tzyy-Nan; Lin, Yi-Ling; Hsueh, Yi-Ping

    2008-01-01

    Membrane-associated guanylate kinase (MAGUK) proteins interact with several synaptogenesis-triggering adhesion molecules. However, direct evidence for the involvement of MAGUK proteins in synapse formation is lacking. In this study, we investigate the function of calcium/calmodulin-dependent serine protein kinase (CASK), a MAGUK protein, in dendritic spine formation by RNA interference. Knockdown of CASK in cultured hippocampal neurons reduces spine density and shrinks dendritic spines. Our analysis of the time course of RNA interference and CASK overexpression experiments further suggests that CASK stabilizes or maintains spine morphology. Experiments using only the CASK PDZ domain or a mutant lacking the protein 4.1–binding site indicate an involvement of CASK in linking transmembrane adhesion molecules and the actin cytoskeleton. We also find that CASK is SUMOylated. Conjugation of small ubiquitin-like modifier 1 (SUMO1) to CASK reduces the interaction between CASK and protein 4.1. Overexpression of a CASK–SUMO1 fusion construct, which mimicks CASK SUMOylation, impairs spine formation. Our study suggests that CASK contributes to spinogenesis and that this is controlled by SUMOylation. PMID:18606847

  8. Research on Spent Fuel Storage and Transportation in CRIEPI (Part 2 Concrete Cask Storage)

    SciTech Connect

    Koji Shirai; Jyunichi Tani; Taku Arai; Masumi Watatu; Hirofumi Takeda; Toshiari Saegusa; Philip L. Winston

    2008-10-01

    Concrete cask storage has been implemented in the world. At a later stage of storage period, the containment of the canister may deteriorate due to stress corrosion cracking phenomena in a salty air environment. High resistant stainless steels against SCC have been tested as compared with normal stainless steel. Taking account of the limited time-length of environment with certain level of humidity and temperature range, the high resistant stainless steels will survive from SCC damage. In addition, the adhesion of salt from salty environment on the canister surface will be further limited with respect to the canister temperature and angle of the canister surface against the salty air flow in the concrete cask. Optional countermeasure against SCC with respect to salty air environment has been studied. Devices consisting of various water trays to trap salty particles from the salty air were designed to be attached at the air inlet for natural cooling of the cask storage building. Efficiency for trapping salty particles was evaluated. Inspection of canister surface was carried out using an optical camera inserted from the air outlet through the annulus of a concrete cask that has stored real spent fuel for more than 15 years. The camera image revealed no gross degradation on the surface of the canister. Seismic response of a full-scale concrete cask with simulated spent fuel assemblies has been demonstrated. The cask did not tip over, but laterally moved by the earthquake motion. Stress generated on the surface of the spent fuel assemblies during the earthquake motion were within the elastic region.

  9. Transportation accident response of a high-capacity truck cask for spent fuel

    SciTech Connect

    O`Connell, W.J.; Glaser, R.E.; Johnson, G.L.; Perfect, S.A.; McGuinn, E.J.; Lake, W.H.

    1995-11-01

    Two of the primary goals of this study were (i) to check the structural and thermal performance of the GA-4 cask in a broad range of accidents and (ii) to carry out a severe-accidents analysis as had been addressed in the Modal Study but now using a specific recent cask design and using current-generation computer models and capabilities. At the same time, it was desired to compare the accident performance of the Ga-4 cask to that of the generic truck cask analyzed in the Modal Study. The same range of impact and fire accidents developed in the Modal Study was adopted for this study. The accident-description data base of the Modal Study categorizes accidents into types of collisions with mobile or fixed objects, non-collision accidents, and fires. The mechanical modes of damage may be via crushing, impact, or puncture. The fire occurrences in the Modal Study data are based on truck accident statistics. The fire types are taken to be pool fires of petroleum products from fuel tanks and/or cargoes.

  10. A 2-D Test Problem for CFD Modeling Heat Transfer in Spent Fuel Transfer Cask Neutron Shields

    SciTech Connect

    Zigh, Ghani; Solis, Jorge; Fort, James A.

    2011-01-14

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 5-10 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper presents results for a simple 2-D problem that is an effective numerical analog for the neutron shield application. Because it is 2-D, solutions can be obtained relatively quickly allowing a comparison and assessment of sensitivity to model parameter changes. Turbulence models are considered as

  11. Engineering evaluation of a sodium hydroxide thermal energy storage module

    NASA Technical Reports Server (NTRS)

    Perdue, D. G.; Gordon, L. H.

    1980-01-01

    An engineering evaluation of thermal energy storage prototypes was performed in order to assess the development status of latent heat storage media. The testing and the evaluation of a prototype sodium hydroxide module is described. This module stored off-peak electrical energy as heat for later conversion to domestic hot water needs.

  12. Testing and evaluation of thermal cameras for absolute temperature measurement

    NASA Astrophysics Data System (ADS)

    Chrzanowski, Krzysztof; Fischer, Joachim; Matyszkiel, Robert

    2000-09-01

    The accuracy of temperature measurement is the most important criterion for the evaluation of thermal cameras used in applications requiring absolute temperature measurement. All the main international metrological organizations currently propose a parameter called uncertainty as a measure of measurement accuracy. We propose a set of parameters for the characterization of thermal measurement cameras. It is shown that if these parameters are known, then it is possible to determine the uncertainty of temperature measurement due to only the internal errors of these cameras. Values of this uncertainty can be used as an objective criterion for comparisons of different thermal measurement cameras.

  13. Cognitive Appraisals Affect Both Embodiment of Thermal Sensation and Its Mapping to Thermal Evaluation

    PubMed Central

    Keeling, Trevor P.; Roesch, Etienne B.; Clements-Croome, Derek

    2016-01-01

    The physical environment leads to a thermal sensation that is perceived and appraised by occupants. The present study focuses on the relationship between sensation and evaluation. We asked 166 people to recall a thermal event from their recent past. They were then asked how they evaluated this experience in terms of 10 different emotions (frustrated, resigned, dislike, indifferent, angry, anxious, liking, joyful, regretful, proud). We tested whether four psychological factors (appraisal dimensions) could be used to predict the ensuing emotions, as well as comfort, acceptability, and sensation. The four dimensions were: the Conduciveness of the event, who/what caused the event (Causality), who had control (Agency), and whether the event was expected (Expectations). These dimensions, except for Expectations, were good predictors of the reported emotions. Expectations, however, predicted the reported thermal sensation, its acceptability, and ensuing comfort. The more expected an event was, the more uncomfortable a person felt, and the less likely they reported a neutral thermal sensation. Together, these results support an embodied view of how subjective appraisals affect thermal experience. Overall, we show that appraisal dimensions mediate occupants' evaluation of their thermal sensation, which suggests an additional method for understanding psychological adaption. PMID:27445877

  14. Cognitive Appraisals Affect Both Embodiment of Thermal Sensation and Its Mapping to Thermal Evaluation.

    PubMed

    Keeling, Trevor P; Roesch, Etienne B; Clements-Croome, Derek

    2016-01-01

    The physical environment leads to a thermal sensation that is perceived and appraised by occupants. The present study focuses on the relationship between sensation and evaluation. We asked 166 people to recall a thermal event from their recent past. They were then asked how they evaluated this experience in terms of 10 different emotions (frustrated, resigned, dislike, indifferent, angry, anxious, liking, joyful, regretful, proud). We tested whether four psychological factors (appraisal dimensions) could be used to predict the ensuing emotions, as well as comfort, acceptability, and sensation. The four dimensions were: the Conduciveness of the event, who/what caused the event (Causality), who had control (Agency), and whether the event was expected (Expectations). These dimensions, except for Expectations, were good predictors of the reported emotions. Expectations, however, predicted the reported thermal sensation, its acceptability, and ensuing comfort. The more expected an event was, the more uncomfortable a person felt, and the less likely they reported a neutral thermal sensation. Together, these results support an embodied view of how subjective appraisals affect thermal experience. Overall, we show that appraisal dimensions mediate occupants' evaluation of their thermal sensation, which suggests an additional method for understanding psychological adaption. PMID:27445877

  15. An assessment methodology for thermal energy storage evaluation

    NASA Astrophysics Data System (ADS)

    Brown, D. R.; Dirks, J. A.; Drost, M. K.; Spanner, G. E.; Williams, T. A.

    1987-11-01

    This report documents an assessment methodology for evaluating the cost, performance, and overall economic feasibility of thermal energy storage (TES) concepts. The methodology was developed by Thermal Energy Storage Evaluation Program personnel at Pacific Northwest Laboratory (PNL) for use by PNL and other TES concept evaluators. The methodology is generically applicable to all TES concepts; however, specific analyses may require additional or more detailed definition of the ground rules, assumptions, and analytical approach. The overall objective of the assessment methodology is to assist in preparing equitable and proper evaluations of TES concepts that will allow developers and end-users to make valid decisions about research and development (R and D) and implementation. The methodology meets this objective by establishing standard approaches, ground rules, assumptions, and definitions that are analytically correct and can be consistently applied by concept evaluators.

  16. An assessment methodology for thermal energy storage evaluation

    SciTech Connect

    Brown, D.R.; Dirks, J.A.; Drost, M.K.; Spanner, G.E.; Williams, T.A.

    1987-11-01

    This report documents an assessment methodology for evaluating the cost, performance, and overall economic feasibility of thermal energy storage (TES) concepts. The methodology was developed by Thermal Energy Storage Evaluation Program personnel at Pacific Northwest Laboratory (PNL) for use by PNL and other TES concept evaluators. The methodology is generically applicable to all TES concepts; however, specific analyses may require additional or more detailed definition of the ground rules, assumptions, and analytical approach. The overall objective of the assessment methodology is to assist in preparing equitable and proper evaluations of TES concepts that will allow developers and end-users to make valid decisions about research and development (R and D) and implementation. The methodology meets this objective by establishing standard approaches, ground rules, assumptions, and definitions that are analytically correct and can be consistently applied by concept evaluators. 15 refs., 4 figs., 13 tabs.

  17. Exploratory evaluation of ceramics for automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1972-01-01

    An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

  18. Impact Analyses and Tests of Metal Cask Considering Aircraft Engine Crash - 12308

    SciTech Connect

    Lee, Sanghoon; Choi, Woo-Seok; Kim, Ki-Young; Jeon, Je-Eon; Seo, Ki-Seog

    2012-07-01

    The structural integrity of a dual purpose metal cask currently under development by the Korea Radioactive Waste Management Cooperation (KRMC) is evaluated through analyses and tests under a high-speed missile impact considering the targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from the literature. The missile impact velocity was set at 150 m/s, and two impact orientations were considered. A simplified missile simulating a commercial aircraft engine is designed from an impact load history curve provided in the literature. In the analyses, the focus is on the evaluation of the containment boundary integrity of the metal cask. The analyses results are compared with the results of tests using a 1/3 scale model. The results show very good agreements, and the procedure and methodology adopted in the structural analyses are validated. While the integrity of the cask is maintained in one evaluation where the missile impacts the top side of the free standing cask, the containment boundary is breached in another case in which the missile impacts the center of the cask lid in a perpendicular orientation. A safety assessment using a numerical simulation of an aircraft engine crash into spent nuclear fuel storage systems is performed. A commercially available explicit finite element code is utilized for the dynamic simulation, and the strain rate effect is included in the modeling of the materials used in the target system and missile. The simulation results show very good agreement with the test results. It is noted that this is the first test considering an aircraft crash in Korea. (authors)

  19. 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).

  20. Development and evaluation of thermal model reduction algorithms for spacecraft

    NASA Astrophysics Data System (ADS)

    Deiml, Michael; Suderland, Martin; Reiss, Philipp; Czupalla, Markus

    2015-05-01

    This paper is concerned with the topic of the reduction of thermal models of spacecraft. The work presented here has been conducted in cooperation with the company OHB AG, formerly Kayser-Threde GmbH, and the Institute of Astronautics at Technische Universität München with the goal to shorten and automatize the time-consuming and manual process of thermal model reduction. The reduction of thermal models can be divided into the simplification of the geometry model for calculation of external heat flows and radiative couplings and into the reduction of the underlying mathematical model. For simplification a method has been developed which approximates the reduced geometry model with the help of an optimization algorithm. Different linear and nonlinear model reduction techniques have been evaluated for their applicability in reduction of the mathematical model. Thereby the compatibility with the thermal analysis tool ESATAN-TMS is of major concern, which restricts the useful application of these methods. Additional model reduction methods have been developed, which account to these constraints. The Matrix Reduction method allows the approximation of the differential equation to reference values exactly expect for numerical errors. The summation method enables a useful, applicable reduction of thermal models that can be used in industry. In this work a framework for model reduction of thermal models has been created, which can be used together with a newly developed graphical user interface for the reduction of thermal models in industry.

  1. Size and transportation capabilities of the existing US cask fleet

    SciTech Connect

    Danese, F.L. ); Johnson, P.E.; Joy, D.S. )

    1990-01-01

    This study investigates the current spent nuclear fuel cask fleet capability in the United States. In addition, it assesses the degree to which the current fleet would be available, as a contingency, until proposed Office of Civilian Radioactive Waste Management casks become operational. A limited fleet of ten spent fuel transportation casks is found to be readily available for use in Federal waste management efforts over the next decade.

  2. Cosmic-ray imaging of spent fuel casks

    NASA Astrophysics Data System (ADS)

    Guardincerri, Elena; Durham, J. Matthew; Morris, Christopher; Poulson, Daniel; Plaud-Ramos, Kenie; Fabritius, Joseph; Bacon, Jeffrey; Winston, Philip; Chichester, David

    2015-10-01

    Muon radiography was used to image the inside of a partially loaded Westinghouse MC-10 dry cask containing spent nuclear fuel at Idaho National Laboratory. We present here the results of a 100 hours long measurement taken in May 2015 with two muon trackers placed outside the cask. The data clearly show the location of the missing fuel bundles and demonstrate the feasibility of using cosmic rays to monitor fuel casks against illicit diversion of their content.

  3. Evaluation of Polyesterimide Nanocomposites Using Methods of Thermal Analysis

    NASA Astrophysics Data System (ADS)

    Gornicka, B.; Gorecki, L.; Gryzlo, K.; Kaczmarek, D.; Wojcieszak, D.

    2016-02-01

    Polyesterimide resin applied for winding impregnation has been modified by incorporating the hydrophilic and hydrophobic nanosilica, montmorillonite and aluminium oxide. For assessment of the resins in liquid and cured states thermoanalytical methods TG/DSC were used. For pure and nanofilled resins the results of investigation of AFM topography, bond strength, dielectric strength and partial discharge resistance have been also presented. It was found that dielectric and mechanical properties of polyesterimide resin containing hydrophilic silica as well aluminium oxide were much improved as compared to pure resin. Based on our investigations we have found that the methods of thermal analysis may be very useful for evaluation of nanocomposites: DSC/TGA study of resins in the liquid state under dynamic conditions can be applied to pre-select nanocomposites; isothermal TG curves of cured resins can be utilized for thermal stability evaluation; in turn, TG study after thermal ageing of cured resins could confirm the barrier properties of nanocomposites.

  4. Evaluation of terrestrial photogrammetric point clouds derived from thermal imagery

    NASA Astrophysics Data System (ADS)

    Metcalf, Jeremy P.; Olsen, Richard C.

    2016-05-01

    Computer vision and photogrammetric techniques have been widely applied to digital imagery producing high density 3D point clouds. Using thermal imagery as input, the same techniques can be applied to infrared data to produce point clouds in 3D space, providing surface temperature information. The work presented here is an evaluation of the accuracy of 3D reconstruction of point clouds produced using thermal imagery. An urban scene was imaged over an area at the Naval Postgraduate School, Monterey, CA, viewing from above as with an airborne system. Terrestrial thermal and RGB imagery were collected from a rooftop overlooking the site using a FLIR SC8200 MWIR camera and a Canon T1i DSLR. In order to spatially align each dataset, ground control points were placed throughout the study area using Trimble R10 GNSS receivers operating in RTK mode. Each image dataset is processed to produce a dense point cloud for 3D evaluation.

  5. Transportation cask decontamination and maintenance at the potential Yucca Mountain repository; Yucca Mountain Site characterization project

    SciTech Connect

    Hartman, D.J.; Miller, D.D.; Hill, R.R.

    1992-04-01

    This study investigates spent fuel cask handling experience at existing nuclear facilities to determine appropriate cask decontamination and maintenance operations at the potential Yucca Mountain repository. These operations are categorized as either routine or nonroutine. Routine cask decontamination and maintenance tasks are performed in the cask preparation area at the repository. Casks are taken offline to a separate cask maintenance area for major nonroutine tasks. The study develops conceptual designs of the cask preparation area and cask maintenance area. The functions, layouts, and major features of these areas are also described.

  6. Rail tiedown tests with heavy casks for radioactive shipments

    SciTech Connect

    Petry, S.F.

    1980-08-01

    A rail tiedown test program was conducted at the Savannah River Plant in July and August 1978. For each test, a 40- or 70-ton cask was secured on a railcar. The railcar was pushed to speeds up to 11 mph and allowed to couple to parked railcars simulating ordinary railyard operations. The test car carrying the cask was heavily instrumented to measure the accelerations and forces generated at strategically selected places. Eighteen test runs were made with different combinations of railcars, couplers, casks, speeds, and tiedown configurations. The major objectives of the test program were to (1) provide test data as a basis to develop a tiedown standard for rail cask shipments of radioactive materials and (2) collect dynamic data to support analytical models of the railcar cask tiedown system. The optimum tiedown configuration demonstrated for heavy casks was a combination of welded, fixed stops to secure the cask longitudinally and flexible cables to restrain vertical and lateral cask movement. Cables alone were inadequate to secure a heavy cask to a standard railcar, and bolting was found disadvantageous in several respects. The use of cushioning coupler mechanisms dramatically reduced the tiedown requirements for the rail coupling operation. The test program and general conclusions are discussed.

  7. Experimental data base for estimating the consequences from a hypothetical sabotage attack on a spent fuel shipping cask

    SciTech Connect

    Sandoval, R.P.; Luna, R.E.

    1986-01-01

    This paper describes the results of a program conducted at Sandia National Laboratories for the US Department of Energy to provide an experimental data base for estimating the radiological health effects that could result from the sabotage of a light water reactor spent fuel shipping cask. The primary objectives of the program were limited to: (1) evaluating the effectiveness of selected high energy devices (HED) in breaching full-scale spent fuel shipping casks, (2) quantifying and characterizing relevant aerosol and radiological properties of the released fuel, and (3) using the resulting experimental data to evaluate the radiological health effects resulting from a hypothetical attack on a spent fuel shipping cask in a densely populated urban area. 3 refs.

  8. Thermal energy effects on articular cartilage: a multidisciplinary evaluation

    NASA Astrophysics Data System (ADS)

    Kaplan, Lee D.; Ernsthausen, John; Ionescu, Dan S.; Studer, Rebecca K.; Bradley, James P.; Chu, Constance R.; Fu, Freddie H.; Farkas, Daniel L.

    2002-05-01

    Partial thickness articular cartilage lesions are commonly encountered in orthopedic surgery. These lesions do not have the ability to heal by themselves, due to lack of vascular supply. Several types of treatment have addressed this problem, including mechanical debridement and thermal chondroplasty. The goal of these treatments is to provide a smooth cartilage surface and prevent propagation of the lesions. Early thermal chondroplasty was performed using lasers, and yielded very mixed results, including severe damage to the cartilage, due to poor control of the induced thermal effects. This led to the development (including commercial) of probes using radiofrequency to generate the thermal effects desired for chondroplasty. Similar concerns over the quantitative aspects and control ability of the induced thermal effects in these treatments led us to test the whole range of complex issues and parameters involved. Our investigations are designed to simultaneously evaluate clinical conditions, instrument variables for existing radiofrequency probes (pressure, speed, distance, dose) as well as the associated basic science issues such as damage temperature and controllability (down to the subcellular level), damage geometry, and effects of surrounding conditions (medium, temperature, flow, pressure). The overall goals of this work are (1) to establish whether thermal chondroplasty can be used in a safe and efficacious manner, and (2) provide a prescription for multi-variable optimization of the way treatments are delivered, based on quantitative analysis. The methods used form an interdisciplinary set, to include precise mechanical actuation, high accuracy temperature and temperature gradient control and measurement, advanced imaging approaches and mathematical modeling.

  9. Evaluation of Oxidation Damage in Thermal Barrier Coating Systems

    NASA Technical Reports Server (NTRS)

    Zhu, Dongming; Miller, Robert A.

    1996-01-01

    A method based on the technique of dilatometry has been established to quantitatively evaluate the interfacial damage due to the oxidation in a thermal barrier coating system. Strain isolation and adhesion coefficients have been proposed to characterize the thermal barrier coating (TBC) performance based on its thermal expansion behavior. It has been found that, for a thermal barrier coating system consisting of ZrO2-8%Y2O3/FeCrAlY/4140 steel substrate, the oxidation of the bond coat and substrate significantly reduced the ceramic coating adherence, as inferred from the dilatometry measurements. The in-situ thermal expansion measurements under 30 deg C to 700 deg C thermal cycling in air showed that the adhesion coefficient, A(sub i) decreased by 25% during the first 35 oxidation cycles. Metallography showed that delamination occurred at both the ceramic/bond coat and bond coat/substrate interfaces. In addition, the strain isolation effect has been improved by increasing the FeCrAlY bond coat thickness. The strain isolation coefficient, Si, increased from about 0.04 to 0.25, as the bond coat thickness changed from 0.1 mm to 1.0 mm. It may be possible to design optimum values of strain isolation and interface adhesion coefficients to achieve the best TBC performance.

  10. Global approach toward the evaluation of thermal infrared countermeasures

    NASA Astrophysics Data System (ADS)

    Verlinde, Patrick S. A.; Proesmans, Marc

    1991-07-01

    This paper proposes a procedure for the numerical evaluation of the efficiency of countermeasures in the thermal infrared. This procedure consists of three phases. In the first phase, the characteristics of different thermal camouflage materials are tested on a lab-scale. These tests comprise measurements of the attenuation of the incident infrared energy and/or of the thermal emissivity factor. With respect to the attenuation measurement, a calibrated infrared sensor is used to determine the radiation patterns of an object. The comparison of these patterns before and after the application of a camouflage system gives an absolute measure of its attenuation. The result of this measurement is important since the attenuation is closely related to the contrast between the camouflaged object and its background and thus to the probability of detection. Contrast, however, is not the only important feature for the detection of an object in a thermal image. That is why in a second phase the countermeasures under evaluation are tested in a real environment. During this phase, a numerical value is given to the efficiency of the considered camouflage in the thermal infrared, using features selected from those which are known to be important for human vision. These include, besides contrast, other features such as correlation and texture. The third and final phase aims at a verification and a validation of the test results. Indeed, it is of a crucial importance to find a link between the performances obtained in the field and the characteristics measured on a lab- scale. It is also necessary to verify that a good correlation exists between the efficiency as determined by human observers or by numerical evaluation. This is now taking place using a database of thermal images taken with a GEC Avionics TICM II (8 to 12 micrometers ). Those images are then presented to human observers and to the machine in a project called Psychotest.

  11. Global approach towards the evaluation of thermal infrared countermeasures

    NASA Astrophysics Data System (ADS)

    Verlinde, Patrick; Wilms, Philippe

    1993-11-01

    This paper proposes a procedure for the numerical evaluation of the efficiency of countermeasures in the thermal infrared. This procedure consists of three phases. In the first phase the characteristics of different thermal camouflage materials are tested on a lab-scale. These tests comprise measurements of the attenuation of the incident infrared energy and/or of the thermal emissivity factor. With respect to the attenuation measurement, a calibrated infrared sensor is used, to determine the radiation patterns of an object. The comparison of these patterns before and after the application of a camouflage system, gives an absolute measure of its attenuation. The result of this measurement is important since the attenuation is closely related to the contrast between the camouflaged object and its background and thus to the probability of detection. Contrast however, is not the only important feature for the detection of an object in a thermal image. That is why in a second phase the countermeasures under evaluation are tested in a real environment. During this phase, a numerical value is given to the efficiency of the considered camouflage in the thermal infrared, using features selected from those which are known to be important for human vision. These include, besides contrast, other features such as correlation and texture. The third and final phase aims at a verification and a validation of the test results. Indeed, it is of crucial importance to find a link between the performance obtained in the field and the characteristics measured on a lab-scale. It is also necessary to verify that a good correlation exists between the efficiency as determined by human observers and the result of this numerical evaluation procedure. This will be done using a database of thermal images taken with a GEC Avionics TICM II (8 to 12 micron). Those images are then presented as well to human observers as to the machine in a project called 'Psychotest'.

  12. Evaluating the ignition sensitivity of thermal battery heat pellets

    SciTech Connect

    Thomas, E.V.

    1993-09-01

    Thermal batteries are activated by the ignition of heat pellets. If the heat pellets are not sensitive enough to the ignition stimulus, the thermal battery will not activate, resulting in a dud. Thus, to assure reliable thermal batteries, it is important to demonstrate that the pellets have satisfactory ignition sensitivity by testing a number of specimens. There are a number of statistical methods for evaluating the sensitivity of a device to some stimulus. Generally, these methods are applicable to the situation in which a single test is destructive to the specimen being tested, independent of the outcome of the test. In the case of thermal battery heat pellets, however, tests that result in a nonresponse do not totally degrade the specimen. This peculiarity provides opportunities to efficiently evaluate the ignition sensitivity of heat pellets. In this paper, a simple strategy for evaluating heat pellet ignition sensitivity (including experimental design and data analysis) is described. The relatively good asymptotic and small-sample efficiencies of this strategy are demonstrated.

  13. Using fractal analysis of thermal signatures for thyroid disease evaluation

    NASA Astrophysics Data System (ADS)

    Gavriloaia, Gheorghe; Sofron, Emil; Gavriloaia, Mariuca-Roxana; Ghemigean, Adina-Mariana

    2010-11-01

    The skin is the largest organ of the body and it protects against heat, light, injury and infection. Skin temperature is an important parameter for diagnosing diseases. Thermal analysis is non-invasive, painless, and relatively inexpensive, showing a great potential research. Since the thyroid regulates metabolic rate it is intimately connected to body temperature, more than, any modification of its function generates a specific thermal image on the neck skin. The shapes of thermal signatures are often irregular in size and shape. Euclidean geometry is not able to evaluate their shape for different thyroid diseases, and fractal geometry is used in this paper. Different thyroid diseases generate different shapes, and their complexity are evaluated by specific mathematical approaches, fractal analysis, in order to the evaluate selfsimilarity and lacunarity. Two kinds of thyroid diseases, hyperthyroidism and papillary cancer are analyzed in this paper. The results are encouraging and show the ability to continue research for thermal signature to be used in early diagnosis of thyroid diseases.

  14. Nuclear thermal rocket nozzle testing and evaluation program

    NASA Technical Reports Server (NTRS)

    Davidian, Kenneth O.; Kacynski, Kenneth J.

    1993-01-01

    Performance characteristics of the Nuclear Thermal Rocket can be enhanced through the use of unconventional nozzles as part of the propulsion system. The Nuclear Thermal Rocket nozzle testing and evaluation program being conducted at the NASA Lewis is outlined and the advantages of a plug nozzle are described. A facility description, experimental designs and schematics are given. Results of pretest performance analyses show that high nozzle performance can be attained despite substantial nozzle length reduction through the use of plug nozzles as compared to a convergent-divergent nozzle. Pretest measurement uncertainty analyses indicate that specific impulse values are expected to be within + or - 1.17 pct.

  15. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    SciTech Connect

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  16. 239Pu Resonance Evaluation for Thermal Benchmark System Calculations

    NASA Astrophysics Data System (ADS)

    Leal, L. C.; Noguere, G.; de Saint Jean, C.; Kahler, A. C.

    2014-04-01

    Analyses of thermal plutonium solution critical benchmark systems have indicated a deficiency in the 239Pu resonance evaluation. To investigate possible solutions to this issue, the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Working Party for Evaluation Cooperation (WPEC) established Subgroup 34 to focus on the reevaluation of the 239Pu resolved resonance parameters. In addition, the impacts of the prompt neutron multiplicity (νbar) and the prompt neutron fission spectrum (PFNS) have been investigated. The objective of this paper is to present the results of the 239Pu resolved resonance evaluation effort.

  17. Neutron shielding analysis for remote handled transuranic waste containers in facility casks at the Waste Isolation Pilot Plant

    SciTech Connect

    Livingston, J.V.; Disney, R.K.

    1984-04-01

    Neutron shielding characteristics of the Waste Isolation Pilot Plant facility cask have been quantified for a variety of combinations of neutron sources and waste matrices which would potentially be handled in waste containers. The neutron attenuation and neutron environment of the waste container and the facility cask have been analyzed to ensure that the design requirement of neutron dose rate will be met under the combinations of the source and waste matrix conditions. The analyses considered the ranges of neutron source spectrum and waste matrices which combine to produce the minimum neutron shielding worth of the facility cask. One-dimensional analyses were performed with discrete ordinate transport theory methods using multigroup neutron cross section data. The results discussed in this report demonstrate the effect of source spectrum and waste container matrix on predicted neutron dose rates adjacent to the unshielded waste container and the surface of the facility cask. An evaluation of the uncertainties in predicted neutron dose rates is provided which results in an assessment of the maximum measured neutron dose rate external to the facility cask. A description of the analytical models developed, the analysis methodology, the neutron source spectra, and the detailed results are described in this report. 10 refs., 50 figs., 39 tabs.

  18. Evaluation of thermal gradients in longitudinal spin Seebeck effect measurements

    NASA Astrophysics Data System (ADS)

    Sola, A.; Kuepferling, M.; Basso, V.; Pasquale, M.; Kikkawa, T.; Uchida, K.; Saitoh, E.

    2015-05-01

    In the framework of the longitudinal spin Seebeck effect (LSSE), we developed an experimental setup for the characterization of LSSE devices. This class of device consists in a layered structure formed by a substrate, a ferrimagnetic insulator (YIG) where the spin current is thermally generated, and a paramagnetic metal (Pt) for the detection of the spin current via the inverse spin-Hall effect. In this kind of experiments, the evaluation of a thermal gradient through the thin YIG layer is a crucial point. In this work, we perform an indirect determination of the thermal gradient through the measurement of the heat flux. We developed an experimental setup using Peltier cells that allow us to measure the heat flux through a given sample. In order to test the technique, a standard LSSE device produced at Tohoku University was measured. We find a spin Seebeck SSSE coefficient of 2.8 × 10 - 7 V K-1.

  19. Evaluation of infrared collimators for testing thermal imaging systems

    NASA Astrophysics Data System (ADS)

    Chrzanowski, K.

    2007-06-01

    Infrared reflective collimators are important components of expensive sophisticated test systems used for testing thermal imagers. Too low quality collimators can become a source of significant measurement errors and collimators of too high quality can unnecessarily increase cost of a test system. In such a situation it is important for test system users to know proper requirements on the collimator and to be able to verify its performance. A method for evaluation of infrared reflective collimators used in test systems for testing thermal imagers is presented in this paper. The method requires only easily available optical equipment and can be used not only by collimator manufactures but also by users of test equipment to verify performance of the collimators used for testing thermal imagers.

  20. Summary of the thermal evaluation of LWBR (LWBR Development Program)

    SciTech Connect

    Lerner, S.; McWilliams, K.D.; Stout, J.W.; Turner, J.R.

    1980-03-01

    This report describes the thermal evaluation of the core for the Shippingport Light Water Breeder Reactor. This core contains unique thermal-hydraulic features such as (1) close rod-to-rod proximity, (2) an open-lattice array of fuel rods with two different diameters and rod-to-rod spacings in the same flow region, (3) triplate orifices located at both the entrance and exit of fuel modules and (4) a hydraulically-balanced movable-fuel system coupled with (5) axial-and-radial fuel zoning for reactivity control. Performance studies used reactor thermal principles such as the hot-and-nominal channel concept and related nuclear/engineering design allowances. These were applied to models of three-dimensional rodded arrays comprising the core fuel regions.

  1. Evaluating Three Active Thermal Architectures for Exploration Missions

    NASA Technical Reports Server (NTRS)

    Cross, Cynthia D.; Hong, Andrew E.; Sheth, Rubik B.; Navarro, Moses; Marett, Susan J.

    2012-01-01

    Mass and cost are typically the two biggest challenges facing space craft designers. Active thermal control systems for crewed space-craft are typically among the more massive and costly systems on the vehicle. A study was completed evaluating three different thermal control system architectures to evaluate overall performance, mass and cost for a typical exploration mission profile. The architectures that were evaluated were 1 - a two-loop system using an internal liquid loop interfacing with an external liquid loop and flow loop with flow through radiators; 2 - a-single loop architecture with flow through radiators utilizing a regenerative heat exchanger and heater; and 3 - a single-loop architecture with heat pipe radiators. Environmental conditions, calculated for a given lunar exploration mission, and mission heat load profiles, generated based on previous Orion time lines, were evalauated through the phases of the on orbit mission. Performance for each of the architectures was evaluated along with the resultant mass of each system. Recommendations include adding a thermal topping system to lunar missions due to the extreme hot environments encountered in near-lunar approaches.

  2. Thermal parametric imaging in the evaluation of skin burn depth.

    PubMed

    Rumiński, Jacek; Kaczmarek, Mariusz; Renkielska, Alicja; Nowakowski, Antoni

    2007-02-01

    The aim of this paper is to determine the extent to which infrared (IR) thermal imaging may be used for skin burn depth evaluation. The analysis can be made on the basis of the development of a thermal model of the burned skin. Different methods such as the traditional clinical visual approach and the IR imaging modalities of static IR thermal imaging, active IR thermal imaging and active-dynamic IR thermal imaging (ADT) are analyzed from the point of view of skin burn depth diagnostics. In ADT, a new approach is proposed on the basis of parametric image synthesis. Calculation software is implemented for single-node and distributed systems. The properties of all the methods are verified in experiments using phantoms and subsequently in vivo with animals with a reference histopathological examination. The results indicate that it is possible to distinguish objectively and quantitatively burns which will heal spontaneously within three weeks of infliction and which should be treated conservatively from those which need surgery because they will not heal within this period. PMID:17278587

  3. Evaluation of laser prostatectomy devices by thermal imaging

    NASA Astrophysics Data System (ADS)

    Molenaar, David G.; van Vliet, Remco J.; van Swol, Christiaan F. P.; Boon, Tom A.; Verdaasdonck, Rudolf M.

    1994-12-01

    The treatment of benign prostatic hyperplasia (BPH) using Nd:YAG laser light has become an accepted alternative to TURP. However, there is no consensus to the dosimetry using the various laser devices. In our study, we evaluate the optical and thermal characteristics of 7 commercially available side firing laser probes. For the thermal analysis, an optical method was used based on `Schlieren' techniques producing color images of the temperature distribution around the laser probe in water. Absolute temperatures were obtained after calibration measurements with thermocouples. Laser probes using metal mirrors for beam deflection heated up entirely. The local temperature rose up to 100 degrees centigrade, thus inducing vapor bubble formation that interfered with the emitted beam. Laser devices, using total internal reflection for deflection, showed far less heating primarily at the exit window, though Fresnel reflections and secondary beams indirectly heated up the (metal) housing of the tip. After clinical application, the absorption at the probe surface and hence temperature increased due to probe deterioration. Color Schlieren imaging is a powerful method for the thermal evaluation of laser devices. The thermal behavior of laser probes can be used as a guidance for the method of application and as an indication of the lifetime of the probes.

  4. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-24

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... issued a final rule on December 22,1994 (59 FR 65898), that approved the Standardized NUHOMS Cask System... 3, 1997 (62 FR 46517), this rule is classified as Compatibility ] Category ``NRC.'' Compatibility...

  5. SRTC criticality safety technical review: Phase 1 criticality analysis for the 9972-9975 family of shipping casks: (SRT-CMA-940003)

    SciTech Connect

    Rathbun, R.

    1994-03-02

    Review of SRT-CMA-940003, ``Phase I Criticality Analysis For The 9972-9975 Family Of Shipping Casks (U). (SRT-CMA-940003).`` January 22, 1994, has been performed by the SRTC Applied Physics Group. The NCSE is a criticality assessment of the 9972-9975 family of shipping casks. This work is a follow-on of a previous criticality safety evaluation, with the differences between this and the previous evaluation are that now wall tolerances are modeled and more sophisticated analytical methods are applied. The NCSE under review concludes that, with one exception, the previously specified plutonium and uranium mass limits for 9972-9975 family of shipping casks do ensure that WSRC Nuclear Criticality Safety Manual requirements (ref. 1) are satisfied. The one exception is that the plutonium mass limit for the 9974 cask had to be reduced from 4.4 to 4.3 kg. In contrast, the 7.5 kg uranium mass limit for the 9974 cask was raised to 14.5 kg, making the uranium mass identical for all casks in this family. This technical review consisted of an independent check of the methods and models employed, application of ANSI/ANS 8.1 and 8.15, and verification of WSRC Nuclear Criticality Safety Manual procedures.

  6. Thermal performance evaluation of the infrared telescope dewar subsystem

    NASA Technical Reports Server (NTRS)

    Urban, E. W.

    1986-01-01

    Thermal performance evaluations (TPE) were conducted with the superfluid helium dewar of the Infrared Telescope (IRT) experiment from November 1981 to August 1982. Test included measuring key operating parameters, simulating operations with an attached instrument cryostat and validating servicing, operating and safety procedures. Test activities and results are summarized. All objectives are satisfied except for those involving transfer of low pressure liquid helium (LHe) from a supply dewar into the dewar subsystem.

  7. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  8. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  9. Interaction of cosmic ray muons with spent nuclear fuel dry casks and determination of lower detection limit

    NASA Astrophysics Data System (ADS)

    Chatzidakis, S.; Choi, C. K.; Tsoukalas, L. H.

    2016-08-01

    The potential non-proliferation monitoring of spent nuclear fuel sealed in dry casks interacting continuously with the naturally generated cosmic ray muons is investigated. Treatments on the muon RMS scattering angle by Moliere, Rossi-Greisen, Highland and, Lynch-Dahl were analyzed and compared with simplified Monte Carlo simulations. The Lynch-Dahl expression has the lowest error and appears to be appropriate when performing conceptual calculations for high-Z, thick targets such as dry casks. The GEANT4 Monte Carlo code was used to simulate dry casks with various fuel loadings and scattering variance estimates for each case were obtained. The scattering variance estimation was shown to be unbiased and using Chebyshev's inequality, it was found that 106 muons will provide estimates of the scattering variances that are within 1% of the true value at a 99% confidence level. These estimates were used as reference values to calculate scattering distributions and evaluate the asymptotic behavior for small variations on fuel loading. It is shown that the scattering distributions between a fully loaded dry cask and one with a fuel assembly missing initially overlap significantly but their distance eventually increases with increasing number of muons. One missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the distributions which is the case of 100,000 muons. This indicates that the removal of a standard fuel assembly can be identified using muons providing that enough muons are collected. A Bayesian algorithm was developed to classify dry casks and provide a decision rule that minimizes the risk of making an incorrect decision. The algorithm performance was evaluated and the lower detection limit was determined.

  10. Fire tests and analyses of a rail cask-sized calorimeter.

    SciTech Connect

    Figueroa, Victor G.; Lopez, Carlos; Suo-Anttila, Ahti Jorma; Greiner, Miles

    2010-10-01

    Three large open pool fire experiments involving a calorimeter the size of a spent fuel rail cask were conducted at Sandia National Laboratories Lurance Canyon Burn Site. These experiments were performed to study the heat transfer between a very large fire and a large cask-like object. In all of the tests, the calorimeter was located at the center of a 7.93-meter diameter fuel pan, elevated 1 meter above the fuel pool. The relative pool size and positioning of the calorimeter conformed to the required positioning of a package undergoing certification fire testing. Approximately 2000 gallons of JP-8 aviation fuel were used in each test. The first two tests had relatively light winds and lasted 40 minutes, while the third had stronger winds and consumed the fuel in 25 minutes. Wind speed and direction, calorimeter temperature, fire envelop temperature, vertical gas plume speed, and radiant heat flux near the calorimeter were measured at several locations in all tests. Fuel regression rate data was also acquired. The experimental setup and certain fire characteristics that were observed during the test are described in this paper. Results from three-dimensional fire simulations performed with the Cask Analysis Fire Environment (CAFE) fire code are also presented. Comparisons of the thermal response of the calorimeter as measured in each test to the results obtained from the CAFE simulations are presented and discussed.

  11. Evaluating CFRP-Masonry Bond Using Thermal Imaging

    NASA Astrophysics Data System (ADS)

    Ross, Joseph C.

    This study presents results from non-destructive testing to evaluate the degradation of the CFRP-masonry bond using thermal imaging. The goal of the research was to identify locations where there was evidence of bond deterioration that could subsequently be verified through destructive pull-off testing. Four full-scale masonry walls were built outdoors at the University of South Florida in 1995 to evaluate the effectiveness of CFRP for repairing settlement damage. Two of the settlement-damaged walls were repaired using single layer, commercially available unidirectional CFRP systems that used Tonen (wall 3) and Henkel (wall 2) epoxies. These two walls were the subject of this investigation. Before non-destructive tests were initiated, historical site data on temperature, humidity and rainfall variation was compiled. Over seventeen years, the walls experienced ambient temperatures as high as 98°F and as low as 25°F. The average rainfall in Tampa is about 34 inches and the annual average high humidity is around 87%. Because of the high temperature and humidity, the CFRP-masonry bond was exposed to a particularly aggressive environment. Three types of thermal evaluation were carried out: thermocouple monitoring and both passive (solar) and active (localized heating) infrared thermal imaging. Twenty-four thermocouples were used to observe the spatial variations in temperature on the wall. Data showed that the surface temperatures of the wall are uneven with one end being hotter than the other. Measurements indicated that the wall temperatures went as high as 103°F during the week of data collection in late March and early April of 2012. In contrast, the highest ambient temperature over the same period was 92°F. The high temperature experienced by the wall is below the glass transition temperature for the epoxies, which ranges from 140°F to 180°F. A FLIR Tau 320 thermal imaging camera was used to identify localized de-bonding. Solar radiation heated the walls and

  12. MSFC Skylab Apollo Telescope Mount thermal control system mission evaluation

    NASA Technical Reports Server (NTRS)

    Hueter, U.

    1974-01-01

    The Skylab Saturn Workshop Assembly was designed to expand the knowledge of manned earth orbital operations and accomplish a multitude of scientific experiments. The Apollo Telescope Mount (ATM), a module of the Skylab Saturn Workshop Assembly, was the first manned solar observatory to successfully observe, monitor, and record the structure and behavior of the sun outside the earth's atmosphere. The ATM contained eight solar telescopes that recorded solar phenomena in X-ray, ultraviolet, white light, and hydrogen alpha regions of the electromagnetic spectrum. In addition, the ATM contained the Saturn Workshop Assembly's pointing and attitude control system, a data and communication system, and a solar array/rechargeable battery power system. This document presents the overall ATM thermal design philosophy, premission and mission support activity, and the mission thermal evaluation. Emphasis is placed on premission planning and orbital performance with particular attention on problems encountered during the mission. ATM thermal performance was satisfactory throughout the mission. Although several anomalies occurred, no failure was directly attributable to a deficiency in the thermal design.

  13. Experimental evaluation of outer planets probe thermal insulation concepts

    NASA Technical Reports Server (NTRS)

    Grote, M. G.; Mezines, S. A.

    1976-01-01

    An experimental program was conducted to evaluate various thermal insulation concepts for use in the Outer Planets Probe (OPP) during entry and descent into the atmospheres of Jupiter, Saturn, and Uranus. Phenolic fiberglass honeycomb specimens representative of the OPP structure were packed and tested with various fillers: Thermal conductivity measurements were made over a temperature range of 300 K to 483 K and pressures from vacuum up to 10 atmospheres in helium and nitrogen gas environments. The conductivity results could not be fully explained so new test specimens were designed with improved venting characteristics, and tested to determine the validity of the original data. All of the conductivity data showed results that were substantially higher than expected. The original test data in helium were lower than the data from the redesigned specimens, probably due to inadequate venting of nitrogen gas from the original specimens. The thermal conductivity test results show only a marginal improvement in probe thermal protection performance for a filled honeycomb core compared to an unfilled core. In addition, flatwise tension tests showed a severe bond strength degradation due to the inclusion of either the powder or foam fillers. In view of these results, it is recommended that the baseline OPP design utilize an unfilled core.

  14. Synchronized Electronic Shutter System (SESS) for Thermal Nondestructive Evaluation

    NASA Technical Reports Server (NTRS)

    Zalameda, Joseph N.

    2001-01-01

    The purpose of this paper is to describe a new method for thermal nondestructive evaluation. This method uses a synchronized electronic shutter system (SESS) to remove the heat lamp's influence on the thermal data during and after flash heating. There are two main concerns when using flash heating. The first concern is during the flash when the photons are reflected back into the camera. This tends to saturate the detectors and potentially introduces unknown and uncorrectable errors when curve fitting the data to a model. To address this, an electronically controlled shutter was placed over the infrared camera lens. Before firing the flash lamps, the shutter is opened to acquire the necessary background data for offset calibration. During flash heating, the shutter is closed to prevent the photons from the high intensity flash from saturating the camera's detectors. The second concern is after the flash heating where the lamps radiate heat after firing. This residual cooling introduces an unwanted transient thermal response into the data. To remove this residual effect, a shutter was placed over the flash lamps to block the infrared heat radiating from the flash head after heating. This helped to remove the transient contribution of the flash. The flash lamp shutters were synchronized electronically with the camera shutter. Results are given comparing the use of the thermal inspection with and without the shutter system.

  15. Synchronized electronic shutter system (SESS) for thermal nondestructive evaluation

    NASA Astrophysics Data System (ADS)

    Zalameda, Joseph N.

    2001-03-01

    The purpose of this paper is to describe a new method for thermal nondestructive evaluation. This method uses a synchronized electronic shutter system (SESS) to remove the heat lamp's influence on the thermal data during and after flash heating. There are two main concerns when using flash heating. The first concern is during the flash when the photons are reflected back into the camera. This tends to saturate the detectors and potentially introduces unknown and uncorrectable errors when curve fitting the data to a model. To address this, an electronically controlled shutter was placed over the infrared camera lens. Before firing the flash lamps, the shutter is opened to acquire the necessary background data for offset calibration. During flash heating, the shutter is closed to prevent the photons from the high intensity flash from saturating the camera's detectors. The second concern is after the flash heating where the lamps radiate heat after firing. This residual cooling introduces an unwanted transient thermal response into the data. To remove this residual effect, a shutter was placed over the flash lamps to block the infrared heat radiating from the flash head after heating. This helped to remove the transient contribution of the flash. The flash lamp shutters were synchronized electronically with the camera shutter. Results are given comparing the use of the thermal inspection with and without the shutter system.

  16. Ultrasonic Evaluation of Thermal Degradation in Adhesive Bonds

    NASA Technical Reports Server (NTRS)

    Lih, Shyh-Shiuh; Mal, Ajit K.; Bar-Cohen, Yoseph

    1994-01-01

    The critical role played by adhesive bonds in lap joints is well known. A good knowledge of the mechanical properties of adhesive bonds in lap joints is a prerequisite to the design and reliable prediction of the performance of these bonded structures. Furthermore, the lap joint may be subject to high-temperature environments in service. Early detection of the degree of thermal degradation in adhesive bonds is required under these circumstances. A variety of ultrasonic nondestructive evaluation (NDE) techniques can be used to determine the thickness and the elastic moduli of adhesively bonded joints. In this paper we apply a previously developed technique based on the leaky Lamb wave (LLW) experiment to investigate the possibility of characterizing the thermal degradation of adhesive bonds in lap joints. The degradation of the adhesive bonds is determined through comparison between experimental data and theoretical calculations.

  17. Thermal Evaluation for the Naval SNF Waste Package

    SciTech Connect

    T.L. Mitchell

    2000-04-25

    The purpose of this calculation is to evaluate the thermal performance of the naval long spent nuclear fuel (SNF) waste package (WP) under multiple disposal conditions in a monitored geologic repository (MGR). The scope of this calculation is limited to determination of thermal temperature profiles upon the surface of, and within, the naval long SNF WP. The objective is to develop a temperature profile history within the WP, at time increments up to 10,000 years of emplacement. The results of this calculation are intended to support the Naval SNF WP Analysis and Model Report (AMR) for Site Recommendation (SR). This calculation was performed to the specifications within its Technical Development Plan (TDP) (Ref. 8.16). This calculation is developed and documented in accordance with the AP-3.12Q/REV. 0IICN. 0 procedure, Calculations.

  18. Thermal Evaluation of the Fort Saint Vrain Codisposal Waste Package

    SciTech Connect

    Adam Scheider; Horia Radulescu

    2001-07-19

    The objective of this calculation is to evaluate the thermal response of the Fort Saint Vrain (FSV) Codisposal Waste Package (WP) design under nominal Monitored Geologic Repository conditions. The objective of the calculation is to provide thermal parameter information to support the FSV waste package design. The information provided by the sketches (Attachment IV) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.124, ''Calculations'' (Ref. 17) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the SDHLW (Defense High Level Waste) / DOE (Department of Energy) Long WP.

  19. Evaluation of thermal and non-thermal effects of UHF RFID exposure on biological drugs.

    PubMed

    Calcagnini, Giovanni; Censi, Federica; Maffia, Michele; Mainetti, Luca; Mattei, Eugenio; Patrono, Luigi; Urso, Emanuela

    2012-11-01

    The Radio Frequency Identification (RFID) technology promises to improve several processes in the healthcare scenario, especially those related to traceability of people and things. Unfortunately, there are still some barriers limiting the large-scale deployment of these innovative technologies in the healthcare field. Among these, the evaluation of potential thermal and non-thermal effects due to the exposure of biopharmaceutical products to electromagnetic fields is very challenging, but still slightly investigated. This paper aims to setup a controlled RF exposure environment, in order to reproduce a worst-case exposure of pharmaceutical products to the electromagnetic fields generated by the UHF RFID devices placed along the supply chain. Radiated powers several times higher than recommended by current normative limits were applied (10 W and 20 W). The electric field strength at the exposed sample location, used in tests, was as high as 100 V/m. Non-thermal effects were evaluated by chromatography techniques and in vitro assays. The results obtained for a particular case study, the ActrapidTM human insulin preparation, showed temperature increases lower than 0.5 °C and no significant changes in the structure and performance of the considered drug. PMID:22717524

  20. Thermal evaluation facility for LMFBR spent fuel transport

    SciTech Connect

    Wesley, D.A.

    1980-04-01

    A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations.

  1. Nuclear cask testing films misleading and misused

    SciTech Connect

    Audin, L. , Ossining, NY )

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  2. Nuclear cask testing films misleading and misused

    SciTech Connect

    Audin, L.

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  3. Modifications to SAS4 to provide cask dose rate profiles

    SciTech Connect

    Napolitano, D.G.; Sweezy, J.E.; Henkel, C.S.

    1997-12-01

    SAS4 of the SCALE code system has been used extensively by NAC International (NAC) to perform storage and transport cask shielding analyses. SAS4 utilizes a one-dimensional XSDRNPM adjoint calculation of the cask to generate biasing parameters for a three-dimensional MORSE-SGC Monte Carlo model of the cask geometry. This technique is very efficient in getting particles to tally at the cask exterior surfaces. However, SAS4/MORSE-SGC is limited to the use of point detectors (next-event estimators) and large surface detectors (surface-crossing estimators). Modifications to SAS4 were made to allow a more flexible use of the surface detectors. This modification allows multiple nonoverlapping surface detectors on each surface and allows each surface detector to be broken into subdetectors. The use of subdetectors enables the user to obtain detailed surface dose rate profiles. Tallies can now be performed on all surfaces of the cask and at user-specified distances from the cask surface. The subdetectors provide an alternative to point detectors and excessive computational time. The NAC version of SAS4 is called SAS4A. A comparison of CPU time and dose rates is made between SAS4 point detectors and SAS4A surface subdetection on the NLI {1/2} transport cask.

  4. Spreader beam analysis for the CASTOR GSF cask

    SciTech Connect

    Clements, E.P.

    1997-04-07

    The purpose of this report is to document the results of the 150% rated capacity load test performed by DynCorp Hoisting and Rigging on the CASTOR GSF special cask lifting beams. The two lifting beams were originally rated and tested at 20,000kg (44,000lb) by the cask manufacturer in Germany. The testing performed by DynCorp rated and tested the lifting beams to 30,000 kg (66,000 lb) +0%, -5%, for Hanford Site use. The CASTOR GSF cask, used to transport isotopic Heat Sources (canisters), must be lifted with its own designed lifting beam system (Figures 1, 2, and 3). As designed, the beam material is RSt 37-2 (equivalent to American Society for Testing and Materials [ASTM] A-570), the eye plate is St 52-2 (equivalent to ASTM A-516), and the lifting pin is St 50 (equivalent to ASTM A-515). The beam has two opposing 58 mm (2.3 in.) diameter by 120 mm(4.7 in.) length, high grade steel pins that engage the cask for lifting. The pins have a manual locking mechanism to prevent disengagement from the casks. The static, gross weight (loaded) of the cask 18,640 kg (41,000 lb) on the pins prevents movement of the pins during lifting. This is due to the frictional force of the cask on the pins when lifting begins.

  5. Evaluation of the Lifetime and Thermal Conductivity of Dysprosia-Stabilized Thermal Barrier Coating Systems

    NASA Astrophysics Data System (ADS)

    Curry, Nicholas; Markocsan, Nicolaie; Östergren, Lars; Li, Xin-Hai; Dorfman, Mitch

    2013-08-01

    The aim of this study was the further development of dysprosia-stabilized zirconia coatings for gas turbine applications. The target for these coatings was a longer lifetime and higher insulating performance compared to today's industrial standard thermal barrier coating. Two morphologies of ceramic top coat were studied: one using a dual-layer system and the second using a polymer to generate porosity. Evaluations were carried out using a laser flash technique to measure thermal properties. Lifetime testing was conducted using thermo-cyclic fatigue testing. Microstructure was assessed with SEM and Image analysis was used to characterize porosity content. The results show that coatings with an engineered microstructure give performance twice that of the present reference coating.

  6. Evaluation of airborne thermal, magnetic, and electromagnetic characterization technologies

    SciTech Connect

    Josten, N.E.

    1992-03-01

    The identification of Buried Structures (IBS) or Aerial Surveillance Project was initiated by the US Department of Energy (DOE) Office of Technology Development to demonstrate airborne methods for locating and identifying buried waste and ordnance at the Idaho National Engineering Laboratory (INEL). Two technologies were demonstrated: (a) a thermal infrared imaging system built by Martin Marietta Missile Systems and (b) a magnetic and electromagnetic (EM) geophysical surveying system operated by EBASCO Environmental. The thermal system detects small differences in ground temperature caused by uneven heating and cooling of the ground by the sun. Waste materials on the ground can be detected when the temperature of the waste is different than the background temperature. The geophysical system uses conventional magnetic and EM sensors. These sensors detect disturbances caused by magnetic or conductive waste and naturally occurring magnetic or conductive features of subsurface soils and rock. Both systems are deployed by helicopter. Data were collected at four INEL sites. Tests at the Naval Ordnance Disposal Area (NODA) were made to evaluate capabilities for detecting ordnance on the ground surface. Tests at the Cold Simulated Waste Demonstration Pit were made to evaluate capabilities for detecting buried waste at a controlled site, where the location and depth of buried materials are known. Tests at the Subsurface Disposal Area and Stationary Low-Power Reactor-1 burial area were made to evaluate capabilities for characterizing hazardous waste at sites that are typical of DOE buried waste sites nationwide.

  7. STRUCTURAL ANALYSES OF FUEL CASKS SUBJECTED TO BOLT PRELOAD, INTERNAL PRESSURE AND SEQUENTIAL DYNAMIC IMPACTS

    SciTech Connect

    Wu, T

    2009-06-25

    Large fuel casks subjected to the combined loads of closure bolt tightening, internal pressure and sequential dynamic impacts present challenges when evaluating their performance in the Hypothetical Accident Conditions (HAC) specified in the Code of Federal Regulations Title 10 Part 71 (10CFR71). Testing is often limited by cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing by using simplified analytical methods. In addition, there are no realistic analyses of closure bolt stresses for HAC conditions reported in the open literature. This paper presents a numerical technique for analyzing the accumulated damages of a large fuel cask caused by the sequential loads of the closure bolt tightening and the internal pressure as well as the drop and crash dynamic loads. The bolt preload and the internal pressure are treated as quasi-static loads so that the finite element method with explicit numerical integration scheme based on the theory of wave propagation can be applied. The dynamic impacts with short durations such as the 30-foot drop and the 40-inch puncture for the hypothetical accident conditions specified in 10CFR71 are also analyzed by using the finite-element method with explicit numerical integration scheme.

  8. Evaluation of diurnal thermal energy storage combined with cogeneration systems

    NASA Astrophysics Data System (ADS)

    Somasundaram, S.; Brown, D. R.; Drost, M. K.

    1992-11-01

    This report describes the results of an evaluation of thermal energy storage (TES) integrated with simple gas turbine cogeneration systems. The TES system captures and stores thermal energy from the gas turbine exhaust for immediate or future generation of process heat. Integrating thermal energy storage with conventional cogeneration equipment increases the initial cost of the combined system; but, by decoupling electric power and process heat production, the system offers the following significant advantages: (1) electric power can be generated on demand, irrespective of the process heat load profile, thus increasing the value of the power produced; (2) although supplementary firing could be used to serve independently varying electric and process heat loads, this approach is inefficient. Integrating TES with cogeneration can serve the two independent loads while firing all fuel in the gas turbine. The study evaluated the cost of power produced by cogeneration and cogeneration/TES systems designed to serve a fixed process steam load. The value of the process steam was set at the levelized cost estimated for the steam from a conventional stand-alone boiler. Power costs for combustion turbine and combined-cycle power plants were also calculated for comparison. The results indicated that peak power production costs for the cogeneration/TES systems were between 25 and 40 percent lower than peak power costs estimated for a combustion turbine and between 15 and 35 percent lower than peak power costs estimated for a combined-cycle plant. The ranges reflect differences in the daily power production schedule and process steam pressure/temperature assumptions for the cases evaluated. Further cost reductions may result from optimization of current cogeneration/TES system designs and improvement in TES technology through future research and development.

  9. Nonlinear Ultrasonic Diagnosis and Prognosis of ASR Damage in Dry Cask Storage

    SciTech Connect

    Qu, Jianmin; Bazant, Zdenek; Jacobs, Laurence; Guimaraes, Maria

    2015-11-30

    Alkali-silica reaction (ASR) is a deleterious chemical process that may occur in cement-based materials such as mortars and concretes, where the hydroxyl ions in the highly alkaline pore solution attack the siloxane groups in the siliceous minerals in the aggregates. The reaction produces a cross-linked alkali-silica gel. The ASR gel swells in the presence of water. Expansion of the gel results in cracking when the swelling-induced stress exceeds the fracture toughness of the concrete. As the ASR continues, cracks may grow and eventually coalesce, which results in reduced service life and a decrease safety of concrete structures. Since concrete is widely used as a critical structural component in dry cask storage of used nuclear fuels, ASR damage poses a significant threat to the sustainability of long term dry cask storage systems. Therefore, techniques for effectively detecting, managing and mitigating ASR damage are needed. Currently, there are no nondestructive methods to accurately detect ASR damage in existing concrete structures. The only current way of accurately assessing ASR damage is to drill a core from an existing structure, and conduct microscopy on this drilled cylindrical core. Clearly, such a practice is not applicable to dry cask storage systems. To meet these needs, this research is aimed at developing (1) a suite of nonlinear ultrasonic quantitative nondestructive evaluation (QNDE) techniques to characterize ASR damage, and (2) a physics-based model for ASR damage evolution using the QNDE data. Outcomes of this research will provide a nondestructive diagnostic tool to evaluate the extent of the ASR damage, and a prognostic tool to estimate the future reliability and safety of the concrete structures in dry cask storage systems

  10. Statistical Evaluation of Molecular Contamination During Spacecraft Thermal Vacuum Test

    NASA Technical Reports Server (NTRS)

    Chen, Philip; Hedgeland, Randy; Montoya, Alex; Roman-Velazquez, Juan; Dunn, Jamie; Colony, Joe; Petitto, Joseph

    1999-01-01

    The purpose of this paper is to evaluate the statistical molecular contamination data with a goal to improve spacecraft contamination control. The statistical data was generated in typical thermal vacuum tests at the National Aeronautics and Space Administration, Goddard Space Flight Center (GSFC). The magnitude of material outgassing was measured using a Quartz Crystal Microbalance (QCNO device during the test. A solvent rinse sample was taken at the conclusion of each test. Then detailed qualitative and quantitative measurements were obtained through chemical analyses. All data used in this study encompassed numerous spacecraft tests in recent years.

  11. Statistical Evaluation of Molecular Contamination During Spacecraft Thermal Vacuum Test

    NASA Technical Reports Server (NTRS)

    Chen, Philip; Hedgeland, Randy; Montoya, Alex; Roman-Velazquez, Juan; Dunn, Jamie; Colony, Joe; Petitto, Joseph

    1997-01-01

    The purpose of this paper is to evaluate the statistical molecular contamination data with a goal to improve spacecraft contamination control. The statistical data was generated in typical thermal vacuum tests at the National Aeronautics and Space Administration, Goddard Space Flight Center (GSFC). The magnitude of material outgassing was measured using a Quartz Crystal Microbalance (QCM) device during the test. A solvent rinse sample was taken at the conclusion of the each test. Then detailed qualitative and quantitative measurements were obtained through chemical analyses. All data used in this study encompassed numerous spacecraft tests in recent years.

  12. Evaluation of Instrumentation and Dynamic Thermal Ratings for Overhead Lines

    SciTech Connect

    Phillips, A.

    2013-01-31

    In 2010, a project was initiated through a partnership between the Department of Energy (DOE) and the New York Power Authority (NYPA) to evaluate EPRI's rating technology and instrumentation that can be used to monitor the thermal states of transmission lines and provide the required real-time data for real-time rating calculations. The project included the installation and maintenance of various instruments at three 230 kV line sites in northern New York. The instruments were monitored, and data collection and rating calculations were performed for about a three year period.

  13. Statistical Evaluation of Molecular Contamination During Spacecraft Thermal Vacuum Test

    NASA Technical Reports Server (NTRS)

    Chen, Philip; Hedgeland, Randy; Montoya, Alex; Roman-Velazquez, Juan; Dunn, Jamie; Colony, Joe; Petitto, Joseph

    1998-01-01

    The purpose of this paper is to evaluate the statistical molecular contamination data with a goal to improve spacecraft contamination control. The statistical data was generated in typical thermal vacuum tests at the National Aeronautics and Space Administration, Goddard Space Flight Center (GSFC). The magnitude of material outgassing was measured using a Quartz Crystal Microbalance (QCM) device during the test. A solvent rinse sample was taken at the conclusion of each test. Then detailed qualitative and quantitative measurements were obtained through chemical analyses. All data used in this study encompassed numerous spacecraft tests in recent years.

  14. Thermal performance evaluation of the Semco (liquid) solar collector

    NASA Technical Reports Server (NTRS)

    1979-01-01

    Procedures used and results obtained during the evaluation test program on a flat plate collector which uses water as the working fluid are discussed. The absorber plate is copper tube soldered to copper fin coated with flat black paint. The glazing consists of two plates of Lo-Iron glass; the insulation is polyurethane foam. The collector weight is 242.5 pounds with overall external dimensions of approximately 48.8 in. x 120.8 in. x 4.1 in. The test program was conducted to obtain thermal performance data before and after 34 days of weather exposure test.

  15. Evaluation of thermal gradients in longitudinal spin Seebeck effect measurements

    SciTech Connect

    Sola, A. Kuepferling, M.; Basso, V.; Pasquale, M.; Kikkawa, T.; Uchida, K.; Saitoh, E.

    2015-05-07

    In the framework of the longitudinal spin Seebeck effect (LSSE), we developed an experimental setup for the characterization of LSSE devices. This class of device consists in a layered structure formed by a substrate, a ferrimagnetic insulator (YIG) where the spin current is thermally generated, and a paramagnetic metal (Pt) for the detection of the spin current via the inverse spin-Hall effect. In this kind of experiments, the evaluation of a thermal gradient through the thin YIG layer is a crucial point. In this work, we perform an indirect determination of the thermal gradient through the measurement of the heat flux. We developed an experimental setup using Peltier cells that allow us to measure the heat flux through a given sample. In order to test the technique, a standard LSSE device produced at Tohoku University was measured. We find a spin Seebeck S{sub SSE} coefficient of 2.8×10{sup −7} V K{sup −1}.

  16. Integrated heat pipe-thermal storage system performance evaluation

    SciTech Connect

    Keddy, E.; Sena, J.T.; Merrigan, M.

    1987-01-01

    Performance verification tests of an integrated heat pipe-thermal energy storage system have been conducted. This system is being developed as a part of an Organic Rankine Cycle-Solar Dynamic Power System (ORC-SDPS) receiver for future space stations. The integrated system consists of potassium heat pipe elements that incorporate thermal energy storage (TES) canisters within the vapor space along with an organic fluid (toluene) heater tube used as the condenser region of the heat pipe. During the insolation period of the earth orbit, solar energy is delivered to the surface of the heat pipe elements of the ORC-SDPS receiver and is internally transferred by the potassium vapor for use and storage. Part of the thermal energy is delivered to the heater tube and the balance is stored in the TES units. During the eclipse period of the orbit, the stored energy in the TES units is transferred by the potassium vapor to the toluene heater tube. A developmental heat pipe element was fabricated that employs axial arteries and a distribution wick connecting the wicked TES units and the heater to the solar insolation surface of the heat pipe. Tests were conducted to verify the heat pipe operation and to evaluate the heat pipe/TES units/heater tube operation by interfacing the heater unit to a heat exchanger.

  17. Quality Considerations for the Evaluation of Thermal Spray Coatings

    NASA Astrophysics Data System (ADS)

    Wigren, Jan; Täng, Kristina

    2007-12-01

    Revealing the true structural and mechanical properties is of utmost importance for the optimized use of thermal sprayed coatings. Only the true properties can be expected to correlate to the spray parameters. During the recent decade, the gas turbine industry has experienced a focus on the laboratory procedures being the weakest link in a frozen and robust process. This article will show several results indicating that the laboratory procedures are more essential to the evaluation results than the spray parameters themselves. With new and robust laboratory techniques, the true properties of thermal spray coatings are revealed, causing a major problem with respect to the quality standards developed 30-40 years ago. In many cases, these old specifications need updates, which is a difficult task from a cost, time, and quality perspective for OEM’s. Coatings that have been successfully used for almost half a century no longer conform to the specification they were optimized to, because of these new appropriate laboratory techniques and procedures. What is actually meant when stating the following? (1) The coating has 5% porosity; (2) No cracks are allowed; (3) Tensile bond is 50 Mpa; (4) Hardness is 1000 HV; and (5) Coating thickness is 100 μm. This article also initiates a discussion on the measurement inaccuracies, for testing of thermally sprayed coatings, with respect to the commonly used general international standards (such as QS9000, ISO17025, AS9003, and ISO10012), as well as with respect to recommendations from the Six Sigma methodology.

  18. Thermal provocation to evaluate microvascular reactivity in human skin

    PubMed Central

    2010-01-01

    With increased interest in predictive medicine, development of a relatively noninvasive technique that can improve prediction of major clinical outcomes has gained considerable attention. Current tests that are the target of critical evaluation, such as flow-mediated vasodilation of the brachial artery and pulse-wave velocity, are specific to the larger conduit vessels. However, evidence is mounting that functional changes in the microcirculation may be an early sign of globalized microvascular dysfunction. Thus development of a test of microvascular reactivity that could be used to evaluate cardiovascular risk or response to treatment is an exciting area of innovation. This mini-review is focused on tests of microvascular reactivity to thermal stimuli in the cutaneous circulation. The skin may prove to be an ideal site for evaluation of microvascular dysfunction due to its ease of access and growing evidence that changes in skin vascular reactivity may precede overt clinical signs of disease. Evaluation of the skin blood flow response to locally applied heat has already demonstrated prognostic utility, and the response to local cooling holds promise in patients in whom cutaneous disorders are present. Whether either of these tests can be used to predict cardiovascular morbidity or mortality in a clinical setting requires further evaluation. PMID:20507974

  19. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    SciTech Connect

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions.

  20. Standard review plan for reviewing safety analysis reports for dry metallic spent fuel storage casks

    SciTech Connect

    Not Available

    1988-01-01

    The Cask Standard Review Plan (CSRP) has been prepared as guidance to be used in the review of Cask Safety Analysis Reports (CSARs) for storage packages. The principal purpose of the CSRP is to assure the quality and uniformity of storage cask reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The CSRP also sets forth solutions and approaches determined to be acceptable in the past by the NRC staff in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a CSAR does not have to follow the solutions or approaches presented in the CSRP. However, applicants should recognize that the NRC staff has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches.

  1. 27. STAINLESS STEEL FERMENTING CASKS MADE BY ZERO MANG OF ...

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

    27. STAINLESS STEEL FERMENTING CASKS MADE BY ZERO MANG OF WASHINGTON, MISSOURI. VIEW LOOKING NORTH TOWARD VAULT OF THE TWELVE APOSTLES - Stone Hill Winery, 401 West Twelfth Street, Hermann, Gasconade County, MO

  2. Feasibility study for a transportation operations system cask maintenance facility

    SciTech Connect

    Rennich, M.J.; Medley, L.G.; Attaway, C.R.

    1991-01-01

    The US Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the development of a waste management program for the disposition of spent nuclear fuel (SNF) and high-level waste (HLW). The program will include a transportation system for moving the nuclear waste from the sources to a geologic repository for permanent disposal. Specially designed casks will be used to safely transport the waste. The cask systems must be operated within limits imposed by DOE, the Nuclear Regulatory Commission (NRC), and the Department of Transportation (DOT). A dedicated facility for inspecting, testing, and maintaining the cask systems was recommended by the General Accounting Office (in 1979) as the best means of assuring their operational effectiveness and safety, as well as regulatory compliance. In November of 1987, OCRWM requested a feasibility study be made of a Cask Maintenance Facility (CMF) that would perform the required functions. 46 refs., 16 figs., 13 tabs.

  3. ALARA studies on spent fuel and waste casks

    SciTech Connect

    Sutherland, S.H.

    1980-04-01

    In this report, some implications of applying the ALARA concept to cask designs for transporting spent fuel, high-level commercial and defense waste, and remote-handled transuranic waste are investigated. The XSDRNPM, one-dimensional radiation transport code, was used to obtain potential shield designs that would yield total dose rates at 1.8 m from the cask surface of 10, 5, and 2 mrem/h. Gamma shields of depleted uranium, lead, and steel were studied. The capacity of the casks was assumed to be 1, 4, or 7 elements or canisters, and the wastes were 1, 3, 5, and 10 years old. Depending on the dose rate, the cask empty weights and lifetime transportation costs were estimated.

  4. Selection and Evaluation of Thermal Interface Materials for Reduction of the Thermal Contact Resistance of Thermoelectric Generators

    NASA Astrophysics Data System (ADS)

    Sakamoto, Tatsuya; Iida, Tsutomu; Sekiguchi, Takeshi; Taguchi, Yutaka; Hirayama, Naomi; Nishio, Keishi; Takanashi, Yoshifumi

    2014-10-01

    A variety of thermal interface materials (TIMs) were investigated to find a suitable TIM for improving the performance of thermoelectric power generators (TEGs) operating in the medium-temperature range (600-900 K). The thermal resistance at the thermal interface between which the TIM was inserted was evaluated. The TIMs were chosen on the basis of their thermal stability when used with TEGs operating at medium temperatures, their electrical insulating properties, their thermal conductivity, and their thickness. The results suggest that the boron nitride (BN)-based ceramic coating, Whity Paint, and the polyurethane-based sheet, TSU700-H, are suitable TIMs for the heat source and heat sink sides, respectively, of the TEG. Use of these effectively enhances TEG performance because they reduce the thermal contact resistance at the thermal interface.

  5. NAC-1 cask dose rate calculations for LWR spent fuel

    SciTech Connect

    CARLSON, A.B.

    1999-02-24

    A Nuclear Assurance Corporation nuclear fuel transport cask, NAC-1, is being considered as a transport and storage option for spent nuclear fuel located in the B-Cell of the 324 Building. The loaded casks will be shipped to the 200 East Area Interim Storage Area for dry interim storage. Several calculations were performed to assess the photon and neutron dose rates. This report describes the analytical methods, models, and results of this investigation.

  6. SPENT FUEL CASK IMPACT LIMITER ATTACHMENT DESIGN DEFICIENCIES

    SciTech Connect

    Leduc, D; Jeffery England, J

    2007-10-16

    A recent structural analysis of the T-3 Spent Fuel Containment Cask found problems with the design of the attachment system. Assumptions in the original SARP concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. Similar weaknesses in the attachment system designs of other casks were also noted. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.

  7. Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

    SciTech Connect

    Allen, G.C.; Beck, D.F.; Harmon, C.D.; Shipers, L.R.

    1992-08-01

    Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program. 2 refs.

  8. Evaluation of protective ensemble thermal characteristics through sweating hot plate, sweating thermal manikin, and human tests.

    PubMed

    Kim, Jung-Hyun; Powell, Jeffery B; Roberge, Raymond J; Shepherd, Angie; Coca, Aitor

    2014-01-01

    The purpose of this study was to evaluate the predictive capability of fabric Total Heat Loss (THL) values on thermal stress that Personal Protective Equipment (PPE) ensemble wearers may encounter while performing work. A series of three tests, consisting of the Sweating Hot Plate (SHP) test on two sample fabrics and the Sweating Thermal Manikin (STM) and human performance tests on two single-layer encapsulating ensembles (fabric/ensemble A = low THL and B = high THL), was conducted to compare THL values between SHP and STM methods along with human thermophysiological responses to wearing the ensembles. In human testing, ten male subjects performed a treadmill exercise at 4.8 km and 3% incline for 60 min in two environmental conditions (mild = 22°C, 50% relative humidity (RH) and hot/humid = 35°C, 65% RH). The thermal and evaporative resistances were significantly higher on a fabric level as measured in the SHP test than on the ensemble level as measured in the STM test. Consequently the THL values were also significantly different for both fabric types (SHP vs. STM: 191.3 vs. 81.5 W/m(2) in fabric/ensemble A, and 909.3 vs. 149.9 W/m(2) in fabric/ensemble B (p < 0.001). Body temperature and heart rate response between ensembles A and B were consistently different in both environmental conditions (p < 0.001), which is attributed to significantly higher sweat evaporation in ensemble B than in A (p < 0.05), despite a greater sweat production in ensemble A (p < 0.001) in both environmental conditions. Further, elevation of microclimate temperature (p < 0.001) and humidity (p < 0.01) was significantly greater in ensemble A than in B. It was concluded that: (1) SHP test determined THL values are significantly different from the actual THL potential of the PPE ensemble tested on STM, (2) physiological benefits from wearing a more breathable PPE ensemble may not be feasible with incremental THL values (SHP test) less than approximately 150-200 W·m(2), and (3) the

  9. TRANSPORTATION CASK RECEIPT AND RETURN FACILITY WORKER DOSE ASSESSMENT

    SciTech Connect

    V. Arakali

    2005-02-24

    The purpose of this design calculation is to estimate radiation doses received by personnel working in the Transportation Cask Receipt and Return Facility (TCRRF) of the repository including the personnel at the security gate and cask staging areas. This calculation is required to support the preclosure safety analysis (PCSA) to ensure that the predicted doses are within the regulatory limits prescribed by the U.S. Nuclear Regulatory Commission (NRC). The Cask Receipt and Return Facility receives NRC licensed transportation casks loaded with spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The TCRRF operation starts with the receipt, inspection, and survey of the casks at the security gate and the staging areas, and proceeds to the process facilities. The transportation casks arrive at the site via rail cars or trucks under the guidance of the national transportation system. This calculation was developed by the Environmental and Nuclear Engineering organization and is intended solely for the use of Design and Engineering in work regarding facility design. Environmental and Nuclear Engineering personnel should be consulted before using this calculation for purposes other than those stated herein or for use by individuals other than authorized personnel in the Environmental and Nuclear Engineering organization.

  10. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.