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Sample records for cask thermal evaluation

  1. (Validation of) computational fluid dynamics modeling approach to evaluate VSC-17 dry storage cask thermal designs

    International Nuclear Information System (INIS)

    This paper presents results from a numerical analysis of the thermal evaluation of a Ventilated Concrete Storage Cask VSC-17 system. Three-dimensional simulations are performed for the VSC-17 system, and the results are compared to experimental data. The VSC-17 is a concrete-shielded spent nuclear fuel (SNF) cask system designed to contain 17 pressurized water reactor (PWR) fuel assemblies for storage and transportation. The system consists of a ventilated concrete cask (VCC) and a multi-assembly sealed basket (MSB). The VCC is a concrete cylindrical vessel, fabricated as a single piece and fitted with a flat plate at the bottom. The concrete cask provides structural support, shielding, and natural convection cooling for the MSB. The MSB has an outer steel shell and an inner fuel guide sleeve assembly that holds canisters containing spent fuel rods. Cooling airflow inside the concrete cask is driven by natural convection. Heat transfer in the cask is a complicated process because of the inherent complexity of the geometry and the fixed and natural convection induced by the radioactive decay process. Other factors that contribute to the overall heat transfer include the heat generation by the spent fuel, the thermal boundary condition, the filling medium within the MSB, and the vertical or horizontal orientation of the cask. Proper thermal analysis of dry storage casks is important for accurate estimation of the peak fuel temperature and peak cladding temperature (PCT). Proper estimation of PCT ensures the integrity of cladding and is important for safety evaluation of independent spent fuel storage installations. Accurate estimation of the peak fuel temperature and peak cladding temperature ensures the integrity of the cladding. The spent nuclear fuel may be exposed to air and oxidize if the cladding is damaged and thus increase the potential for release of radioactivity. In the current analysis, numerical simulations are carried out using the computational fluid

  2. Spent Fuel Dry Storage Cask Thermal Test

    International Nuclear Information System (INIS)

    Most nuclear power plants maintain their spent fuel discharged at a reactor in wet storage pools. However, after several years of use, many pools are filled to capacity. Therefore, finding a sufficient capacity for storage is essential because of the continued delays in obtaining a safe, interim storage facility if nuclear power plants are to be allowed to continue to operate. Dry storage cask will be one solution for solving an interim storage problem. The dry storage cask consists of two separate components: an over-pack, and a canister. The structure strength part of the over-pack is made of carbon steel, and the inner cavity of the structure strength part is filled with concrete, which accomplishes the role as a radiation shield. The outer diameter of the dry storage cask is 3,550 mm and the its overall height is 5,885 mm. It weighs approximately 135 tons. The dry storage cask accommodates 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat from the 24 PWR spent fuel assemblies is 25.2 kW This paper discusses the experimental approach used to evaluate the heat transfer characteristics of the dry storage cask

  3. CASKET: a computer code system for thermal and structural analyses of radioactive material transport and/or storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-05-01

    A computer code system CASKET (CASK thermal and structural analyses and Evaluation code system) for the thermal and structural analyses which are indispensable for radioactive material transport and/or storage cask designs has been developed. The CASKET is a simplified computer code system to perform parametric analyses on sensitivity evaluations in designing a cask and conducting its safety analysis. Main features of the CASKET are as follow: (1) it is capable to perform impact analysis of casks with shock absorbers, (2) it is capable to perform impact analysis of casks with fins. (3) puncture analysis of casks is capable, (4) rocking analysis of casks during seismic load is capable, (5) material property data library are provided for impact analysis of casks, (6) material property data library are provided for thermal analysis of casks, (7) fin energy absorption data library are provided for impact analysis of casks with fins are and (8) not only main frame computers (OS MSP) but also work stations (OS UNIX) and personal computers (OS Windows 3.1) are available. In the paper, brief illustrations of calculation methods are presented. Some calculation results are compared with experimental ones to confirm the computer programs are useful for thermal and structural analyses. (author)

  4. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

    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

  5. Safety evaluation for packaging (onsite) SERF cask

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Economic evaluation of nuclear waste transportation casks

    International Nuclear Information System (INIS)

    A method is described which allows the systematic economic evaluation of transportation cask designs which meet the requirements of the Test and Evaluation Facility (TEF) program. The heart of the method described is the Waste Management Transportation Model. This model uses a set of computer-based algorithms to assemble specific case information input, combine this input with the data base of transportation information maintained within the model, and calculate the cask types and quantities necessary, the cask utilization factors, and the total costs for each transport line specified. The model is capable of handling a large variety of transportation problems given the specific input related to each type. Three combinations of waste packaging facilities were examined. The first assumes all consolidation and packaging occurs at an existing hot cell. The second assumes all consolidation and packaging is done at the TEF site. The third combination assumes that spent fuels are consolidated at an existing hot cell while waste packaging occurs at the TEF site. Some of the general findings are: (1) defense high-level waste (DHLW) is generally lower in cost than SF as the prime waste form because of the fewer number of shipments required prior to the waste consolidation activity; (2) when DHLW is the prime waste form, it is beneficial to locate the packaging facility (PF) close to the TEF site because the packaged waste form is heavier, more costly to transport; (3) when SF is the prime waste form, it is beneficial to locate the PF close to the waste source to reduce the length of the transport links containing unconsolidated spent fuel assemblies; and (4) truck casks, and legal weight truck casks in particular, are generally superior to the rail casks on an economic basis

  7. Thermal test and analysis of a spent fuel storage cask

    International Nuclear Information System (INIS)

    A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

  8. Conceptual evaluation of metal storage cask for conditioned spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Seo, K. S.; Shin, H. S.; Lee, J. C.; Bang, K. S.; Kim, H. D.; Park, S. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-07-01

    The storage parameters of spent PWR fuel are radioactivity, heat power and its volume. Those values could be reduced to about a quarter by an Advanced spent fuel Conditioning Process (ACP). Firstly, a storage concept and scenario were established considering the characteristics of conditioned spent fuel. If the efficiency of the cooling system is improved and the appropriate quantities of the conditioned spent fuel are stored, the conditioned spent fuels could be stored at a four times higher level of spent fuel storage. One storage unit of conditioned spent fuel was designed to have its capacity equivalent to one PWR spent fuel. It was supposed that a metal storage cask has 7 baskets that can load 28 storage units. Those capacities means that 28 spent PWR fuels in metal casks can be stored. The conceptual evaluations of the critical, shielding, thermal and structural fields were performed. In conclusion, the results of the conceptual evaluations show that the proposed metal cask satisfied with the important design criteria at a smaller size than the existing systems.

  9. Evaluation of improvement potential for spent fuel cask handling

    International Nuclear Information System (INIS)

    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

  10. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    International Nuclear Information System (INIS)

    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

  12. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Numerical simulation of ambient flow and thermal distributions in a spent fuel storage cask array

    Energy Technology Data Exchange (ETDEWEB)

    Michener, T. [Pacific Northwest National Laboratory, Richland WA (United States); Trent, D.S.; Guttmann, J.; Bajwa, C. [United States Nuclear Regulatory Commission, One White Flin North, Rockville MD (United States)

    2001-07-01

    At the request of the U.S. Nuclear Regulatory Commission (USNRC), the staff at the Pacific Northwest National Laboratory (PNNL) analyzed the thermal performance of the Utah Private Fuel Storage (PFS) using the TEMPEST computational fluid dynamics software. A three-dimensional section of the PFS with a total of 20 casks was modeled to estimate the ambient flow and temperature distributions surrounding the casks. The purpose of this analysis was to compute the cask inlet vent air temperature to be used for boundary conditions in a detailed analysis of an individual Holtec Hi-Storm 100 cask using the COBRA-SFS (Spent Fuel Storage) thermal hydraulic computer software. (author)

  15. Structural evaluation and analysis under normal conditions for spent fuel concrete storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Taechul; Baeg, Changyeal; Yoon, Sitae [Korea Radioactive waste Management Agency, Daejeon (Korea, Republic of); Jung, Insoo [Korea Nuclear Engineering and Service Co., Daejeon (Korea, Republic of)

    2014-05-15

    The purpose of this paper is the verification of stabilities of the structural elements that influence the safety of a concrete storage cask. The evaluation results were reviewed with respect to every design criterion, in terms of whether the results satisfy the criteria, provided by 10CFR 72 and NUREG-1536. The basic information on the design is partially explained in 2. Description of spent fuel storage system and the maintainability and assumptions included in the analysis were confirmed through detailed explanations of the acceptable standards, analysis model, and analysis method. ABAQUS 6.10, a widely used finite element analysis program, was used in the structural analysis. The storage cask shall maintain the sub-criticality, shielding, structural integrity, thermal capability and confinement in accordance with the requirements specified in US 10 CFR 72. The safety of storage cask is analyzed and it has been confirmed to meet the requirements of US 10 CFR 72. This paper summarizes the structural stability evaluation results of a concrete storage cask with respect to the design criteria. The evaluation results of this paper show that the maximum stress was below the allowable stress under every condition, and the concrete storage cask satisfied the design criteria.

  16. Structural evaluation and analysis under normal conditions for spent fuel concrete storage cask

    International Nuclear Information System (INIS)

    The purpose of this paper is the verification of stabilities of the structural elements that influence the safety of a concrete storage cask. The evaluation results were reviewed with respect to every design criterion, in terms of whether the results satisfy the criteria, provided by 10CFR 72 and NUREG-1536. The basic information on the design is partially explained in 2. Description of spent fuel storage system and the maintainability and assumptions included in the analysis were confirmed through detailed explanations of the acceptable standards, analysis model, and analysis method. ABAQUS 6.10, a widely used finite element analysis program, was used in the structural analysis. The storage cask shall maintain the sub-criticality, shielding, structural integrity, thermal capability and confinement in accordance with the requirements specified in US 10 CFR 72. The safety of storage cask is analyzed and it has been confirmed to meet the requirements of US 10 CFR 72. This paper summarizes the structural stability evaluation results of a concrete storage cask with respect to the design criteria. The evaluation results of this paper show that the maximum stress was below the allowable stress under every condition, and the concrete storage cask satisfied the design criteria

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. An economic evaluation of a storage system for casks with burnup credit

    International Nuclear Information System (INIS)

    It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

  19. Structural evaluation of spent fuel dry storage cask

    International Nuclear Information System (INIS)

    In a various regulations and standards related to the spent fuel storage, the storage casks should be designed to sustain the structural integrity under the accident conditions of predicted operation and design criteria. These conditions for the structural evaluation requires the drop, tip-over, wind like tornado and typhoon, flood and earthquake. This paper describes the load cases and conceptual evaluation method for the structural evaluation. Preliminary safety analysis of the concrete storage system were performed

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

    International Nuclear Information System (INIS)

    The development of the Source Term Analyses for Containment Evaluations (STACE) methodology provides a unique means for estimating the probability of cladding breach within transport casks, quantifying the amount of radioactive material released into the cask interior, and calculating the releasable radionuclide concentrations and corresponding maximum permissible leakage rates. 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 team 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, volitile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking in which experimental validation is planned. Finally, the ANSI N14.5 recommendation that 3% and 100% of the fuel rods fail during normal and hypothetical accident conditions of transport, respectively, has been show to be overly conservative by several degrees of magnitude for these example analyses. Furthermore, the maximum permissible leakage rates for this example assembly under normal and hypothetical accident conditions are significanly higher that the leaktight requirements. By relaxing the maximum permissible leakage rates, the source term methodology is expected to significantly improvecask economics and safety

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

  3. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    International Nuclear Information System (INIS)

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  4. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-26

    The purpose of this design calculation is to demonstrate that the handling operations of transportation casks performed in the Transportation Cask Receipt and Return Facility (TCRRF) and Buffer Area meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC [Bechtel SAIC Company] 2004 [DIRS 171599], Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''Transportation Cask Receipt/Return Facility Description Document'' (BSC 2004 [DIRS 170217], Section 3.2.3). Specific scope of work contained in this activity consists of the following items: (1) Evaluate criticality effects for both dry and fully flooded conditions pertaining to TCRRF and Buffer Area operations for defense in depth. (2) Evaluate Category 1 and 2 event sequences for the TCRRF as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). This evaluation includes credible fuel reconfiguration conditions. In addition to the scope of work listed above, an evaluation was also performed of modeling assumptions for commercial spent nuclear fuel (CSNF) regarding inclusion of plenum and end regions of the active fuel. This calculation is limited to CSNF and US Department of Energy (DOE) SNF. it should be mentioned that the latter waste form is evaluated more in depth in the ''Canister Handling Facility Criticality Safety Calculations (BSC 2004 [DIRS 167614]). Further, the design and safety analyses of the naval SNF canisters are the responsibility of the US Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the TCRRF and Buffer Area and may not reflect the ongoing design evolution of the facility. However, it is anticipated that design changes to the facility layout will

  5. A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation cask

    International Nuclear Information System (INIS)

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. To overcome modeling difficulties arising from the complexity of geometry in large PWR metal casks, a multiple cylinder model is used to calculate the temperature profile of a cylindrical cask body in the first step analysis. In the second step analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three-dimensional conduction analysis model. An existing HEATING 7.2f code has been used in the present two-step numerical analyses. Effects of aluminium heat transfer fin and the cask ambient conditions on the maximum fuel temperature have been examined as a parametric study. A comparison between the predicted maximum fuel temperature and the data of Nuclear Assurance Corporation Storage and Transportation Canister Safety Analysis Report (NAC-STC SAR) shows good agreement

  6. A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

    International Nuclear Information System (INIS)

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

  7. Evaluation of Impact Resistance of Concrete Overpack of Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sanghoon; Kim, Ki-Young; Jeon, Je-Eon; Seo, Ki-Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The concrete overpack of the cask provides radiation shielding as well as physical protection for inner canister against external mechanical shock. When the overpack undergoes a severe missile impact which might be caused by tornado or aircraft crash, it should sustain minimal level of structural integrity so that the radiation shielding and the retrievability of canister are maintained. Empirical formulas have been developed for the evaluation of concrete damage but those formulas can be used only for local damage evaluation and not for global damage evaluation. In this research, a series of numerical simulations and tests have been performed to evaluate the damage of two types of concrete overpack segment models under high speed missile impact. It is shown that appropriate modeling of material failure is crucial in this kind of analyses and finding the correct failure parameters may not be straightforward. When comparing the simulation results with the test results, it is shown that neither setting, case 1 and 2 provides results with consistent agreement with test results. That is, case 1 setting is more close to reality in Type 1 model analysis, but for Type 2, case 2 setting provides more close results to the reality. In both the case, not enough deformation is predicted by simulation compared to the tests. Weak failure and eroding criteria give larger penetration depth with insufficient overall damage due to energy loss with element erosion.

  8. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear and Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Choi, Kyung Joo; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun; Kim, Seong Jong [Chungnam National Univ., Taejon (Korea, Republic of)

    2002-03-15

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology.

  9. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology

  10. Drop accident analyses of dry metal cask without impact limiter and evaluation of leak rate

    International Nuclear Information System (INIS)

    In Japan, utilities are preparing to commence a private Interim Storage Facility (ISF) of spent fuels in Mutucity, Aomori prefecture from 2010. In the ISF, metal casks for transport and storage will be adopted and handled without an impact limiter. Cask drop tests without the impact limiter using an actual size simulated cask were carried out by CRIEPI (Central Research Institute of Electric Power Industry) in 2005. Then cases of cask drop tests were analyzed and the leak rate characteristics of a metal gasket were investigated. A general non-linear dynamic simulation computer code LS-DYNA was used in analyses. The collision velocity of the cask was calculated assuming free drop from an initial position for both horizontal drop and rotational drop. Although the drop height was 1 m in the tests, it was changed to 1.5 m and 2.0 m as parameters in the calculation for investigation of the leak rate characteristic. It was supposed that the increase of the leak rate was not only due to an increase of the total sliding movement of the lid but also caused by plastic deformation of flange or bolts. A correlation curve between total sliding movement of lid and leak rate was settled for leak rate of cask drops without the impact limier, based on results of the previous test using small-scale sized model (small scale test). Under these postulations, the leak rate could be evaluated by the correlation curve and obtained total sliding movement of the lid. In the simulated cask used for the test, a clearance between the lid and the cask body was small and the total sliding movement was limited. The leak rate estimation methodology would be applicable to the actual cask drop accident without the impact limiter, if the plastic deformation were not occurred at the flange. (author)

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

    International Nuclear Information System (INIS)

    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

  12. Thermal analysis of spent fuel shipping cask for application of metalized fuel

    International Nuclear Information System (INIS)

    Thermal analysis of spent fuel shipping cask loaded with 4 spent PWR fuel assemblies has been carried out using the fluent code. And the temperature distribution of cask for application of 4 metalized fuels equivalent to 16 PWR fuels has been also calculated. Total decay heat from 4 spent PWR fuels and 4 metalized spent fuels are 2.2 kW and 4.4 kW, respectively. The calculated temperatures for 4 spent PWR fuels were compared with the proven data presented from the safety analysis report of shipping cask. It has good agreement between two results. The maximum fuel rod temperatures inside the canisters of square and hexagonal types are estimated to be 269 .deg. C and 212 .deg. C, respectively. Therefore, it is found that the hexagonal canister loaded with metalized fuel rods is more advantageous in aspect of thermal characteristics and storage efficiency. Fuel temperature in the cavity of helium gas for hexagonal canister is lower than the temperature for spent PWR fuel

  13. Drop accident analyses of dry metal cask without impact limiter and evaluation of leak rate

    International Nuclear Information System (INIS)

    rotational impact. Although the drop height was 1 m in both horizontal drop and rotational impact tests, the drop height was changed to 1.5 m and 2.0 m as a parameter in the calculation. The increase of the leak rate is not only due to the increase of the total sliding movement of the lid but also suspected to be caused by plastic deformation of flange or bolts. The applicability of the correlation curve between the total sliding movement of the lid and leak rate postulates that structural parts composing the leak tightness of the lid maintain the integrity without plastic deformation even after the event. Using this correlation and the total sliding movement of the lid, the leak rate could be evaluated. In the evaluation, the difference in hoop diameter, and effects of aging and elapsed time between the event and the measurement should be corrected. Data is shown, obtained from the small scale tests for seal performance of lid with aged metal gasket, the correlation curve with 95 % confidence level and the calculated total sliding movement. For the simulated cask used for the test, the clearance between the lid and the cask body is small and the total sliding movement is limited. Therefore, increase of the leak rate due to large total sliding movement is difficult to occur. In the rotational impact event, the top of the cask near the seal collides against the floor. In the calculation for the 2 m rotational impact case, plastic deformation was observed around the top of the cask but not near the flange. The present calculations were done under room temperature conditions. The leak rate estimation methodology is applicable to the actual cask drop accident, if plastic deformation does not occur near the flange

  14. Design and operational experience of dry cask storage systems

    International Nuclear Information System (INIS)

    This paper (Power Point presentation) describes cask storage design features and available dry cask storage technology, cask types used for dry storage, design characteristics of CASTOR casks, the German licensing basis for cask storage systems, shielding requirements, thermal layout, mechanical design, criticality safety and containment, licensing procedure, operational experience of dry cask storage in Germany and worldwide

  15. Development of strain gauge evaluation channels for use in dynamic testing of shipping casks

    International Nuclear Information System (INIS)

    The Transportation System Development Department at Sandia National Laboratories (SNL) frequently evaluates the structural response of casks being developed to transport radioactive materials. A major part of this activity includes gathering instrumentation data from dynamic impact tests of cask models. The acquisition of reliable, high-quality instrumentation data is an important component of cask certification. One method to evaluate instrumentation error during testing is to include evaluation channels for the various structural transducers. Evaluation channels have been produced by some manufacturers of accelerometers used for structural evaluations of casks and are commercially available. These particular devices produce very low output or no output to applied shock acceleration. However, it was found that a packaged strain gauge evaluation channel is not commercially available. Consequently, strain gauge evaluation channels have been developed at SNL to evaluate non-strain-induced resistance changes from environmental factors that could affect resistance strain measurement data. These unwanted nonstrain-induced resistance changes could be caused, for example, by resistance changes in the interconnecting cabling, electromagnetic noise, or grounding effects

  16. Safety analysis report for EPMA irradiated specimen cask

    Energy Technology Data Exchange (ETDEWEB)

    Ku, J. H.; Lee, J. C.; Seo, K. S.; Bang, K. S.; Park, S. W.; Min, D. K

    2000-11-01

    For the effective examination of spent fuels and radioactive materials by using EPMA in IMEF besides using SEM in PIEF, a special purpose EPMA cask was developed. It will be used to transport a specimen from the hot-cell in PIEF to the shielded glove box in IMEF. This cask should be easy to handle and transport by hand carry. It also has to be safe to maintain the shielding safety as well as the thermal and structural integrities under prescribed load conditions by the regulatory requirements. This cask was designed compactly to be docked perfectly maintaining shielding integrity without the modification of the interfaces of hot-cell and shielded glove box. Accordingly, the main features of cask were analyzed with functional capabilities, and the integrities of cask under required load conditions were evaluated. It was verified that the EPMA cask is suitable to use at handy transport of irradiated specimen between the PIEF and IMEF facilities in KAERI.

  17. 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)

    International Nuclear Information System (INIS)

    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

  18. POST-CASKETSS: a graphic computer program for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    A computer program POST-CASKETSS has been developed for the purpose of calculation result representation for thermal and structural analysis computer code system CASKETSS (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of POST-CASKETSS are as follows; (1) Function of calculation result representation for thermal and structural analysis computer programs is provided in the program. (2) Two and three dimensional graphic representation for finite element and finite difference programs are available in the program. (3) The capacity of graphics of geometry, temperature contor and temperature-time curve are provided for thermal analysis. (4) The capacity of graphics of geometry, deformation, stress contor, displacement-time curve, velocity-time curve, acceleration-time curve, stress-time curve, force-time curve and moment-time curve are provided for structural analysis. (5) This computer program operates both the time shearing system and the batch system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  19. THERMLIB: a material property data library for thermal analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    The paper describes an heat conduction data library and graphical program for analysis of radioactive material transport casks. More than 1000 of material data are compiled in the data library which was produced by Lawrence Livermore Laboratory. Thermal data such as, density, thermal conductivity, specific heat, phase-change or solid-state, transition temperature and latent heat have been tabulated. Using this data library, a data library processing program THERMLIB for thermal analysis has been developed. Main features of THERMLIB are as follows: (1) data have been tabulated against temperature, (2) more than 1000 material data are available, (3) it is capable of graphical representations for thermal data and (4) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of THERMLIB. In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides an user`s guide for computer program and input data for THERMLIB. (author)

  20. THERMLIB: a material property data library for thermal analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    The paper describes an heat conduction data library and graphical program for analysis of radioactive material transport casks. More than 1000 of material data are compiled in the data library which was produced by Lawrence Livermore Laboratory. Thermal data such as, density, thermal conductivity, specific heat, phase-change or solid-state, transition temperature and latent heat have been tabulated. Using this data library, a data library processing program THERMLIB for thermal analysis has been developed. Main features of THERMLIB are as follows: (1) data have been tabulated against temperature, (2) more than 1000 material data are available, (3) it is capable of graphical representations for thermal data and (4) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of THERMLIB. In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides an user's guide for computer program and input data for THERMLIB. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  3. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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

  4. A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit

    International Nuclear Information System (INIS)

    The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years

  5. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  7. Evaluation of stress corrosion cracking in aqueous solution neutron shield of transport/storage cask for spent fuel

    International Nuclear Information System (INIS)

    Experimental evaluation proved that no chloride induced stress corrosion cracking will occur on the metal cask which utilizes propylene glycol aqueous solution as neutron shield. Crevice corrosion, precursor of cracking, occurs at about 0.4V vs. 0.1M-KCl silver silver-chloride reference electrode in aqueous solution with chloride concentration of more than 5 times higher than limit value. On the other hand, the electrochemical potential (ECP) of cask material was 0.08V in air saturated aqueous solution. Since ECP is much smaller than the crevice corrosion potential below which no crevice corrosion is expected, the possibility is very small for chloride induced stress corrosion cracking to occur on the cask. (author)

  8. Evaluation of Equivalent Dose Rate of Interim Dry Storage Casks Loaded with Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Equivalent dose rate calculations of the CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks were performed using SCALE 4.3 computer codes system. These casks are planned for an interim storage of spent nuclear fuel at Ignalina NPP. The dose rate calculations were made on the sidelong, upper and lower surface of the cask and at the certain distance. Results show that dose rate values on the surface of the cask are much less then permissible value 1000 μSv/h when average burnup of fuel assembly is 20 GWd/tU. (author)

  9. Thermal hydraulic and neutronic analysis of dry cask storage systems for spent nuclear fuels

    International Nuclear Information System (INIS)

    Interim spent fuel storage systems must provide for the safe receipt, handling, retrieval and storage of spent nuclear fuel before reprocessing or disposal. In the context of achieving these objectives, the following features of the design were taken into consideration for metal shielded type storage systems; to maintain fuel subcritical, to remove spent fuel residual heat, to provide for radiation protection. These features in the design of a dry cask storage system were analyzed by employing COBRA-SFS and SCALE4.4 (ORIGEN, XSDOSE, CSAS6 ) codes for normal operation of the system under study. In accordance with safety assurance limits of International Atomic Energy Authority (IAEA), appropriate designs for Dry Cask Storage Systems (DCSS) were reached for 33000, 45000, and 55000 MWd/t burnup values and 5 and 10 years of cooling periods for spent fuel to be stored (Table 1)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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

  12. Evaluation of computer programs used for structural analyses of impact response of spent fuel shipping casks

    International Nuclear Information System (INIS)

    This report presents the results of a study of impact analyses of a generic spent-fuel cask. The study compares the use and results of three different finite element computer codes. Seven different cask-like model analyses are considered. The models encompass both linear and nonlinear geometric and material behavior. On the basis of the analyses results, this report recommends what parameters are useful in the comparison of different structural finite element computer programs. 5 references, 36 figures, 11 tables

  13. Evaluation of computer programs used for structural analyses of impact response of spent fuel shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B A; Gwinn, K W

    1984-05-01

    This report presents the results of a study of impact analyses of a generic spent-fuel cask. The study compares the use and results of three different finite element computer codes. Seven different cask-like model analyses are considered. The models encompass both linear and nonlinear geometric and material behavior. On the basis of the analyses results, this report recommends what parameters are useful in the comparison of different structural finite element computer programs. 5 references, 36 figures, 11 tables.

  14. Comparisons of prediction methods for peak cladding temperature and effective thermal conductivity in spent fuel assemblies of transportation/storage casks

    International Nuclear Information System (INIS)

    Highlights: • Peak cladding temperature (PCT) of spent fuel were evaluated by various methods. • The methods are Wooton–Epstein correlation, two-region model, and CFD. • Temperature difference between two-region and CFD ranges from −0.2 to 9 K. • CFD could be used to calculate PCT because of over-predicting PCT of two-region. • Application using CFD was conducted for spent fuel assembly used in Republic of Korea. - Abstract: When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature. There are three models to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas’s two-region model, Bahney Lotz’s effective thermal conductivity model, and Wooton–Epstein correlation. The peak cladding temperatures of Babcock and Wilcox (B and W) 15 × 15 PWR spent fuel assembly under helium backfill gas were evaluated by using two-dimensional CFD simulation and compared with two models (Wooton–Epstein correlation, two-region model). The peak cladding temperature difference between the two-region model and CFD simulation ranges from −0.2 K to 9 K. Two-region model over-predicts the measured peak cladding temperature that performs in a spent fuel dry storage cask. Therefore the simulation could be used to calculate peak cladding temperature of spent fuel assemblies. Application using CFD simulation was conducted to investigate the peak cladding temperature and effective thermal conductivity of spent fuel assembly used in Korea NPPs: 16 × 16 (CE type) and 17 × 17 (WH type) PWR spent fuel assembly. CFD simulation results are similar to each other, and the difference of temperature drop between the three arrays occurs slightly in all basket wall temperatures. The effective thermal conductivity calculated from the 16 × 16 PWR spent fuel assembly results was more conservative

  15. Dry interim spent fuel storage casks. Licensing, evaluation and operational experience

    International Nuclear Information System (INIS)

    The German concept for the external dry interim storage of spent fuel and high level wastes is based on the used of monolithic ductile iron casks which are licensed according to the transport regulations and the national Atomic Energy Act. The casks ensure the safe confinement of the radioactive inventory over long term storage periods of up to 40 years. Essential for that purpose is the double barrier containment system, consisting of two independent lids sealed with long term resistant metallic gaskets and equipped with an interspace pressure monitoring device. Since the establishment of this dry interim storage concept in Germany in the early 1980s, a great deal of experience has been accumulated and now spent fuel elements from the THTR reactor at Hamm-Uentrop and from the AVR research reactor at Juelich are loaded into CASTOR-THTR/AVR casks under dry conditions and stored in the licensed external dry interim storage facilities in Ahaus and Juelich. These are now routine procedures that started in 1992 and has so far comprise more than 200 casks. A great deal of operational experience exists and has also been gained in process optimization without any serious problems. Much more difficult are the drying and evacuation procedures for casks loaded under wet conditions in the spent fuel storage pond of a nuclear power plant. In this case, special operational procedures involving humidity measurements are applied. Different loading operations in several German power plants have been carried out since 1982 and the first wet loaded cask proposed for storage in the licensed external dry interim storage facility at Gorleben came into operation in July 1994. (author). 4 refs, 5 figs, 1 tab

  16. Seismic considerations for spent nuclear fuel storage in dry casks

    Institute of Scientific and Technical Information of China (English)

    John L Bignell; Jeffrey A Smith; Christopher A Jones; Susan Y Pickering

    2013-01-01

    To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems.Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters.The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g.A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping.In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask.The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over).The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask.Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

  19. Thermal safety analysis of a dry storage cask for the Korean standard spent fuel - 16159

    International Nuclear Information System (INIS)

    A conceptual dry storage facility, which is based on a commercial dry storage facility, was designed for the Korea standard spent nuclear fuel (SNF) and preliminary thermal safety analysis was performed in this study. To perform the preliminary thermal analysis, a thermal analysis method was proposed. The thermal analysis method consists of 2 parts. By using the method, the surface temperature of the storage canister corresponding to the SNF clad temperature was calculated and the adequate air duct area was decided using the calculation result. The initial temperature of the facility was calculated and the fire condition and half air duct blockage were analyzed. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L.; Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R.; Barrett, P.R. [ANATECH Research Corp., La Jolla, CA (United States); Malinauskas, A.P. [Oak Ridge National Lab., TN (United States); Einziger, R.E. [Pacific Northwest Lab., Richland, WA (United States); Jordan, H. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant; Duffey, T.A.; Sutherland, S.H. [APTEK, Inc., Colorado Springs, CO (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  4. SNF shipping cask shielding analysis

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Pace, J.V. III

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  7. A numerical study of transportation casks subjected to puncture loads

    International Nuclear Information System (INIS)

    A nonlinear dynamic finite element analysis has been performed to study the structural response of casks subjected to puncture load. Particular attention is placed on the Multipurpose Canister (MPC) and General Atomic (GA) casks that are currently under development. The structural response of the casks subjected to both regulatory hypothetical accidents and accidents beyond regulatory requirements were evaluated. A performance map was presented for casks subjected to regulatory formula puncture tests, and the structural contribution of the various layers backing the steel cask shell has been studied

  8. 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

    International Nuclear Information System (INIS)

    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

  9. 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

    Energy Technology Data Exchange (ETDEWEB)

    Mok, G.C.; Thomas, G.R.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L. [Lawrence Livermore National Lab., CA (United States)

    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.

  10. Evaluation of Helium Purge and Vent Process to Reduce Oxygen Concentrations in the Large Diameter Container and Cask Void Volumes at K Basin

    International Nuclear Information System (INIS)

    The purpose of this document is to provide calculations to model the following activities and associated procedures: (1) Model a Helium Purge System (HPS) to reduce the oxygen concentration (i.e., O2 mole fraction) to less than 1% in the single Large Diameter Container (LDC) void volume, by a direct purge and vent process, after sludge load out is complete. (2) Model a HPS to reduce oxygen concentration (i.e., O2 mole fraction) to less than 1% in the Cask and filter-connected LDC void volumes prior to transport to T-Plant. This document will evaluate and determine the following items, in order, to address the issues noted above: (1) Demonstrate the purge system process and methodology will ensure the Cask and LDC void volumes can be purged below 1% oxygen for both models defined above. (2) Based on previous item (1), determine the number of purge/vent cycles for the Cask/LDC, and single LDC, to enable the LDC void volume to obtain an oxygen concentration below 1%. (3) Based on previous items (1) and (2), determine the length of purge/vent time for each cycle and model using a reduced final purge cycle pressure in single LDC (i.e., 35 psig) and Cask/LDC (37 psig) and using an increased final vent cycle pressure in single LDC (i.e., 4 psig) and Cask/LDC (7 psig). Revision 2 of this document provides a greater purge pressure and reduced vent pressure per cycle which increases process cycle durations but decreases oxygen concentrations per cycle. (4) Determine a recommended dynamic pressure setting on the helium purge feed regulator setting or volumetric helium feed flow to meet the proposed cycle times in previous items (1) through (3). (5) Determine a final Cask purge pressure based on the single LDC process run data. The final pressure shall ensure the process avoids damaging the filter media between the two void volumes (6) Provide any special design change recommendations or specific design requirements that the purge system must meet to adequately optimize the

  11. Solar Thermal Concept Evaluation

    Science.gov (United States)

    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

  12. Occupant thermal comfort evaluation

    Science.gov (United States)

    Ghiardi, Gena L.

    1999-03-01

    Throughout the automotive industry there has been an increasing concern and focus on the thermal comfort of occupants. Manufacturers are continuously striving to improve heating and air conditioning performance to comply with expanding customer needs. To optimize these systems, the technology to acquire data must also be enhanced. In this evaluation, the standard use of isolated thermocouple location technology is compared to utilizing infrared thermal vision in an air conditioning performance assessment. Infrared data on an actual occupant is correlated to breath and air conditioning output temperatures measured by positioned thermocouples. The use of infrared thermal vision highlights various areas of comfort and discomfort experienced by the occupant. The evaluation involves utilizing an infrared thermal vision camera to film an occupant in the vehicle as the following test procedure is run. The vehicle is soaked in full sun load until the interior temperature reaches a minimum of 150 degrees F (65.6 degrees Celsius). The occupant enters the vehicle and takes an initial temperature reading. The air conditioning is turned on to full cold, full fan speed, and recirculation mode. While being filmed, the driver drives for sixty minutes at 30 miles per hour (48.3 kph). The thermocouples acquire data in one minute intervals while the infrared camera films the cooling process of the occupant.

  13. Testing of Metal Cask and Concrete Cask

    International Nuclear Information System (INIS)

    In Japan, the first interim spent fuel storage facility (ISF) outside of nuclear power plant site in use of dual-purpose metal cask is being planned to start its commercial operation in 2012 in Mutsu city, Aomori prefecture. The CRIEPI (Central Research Institute of Electric Power Industry) has executed several study programs on demonstrative testing for interim storage of spent fuel, mainly related to metal cask and concrete cask storage technology to reflect in Japanese safety requirements for dry casks issued by NISA/METI (Nuclear and Industrial Safety Agency, Ministry of Economy and Trade Industry). On top of that, the Japan Nuclear Energy Safety Organization (JNES) has executed study programs on spent fuel integrity, etc. This paper introduces the summary of these research programs. (author)

  14. Thermal evaluation of buildings

    OpenAIRE

    Barajas, Luís M; Roset Calzada, Jaime; La Ferla, Giuseppe

    2015-01-01

    To COST ACTION TU 1104 "Smart Energy Regions" Prof. Aleksandra Djukic and Prof. Aleksandra Krstic-Furundzic of the Faculty of Architecture of the University of Belgrade, Serbia, that gave us the opportunity of be part of the training school imparted from Monday 20th to Thursday 23rd April 2015, where we can teach the topic. The convenience of the use of environmental building evaluation tools, to know design conditions and thermal behavior, by using bioclimatic strategies fo...

  15. CFD analysis of a cask for spent fuel dry storage: Model fundamentals and sensitivity studies

    International Nuclear Information System (INIS)

    Highlights: • A dry storage cask has been evaluated by a CFD code, FLUENT 14. • An alternative methodology for thermal-fluid dynamic modeling has been performed. • Fuel maximum temperature obtained is around 50 K below the regulation limit (673 K). • Even in the most unfavorable heat load distribution temperature increase is smaller than 4%. - Abstract: Dry storage technology must ensure spent fuel cooling under any conditions. This turns thermo-fluid dynamics within dry storage casks a key aspect to investigate, as it would heavily affect fuel rod temperatures. This paper introduces a Computational Fluid Dynamic (CFD) model and analyses of a HI-STORM 100S cask with FLUENT 14.0. Fuel assemblies have been modeled as a porous medium characterized by a thermal conductivity and pressure drop that have been derived from specific approximations, algorithms and methods. This approach has been verified by comparing its results to those published by Holtec International for the HI-STORM cask. The application of the 3D model to HI-STORM 100S cask type under normal conditions, confirms that fuel maximum temperatures more than about 50 K below the regulation limit (673 K) should be expected. In addition, the effect on these results of aspects such as cask design (inlet/outlet orientation), heat load (regionalization) and local climate (external temperature), have been explored. The results indicate that the most relevant factor is heat load distribution and that, even in the most unfavorable regionalization feasible, temperature increase is smaller than 4%. Nonetheless, it should be highlighted that thermal margin to regulatory setting might be reduced down to around 40%

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

    Energy Technology Data Exchange (ETDEWEB)

    Verst, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    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.

  17. SNS Inner Plug Shipping Cask Analysis

    International Nuclear Information System (INIS)

    Calculations were performed to evaluate the dose rates outside the shipping cask containing the Spallation Neutron Source (SNS) inner plug assembly. The analysis consisted of simulating the proton beam interaction with the SNS target, activation calculations with the determined neutron flux levels and assumed SNS operation schedule, and calculation of the decay gamma-rays propagation through the inner plug and shipping cask. Several materials were considered for the inner plug. The results provide guidance for the finalization of the plug design

  18. Design report for cask transportation equipment

    International Nuclear Information System (INIS)

    In Korea, the spent fuels stored in the spent fuel storage pools in the domestic nuclear power plants significantly affects the continuation of the power plant operation. To solve this problem, KAERI has developed KSC-4 spent fuel shipping cask, which can transport 4 PWR spent fuel assemblies. Besides the development of the cask, KAERI developed transportation equipment which needed to use of KSC-4 cask. These equipment consist of cask handling tools such as lifting yoke, lid handling tool and spent fuel handling tool, etc. and transportation equipment such as trailer. In this report the usages, structures and functions of these tools and equipment were described, and the safety evaluation was carried out for each equipment

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

    International Nuclear Information System (INIS)

    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 ANSYSR 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)

  20. Cask Processing Enclosure Specification/Operation - 12231

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. A cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Experimental investigation of heat removal performance of a concrete storage cask

    International Nuclear Information System (INIS)

    Highlights: • Thermal tests were performed to evaluate the heat removal performance of the concrete storage cask. • Passive heat removal system was well designed and worked adequately. • Half-blockage of the inlet has a relatively small effect. • Thermal integrity of the concrete is maintained under accident conditions. - Abstract: Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A concrete storage cask to safely store spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. Moreover, the concrete storage cask must ensure that the temperatures of the spent nuclear fuel assemblies are maintained within the allowable values for normal, off-normal, and accident conditions. Therefore, the concrete storage cask must be designed to have heat removal capabilities with appropriate reliability. However, the thermal conductivity of concrete is not good and the allowable temperature of concrete is lower than that of steel. In this study, a thermal test was performed to evaluate the heat removal performance of the concrete storage cask under development by KORAD (Korea Radioactive Waste Agency), under normal and off-normal conditions. In addition, a thermal test was performed to evaluate the thermal integrity of the concrete under accident conditions. The heat transfer rate to the ambient atmosphere by convective airflow through the passive heat removal system of the concrete storage cask was found to reach 93.5% under normal conditions. Thus, it was confirmed that the passive heat removal system was well designed and worked adequately. In addition, the heat transfer rate to the ambient atmosphere by convective airflow through the passive heat removal system under off-normal conditions was estimated to reach 87.4%. Therefore, it was deduced that the half-blockage of the inlet openings has a relatively small effect on the maximum temperatures and temperature distributions

  4. Safety analysis report for packaging: the ORNL loop transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Nelms, H.A.; Crowley, W.K.; Just, R.A.

    1977-11-01

    An evaluation of the ORNL loop transport cask demonstrating its compliance with the regulations governing the transportation of radioactive and fissile materials is presented. A previous review of the cask is updated to demonstrate compliance with current regulations, to present current procedures, and to reflect the more recent technology.

  5. Thermo-mechanical finite element analyses of bolted cask lid structures

    Energy Technology Data Exchange (ETDEWEB)

    Wieser, G.; Qiao Linan; Eberle, A.; Voelzke, H. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    The analysis of complex bolted cask lid structures under mechanical or thermal accident conditions is important for the evaluation of cask integrity and leak-tightness in package design assessment according to the Transport Regulations or in aircraft crash scenarios. In this context BAM is developing methods based on Finite Elements to calculate the effects of mechanical impacts onto the bolted lid structures as well as effects caused by severe fire scenarios. I n case of fire it might be not enough to perform only a thermal heat transfer analysis. The complex cask design in connection with a severe hypothetical time-temperature-curve representing an accident fire scenario will create a strong transient heating up of the cask body and its lid system. This causes relative displacements between the seals and its counterparts that can be analyzed by a so-called thermo-mechanical calculation. Although it is currently not possible to correlate leakage rates with results from deformation analyses directly an appropriate Finite Element model of the considered type of metallic lid seal has been developed. For the present it is possible to estimate the behaviour of the seal based on the calculated relative displacements at its seating and the behaviour of the lid bolts under the impact load or the temperature field respectively. Except of the lid bolts the geometry of the cask and the mechanical loading is axial-symmetric which simplifies the analysis considerably and a two-dimensional Finite Element model with substitute lid bolts may be used. The substitute bolts are modelled as one-dimensional truss or beam elements. An advanced two-dimensional bolt submodel represents the bolts with plane stress continuum elements. This paper discusses the influence of different bolt modelling on the relative displacements at the seating of the seals. Besides this, the influence of bolt modelling, thermal properties and detail in geometry of the two-dimensional Finite Element models on

  6. Seismic tipping analysis of a spent nuclear fuel shipping cask sitting on a crush pad

    International Nuclear Information System (INIS)

    A crush pad has been designed and analyzed to absorb the kinetic energy of an accidentally dropped spent nuclear fuel shipping cask into a 44 ft. deep cask unloading pool. Conventional analysis techniques available for evaluating a cask for tipping due to lateral seismic forces assume that the cask rests on a rigid surface. In this analysis, the cask (110 tons) sits on a stainless steel encased (0.25 in. top plate), polyurethane foam (4 ft. thick) crush pad. As the cask tends to rock due to horizontal seismic forces, the contact area between the cask and the crush pad is reduced, increasing the bearing stress, and causing the pivoting corner of the cask to depress into the crush pad. As the crush pad depresses under the cask corner, the pivot point shifts from the corner toward the cask center, which facilitates rocking and potential tipping of the cask. Subsequent rocking of the cask may deepen the depression, further contributing to the likelihood of cask tip over. However, as the depression is created, the crush pad is absorbing energy from the rocking cask. Potential tip over of the cask was evaluated by performing a non-linear, dynamic, finite element analysis with acceleration time history input. This time history analysis captured the effect of a deforming crush pad, and also eliminated conservatisms of the conventional approaches. For comparison purposes, this analysis was also performed with the cask sitting on a solid stainless steel crush pad. Results indicate that the conventional methods are quite conservative relative to the more exacting time history analysis. They also indicate that the rocking motion is less on the foam crush pad than on the solid stainless steel pad

  7. Evaluation of mechanical properties and low velocity impact characteristics of balsa wood and urethane foam applied to impact limiter of nuclear spent fuel shipping cask

    International Nuclear Information System (INIS)

    The paper aims to evaluate the low velocity impact responses and mechanical properties of balsa wood and urethane foam core materials and their sandwich panels, which are applied as the impact limiter of a nuclear spent fuel shipping cask. For the urethane foam core, which is isotropic, tensile, compressive, and shear mechanical tests were conducted. For the balsa wood core, which is orthotropic and shows different material properties in different orthogonal directions, nine mechanical properties were determined. The impact test specimens for the core material and their sandwich panel were subjected to low velocity impact loads using an instrumented testing machine at impact energy levels of 1, 3, and 5J. The experimental results showed that both the urethane foam and the balsa wood core except in the growth direction (z-direction) had a similar impact response for the energy absorbing capacity, contact force, and indentation. Furthermore, it was found that the urethane foam core was suitable as an impact limiter material owing to its resistance to fire and low cost, and the balsa wood core could also be strongly considered as an impact limiter material for a lightweight nuclear spent fuel shipping cask

  8. Computational fluid dynamics analysis for K24B cask design with burnup credit

    International Nuclear Information System (INIS)

    Korea Nuclear Engineering Service Corp. (KONES) has designed K24B cask for the storage and the transportation of 24 (CE-type 16x16) PWR assemblies. K24B cask is designed with considering burnup credit of spent fuel. In order to remove heat from the fuel assemblies effectively, the flow channels in the upper and the lower part of fuel assemblies are set up to promote the natural convection. Computational fluid dynamics analysis is carried out to estimate and assure the thermal integrity of K24B cask. Conduction and radiation heat transfer through the cask components and the natural convective heat transfer in the cask are simulated. As a result of the analysis, the maximum temperatures of the cask components are maintained below the operating temperature for the safety. Therefore, the design of K24B cask can satisfy the safety limit. (author)

  9. LMFBR thermal-striping evaluation

    International Nuclear Information System (INIS)

    Thermal striping is defined as the fluctuating temperature field that is imposed on a structure when fluid streams at different temperatures mix in the vicinity of the structure surface. Because of the uncertainty in structural damage in LMFBR structures subject to thermal striping, EPRI has funded an effort for the Rockwell International Energy Systems Group to evaluate this problem. This interim report presents the following information: (1) a Thermal Striping Program Plan which identifies areas of analytic and experimental needs and presents a program of specific tasks to define damage experienced by ordinary materials of construction and to evaluate conservatism in the existing approach; (2) a description of the Thermal Striping Test Facility and its operation; and (3) results from the preliminary phase of testing to characterize the fluid environment to be applied in subsequent thermal striping damage experiments

  10. Al analysis and design of dry storage cask of spent nuclear fuel

    International Nuclear Information System (INIS)

    According to thermal analysis of the existing concrete cask, the maximum temperature occurred at the upper side of cask. If the cask lid is made of concrete, the temperature of concrete in lid exceeds the allowable value. Based on that result, research is progressed focusing on two strategies - one is to increase thermal margin, another is to decrease the lid concrete temperature. Here, thermally - enhanced design is suggested and analyzed. This design features the air flow duct in the lid and the thermal shielding disk installed between canister and lid. Air flow duct on the center of lid concrete connected to existing air outlet can decrease temperature by promoting the convection heat transfer, and thermal shielding disk bearing smaller hole located on the center can maintain the increased convection heat transfer and minimize radiation heat transfer from canister to lid concrete for the lid concrete temperature not to be over the allowable value. Thermal analysis result for this design shows that it can be very successful to achieve these objectives. The overall component of cask temperature decrease by 2-10 .deg. C, and the lid concrete temperature dropped from above 100 to 87.5 .deg. C which is below the allowable value 93 .deg. C. In addition, heat removal of cask depending on distance between casks was investigated. Cask heat is removed by convection and radiation heat transfer at an external surface to environment. Naturally, these heat transfers are mainly affected by ambient temperature. The ambient temperature can be affected if the thermal boundary layer is overlapped. So, thermal boundary layer thickness of cask was calculated to estimate to see if the ambient temperature is affected by other cask. Boundary layer thickness is calculated is too small just about 5cm. It is concluded that distance between casks can do little impact on heat removal of cask in a practical view

  11. Simplified computer codes for cask impact analysis

    International Nuclear Information System (INIS)

    In regard to the evaluation of the acceleration and deformation of casks, the simplified computer codes make analyses economical and decrease input and calculation time. The results obtained by the simplified computer codes have enough adequacy for their practical use. (J.P.N.)

  12. Surface storage cask test summarization report

    International Nuclear Information System (INIS)

    From December 1978 to September 1982, as part of DOE's Spent Fuel Handling and Packaging Program and Commercial Waste and Spent Fuel Packaging Program, a pressurized water reactor (PWR) spent nuclear fuel assembly with an initial decay heat level of approximately 1.0 kilowatt (kW) was emplaced in a concrete cask at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility in Area 25 of the Nevada Test Site. Temperatures were monitored during the emplacement period to determine the thermal response of the cask, the canister, and the fuel assembly. During and following the test, the atmosphere of the canister containing the fuel assembly was sampled to determine if fission product gases had been released by the fuel assembly. This 45-month Surface Storage Cask (SSC) test was the first demonstration of interim storage of a PWR spent fuel assembly in a dry storage cask. The receipt, handling, packaging, emplacement and retrieval operations have been demonstrated as directly applicable to similar operations in federal interim storage and repository related activities. 7 references, 35 figures, 7 tables

  13. Overview of research and development of metal cask for transport and storage of spent nuclear fuel in Japan

    International Nuclear Information System (INIS)

    The paper overviews experimental studies of dual-purpose metal casks carried out in Japan. Full-scale casks were dropped onto a reinforced concrete target simulating hypothetical accidental drop during handling procedure in a storage facility. In some cases, leakage from the primary lid was detected, but no leakage from the secondary lid. A heavy weight drop test was carried out onto a full-scale cask simulating hypothetical collapse of a storage building due to earthquake, etc. The cask maintained its integrity. A full-scale cask was covered with a thermal insulator simulating a hypothetical burial by debris due to a building collapse in earthquake, etc. Some components might need to be recovered from the debris before reaching their critical temperature. A scale-model of a cask was subjected to seismic motion on a shaking table simulating an earthquake. The cask was rocking more for an earthquake with longer wavelength. Long-term containment of metal gaskets in double lid structure of casks has been tested with full-scale lid model. Transportability of cask after long-term storage was tested simulating degradation of cask components. Effects of aging of cask body metal, basket metal, seal and neutron shielding materials were investigated. With those degradations, cask performance in terms of shielding, sub-criticality, heat removal and containment were investigated. (author)

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

    International Nuclear Information System (INIS)

    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

  15. IMPACT ANALYSIS OF SPENT FUEL DRY CASKS UNDER ACCIDENTAL DROP SCENARIOS

    International Nuclear Information System (INIS)

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister OLIIpC that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Cask development, testing, and licensing

    International Nuclear Information System (INIS)

    The NuPac 125-B Rail Cask was developed to provide a safe means of transporting the damaged core of Three Mile Island Unit 2 from the TMI site at Middletown, PA, to the Idaho National Engineering laboratory (INEL) at Idaho Falls, ID. The development of the NuPac 125-B Rail Cask posed two engineering and technical management challenges; Licensing Strategy - The NuPac 125-B Rail Cask represented the first irradiated fuel rail cask developed within the United States in the past decade, a decade characterized by changing nuclear regulations, and Accelerated Schedule - The TMI-2 defueling schedule demanded a cask development schedule one-third as long as normally required. These challenges governed the overall development and licensing process for the cask. First, a high degree of conservation was incorporated into the design to allow quick, simplified demonstrations of adequacy to regulatory staff. Second, redundant design techniques were employed in all areas of uncertainty. The testing program eliminated performance uncertainties and validated predictions and predictive models. Drop tests of a quarter-scale model of the cask were conducted, and results were correlated with analytic predictions to verify structural and mechanical performance of the cask. Full-scale tests of the canisters were conducted to verify structural behavior of canister internals which provide criticality control. This paper describes the testing program for the NuPac 125-B Rail Cask, presents results therefrom, and correlates findings with Regulation 10 CFR 71 of the U.S. Nuclear Regulatory Commission

  19. Safety analysis report for packaging: the ORNL HFIR spent-fuel-element shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Eversole, R.E.; Just, R.A.; Llewellyn, G.H.

    1977-11-01

    The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) spent-fuel-element shipping cask is used to transport HFIR, Oak Ridge Research Reactor (ORR), and other reactor fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures and tests were used to determine behavior of the cask relative to the general standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Shielding analysis of dual purpose casks for spent nuclear fuel under normal storage conditions

    International Nuclear Information System (INIS)

    Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, UO2 pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a 2 x 10 cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the 2 x 10 cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the 2 x

  2. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P.; Ribeiro, M.I.; Aparicio, P. [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  3. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  4. CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. Inspection of Used Fuel Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yoshimura, H.R.; Bronowski, D.R.; Uncapher, W.L.; Attaway, S.W.; Bateman, V.I.; Carne, T.G.; Gregory, D.L. (Sandia National Labs., Albuquerque, NM (USA)); Huerta, M. (Southwest Engineering Associates, El Paso, TX (USA))

    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.

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

    International Nuclear Information System (INIS)

    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. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  9. Contract Report for Usage Inspection of KN-12 Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K

    2007-03-15

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Cask fleet operations study

    International Nuclear Information System (INIS)

    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. Cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. SLI Thermal Imaging Requirements Evaluation

    Science.gov (United States)

    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.

  14. CASKETSS-DYNA2D: a nonlinear impact analysis computer program for nuclear fuel transport casks in two dimensional geometries

    International Nuclear Information System (INIS)

    A nonlinear impact analysis computer program DYNA2D, which was developed by Hallquist, has been introduced from Lawrence Livermore National Laboratory for the purpose of using impact analysis of nuclear fuel transport casks. DYNA2D has been built in CASKETSS code system (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of DYNA2D are as follows; (1) This program has been programmed to provide near optimal speed on vector processing computers. (2) An explicit time integration method is used for fast calculation. (3) Many material models are available in the program. (4) A contact-impact algorithm permits gap and sliding along structural interfaces. (5) A rezoner has been embedded in the program. (6) The graphic program for representations of calculation is provided. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    International Nuclear Information System (INIS)

    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

  17. 78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-12-27

    ... RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS Cask... final rule amended the NRC's spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to...

  18. Safety assessment of a metal cask under aircraft engine crash

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of); Choi, Woo Seok; Seo, Ki Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD) was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is free standing on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact load-time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

  19. Seismic Response Analysis of Spent Nuclear Fuel Metal Storage Cask considering Soil- Structure Interaction Effects

    Energy Technology Data Exchange (ETDEWEB)

    Baeg, Chang-Yeal; Lee, Kyung-Ho; Lee, Dae-Ki [Nuclear Engineering and Technology Institute, Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Jung, In-Su; Song, Won-Tae; Jin, Han-Uk; Kim, Jong-Soo [KONES, Seoul (Korea, Republic of)

    2008-05-15

    Maintaining of the structure safety for the metal storage cask is important to store spent nuclear fuel under a seismic events. Sliding and overturning behavior must be estimated because the metal cask systems are to be installed as free standing structures on reinforced concrete pads. This behavior can cause a serious problem in the integrity of spent nuclear fuel by the impact between neighboring casks. Also, soil condition should be considered since the cask's behavior is strongly affected by the characteristics of the base soil condition. In this study, the seismic response analysis was carried out in order to evaluate the behavior of the metal storage cask under earthquake envelopment considering Soil-Structure Interaction (SSI) effects.

  20. GNS spent fuel cask experience

    International Nuclear Information System (INIS)

    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. GNS spent fuel cask experience

    Energy Technology Data Exchange (ETDEWEB)

    Weh, R. (Gesellschaft fuer Nuklear-Service mbH, Hannover (Germany))

    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.

  2. Initiatives in transport cask license

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, John [NAC International, Aiken, SC (United States). Foreign Research Reactor Liaison]. E-mail: nacaiken@aol.com

    1998-07-01

    The variations in research reactor fuel form, configuration, irradiation characteristics, and transport cask have required a substantial number of transport cask licensing actions associated with foreign research reactor spent fuel transportation. When compounded by limited time for shipment preparations, due to contract timing or delayed receipt of technical data, the number and timing of certifications has adversely impacted the ability of regulatory agencies to support intended shipping schedules. This issue was brought into focus at a april, 1998 meeting among DOE, the US Nuclear Regulatory Commission, and DOE's spent fuel transportation contractors. (author)

  3. Source storage and transfer cask: Users Guide

    International Nuclear Information System (INIS)

    The storage and shield cask for the dual californium source is designed to shield and transport up to 3.7 mg (2 Ci) of 252Cf. 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

  4. Structural analysis of closure bolts for shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Mok, G.C.; Fischer, L.E.

    1993-04-01

    This paper identifies the active forces and moments in a closure bolt of a shipping cask. It examines the interactions of these forces/moments and suggest simplified methods for their analysis. The paper also evaluates the role that the forces and moments play in the structure integrity of the closure bolt and recommends stress limits and desirable practices to ensure its integrity.

  5. Monte Carlo shipping cask calculations using an automated biasing procedure

    International Nuclear Information System (INIS)

    This paper describes an automated biasing procedure for Monte Carlo shipping cask calculations within the SCALE system - a modular code system for Standardized Computer Analysis for Licensing Evaluation. The SCALE system was conceived and funded by the US Nuclear Regulatory Commission to satisfy a strong need for performing standardized criticality, shielding, and heat transfer analyses of nuclear systems

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

    International Nuclear Information System (INIS)

    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%

  7. The evaluation of thermal hotels' online reviews

    OpenAIRE

    BERTAN, Serkan; Bayram, Murat; Benzergil, Nisan

    2015-01-01

    The main objective of this study was to evaluate the perceptions related to the online user reviews of thermal hotels. Specifically, it was investigated whether perceptions towards value (V), location (L), sleep quality (SQ), rooms (R), cleanliness (C), service (S) and factors influencing general evaluation depend on the star numbers of hotels, the location of the hotels and the nationalities of participants. In order to obtain data on perceptions of consumers towards thermal hotels in Turkey...

  8. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

  9. Heat transfer investigations within dry spent fuel casks

    International Nuclear Information System (INIS)

    For studying the heat transfer processes and predicting the maximum spent fuel element surface temperature in a spent fuel assembly (SFA) transported in a dry cask, model experiments have been performed with a gas-filled model cask containing a simplified electrically heated model of a WWER-type SFA with 90 fuel elements. The temperature distribution of the SFA model is measured for different heat rates under vacuum in the model cask, and under normal pressure and overpressure (0.1 ... 0.7 MPa) for several cooling gases (air, argon, helium) in order to separately investigate heat transfer processes by radiation and convection/conduction. The measuring results were compared with the calculations. Computer programmes as well as simplified calculation methods for temperature prediction were developed and checked. The results obtained are also useful for thermal analyses in the field of the dry storage of SFAs in a cask or can. Specifically it was found that: The heat removal from the SFA can be considerably improved by increasing the internal cask pressure or by using helium as coolant. The radiant heat exchange in the SFA model can be calculated with sufficient accuracy by means of a computer programme developed in 1978 or by means of a simplified analytical representation shown in the final report. Both methods are directly applicable to the original SFA and useful in order to approximately calculate the maximum SFE surface temperature under normal pressure, if the fraction of heat transferred by radiation is allowed for. For the calculation of the total heat transfer a computer programme was developed and verified, which completely permits the temperature prediction of the SFA model in dependence on heat rate, type of gaseous coolant and coolant pressure. This computer programme can be directly applied to the original SFA for the calculation of the maximum SFE surface temperature

  10. Genetics Home Reference: CASK-related intellectual disability

    Science.gov (United States)

    ... Conditions CASK-related intellectual disability CASK-related intellectual disability Enable Javascript to view the expand/collapse boxes. ... Open All Close All Description CASK -related intellectual disability is a disorder of brain development that has ...

  11. Histopathological evaluation of tissue undergoing thermal insult

    Science.gov (United States)

    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

  12. Interim Dry Storage of Spent Fuel in Casks

    International Nuclear Information System (INIS)

    French option for the back end of the fuel cycle is reprocessing of used fuel and recycling the fissile material, except some very specific fuel stored in vaults (dry conditions). Used fuel management solutions studied by AREVA for various countries allow for either direct transport to the reprocessing plant, or interim storage and transport after storage of used fuel. Interim storage solutions are wet storage or dry storage (DSC, metal casks or vault systems). When the decision on used fuel management has been postponed, some extension of interim storage duration is considered, therefore it becomes necessary to study used fuel and cask material behaviour and deterioration mechanisms. One objective of this R&D was to review research efforts on spent fuel behaviour and Dry storage experience in casks. Particularly we were interested in the assessment of retrievability of fuel after storage for further use. A review therefore, was made of the effect of storage time/ temperatures and of loading/ drying operation on used fuel integrity. R&D programmes were also carried out on the evaluation of cask materials in long term, especially materials susceptible to degradation

  13. Test Plan for Cask Identification Detector

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-29

    This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

  14. Radioactive fuel cask railcar humping study

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, L.T. (comp.)

    1978-01-01

    The response of two radioactive shipping casks due to railroad humping shocks was calculated using a spring-mass model. The two railcars for these casks had different coupling mechanisms and different tiedown arrangements. Humping tests had been performed on one of the railcars (ATMX-600) and the resulting shock spectra was used to adjust the spring-mass model to get matching results. One car (designed for cask shipment) was equipped with Freightmaster E-15 end of car coupler and had about /sup 1///sub 8/ in. free travel of the cask skid relative to the car. The other car (ATMX-600), equipped with Miner RF-333 draft gear, was designed for nuclear weapon shipment and adapted to nuclear waste shipment by fastening the casks to the floor. Both car frames were built by the same manufacturer and are very similar. The response of the casks was put in shock spectra format and a parametric study was performed with various cask weights. Additional studies were done on the effects of fastening the loose cask, and using the Freightmaster end of car coupler on the ATMX car. Half-sine response spectra were overlaid to include the natural frequency of the cask tiedown. The resulting shock amplitude was plotted against the cask weight for each car. The results show a constant acceleration level for all the weights on the car with hydraulic end-of-car coupler which results from constant force at that impact velocity. The cask acceleration can be reduced by fastening it to the car, rather than allowing it to move freely through some small space. This study also shows that the cask response can be optimized on railcars without hydraulic draft gear by adjusting the tiedown stiffness to keep the tiedown frequency different than car frequencies.

  15. Development of cask and transportation system

    International Nuclear Information System (INIS)

    Transportation of spent fuels to the AFR interim storage facility and disposal repository are necessary in Korea. Therefore, an emphasis has been concentrated to develop the design and fabrication technology of commercial casks. A conceptual design of the temperature and deformation measuring systems in the cask, which will be used for mock-up tests has been performed. Preliminary design data of the cask for 7 spent PWR fuels have been obtained in the course of study. (author)

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

    Science.gov (United States)

    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.

  17. Used Fuel Cask Identification through Neutron Profile

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Laboratory

    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.

  18. Transportation capabilities of the existing cask fleet

    International Nuclear Information System (INIS)

    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 465 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the 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

  19. Feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer based on single-element transfer cask experience

    International Nuclear Information System (INIS)

    Spent fuel transportation casks and canister-based storage systems are generally loaded underwater in a nuclear plant's spent fuel pool/cask loading pit. Several reasons exist for exploring the feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer. These include: the accommodation of plants which do not have sufficient crane capacity to handle large 90 tonne (100 ton) storage canisters or shipping casks, and construction of an MRS without the need for extensive hot cell facilities. In the case of DOE's ''Multi-Purpose Canister'' (MPC) scenario, use of such a transfer system would allow all plants with adequate transport routes to use large canisters at-reactor, and those without adequate transport routes to use the MRS for loading of large canisters without the need for hot cell facilities. The dry transfer option would not only allow the use of large canisters for all fuel, but would assist DOE in meeting MRS deadlines since licensing and construction of hot-cell facilities significantly affect schedule. This paper reviews the regulatory issues and technical design considerations for a single-element dry transfer system. Also summarized are lessons learned from the TMI-2 fuel transfer system which are directly applicable to the design, testing, startup, and use of a future dry cask-to-cask or pool-to-cask transfer system

  20. SAS1 and SAS4, two new shielding analysis sequences for spent fuel casks

    International Nuclear Information System (INIS)

    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. 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

  1. Evaluated neutron data for thermal reactor calculations

    International Nuclear Information System (INIS)

    The paper describes a library of evaluated neutron data designed for thermal reactor calculations and other low energy neutron physics applications. The name of the library is KORT (Evaluated Thermal Reactor Constants). The following information is given in KORT: a general characterization of the nucleus (mass, energy of capture and fission reactions, parameters of radioactive decay); partial cross-sections for neutrons of thermal energy, and the number of secondary fission neutrons (estimated errors in the measurements of these quantities are indicated); coefficients defining the deviation of capture and fission cross-sections from the 1/v law in a Maxwellian spectrum; resonance capture and fission integrals and the estimated errors in these quantities (for nuclei with Z>=90); detailed energy dependence of the cross-sections in the 10-4-5 eV region at T=300 K

  2. Shielding Analysis of the 5320 Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, A. [Westinghouse Savannah River Company, AIKEN, SC (United States); Nathan, S. [Westinghouse Safety Management Solutions, Aiken, SC (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Kaushik [ORNL; Scaglione, John M [ORNL

    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.

  5. Design review report FFTF interim storage cask

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. 78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

    ... RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9... amends the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 9...

  7. Thermal processes evaluation for RWMC wastes

    International Nuclear Information System (INIS)

    The objective of this activity was to provide a white paper that identifies, collects information, and presents a preliminary evaluation of ''core'' thermal technologies that could be applied to RWMC stored and buried mixed waste. This paper presents the results of the following activities: General thermal technology identification, collection of technical and cost information on each technology, identification of thermal technologies applicable to RWMC waste, evaluation of each technology as applied to RWMC waste in seven process attributes, scoring each technology on a one to five scale (five highest) in each process attribute. Reaching conclusions about the superiority of one technology over others is not advised based on this preliminary study alone. However, the highly rated technologies (i.e., overall score of 2.9 or better) are worthy of a more detailed evaluation. The next step should be a more detailed evaluation of the technologies that includes onsite visits with operational facilities, preconceptual treatment facility design analysis, and visits with developers for emerging technologies. 2 figs., 6 tabs

  8. Bonner sphere neutron spectrometry at spent fuel casks

    CERN Document Server

    Rimpler, A

    2002-01-01

    For transport and interim storage of spent fuel elements from power reactors and vitrified highly active waste (HAW) from reprocessing, various types of casks are used. The radiation exposure of the personnel during transportation and storage of these casks is caused by mixed photon-neutron fields and, frequently, the neutron dose is predominant. In operational radiation protection, survey meters and even personal dosemeters with imperfect energy dependence of the dose-equivalent response are used, i.e. the fluence response of the devices does not match the fluence-to-dose equivalent conversion function. In order to achieve more accurate dosimetric information and to investigate the performance of dosemeters, spectrometric investigations of the neutron fields are necessary. Therefore, fluence spectra and dose rates were measured by means of a simple portable Bonner multisphere spectrometer (BSS). The paper describes briefly the experimental set-up and evaluation procedure. Measured spectra for different locat...

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  10. Interactions between cask components and content of packaging for the transport of radioactive material during drop tests

    International Nuclear Information System (INIS)

    This paper describes the analytical, numerical and experimental investigations on the phenomenon of interactions between cask components and content of packages for the transport of radioactive material during drop tests required according to the IAEA Regulations for the Safe Transport of Radioactive Material. Radial and axial gaps between cask components and content are usually necessary for thermal reasons but larger gaps can exist because of the geometrical dimensions of the specified content. Consequently interactions between content and cask components (lid system, cask body, etc.) are possible and can not be excluded during drop tests. Interactions in this context are relative movements between cask and content which are mainly due to elastic spring effects after releasing the cask for the free drop. These relative movements can cause interior collisions between content and cask during the main impact of the package onto the unyielding target. Drop tests with various types of Type A and Type B packages fully instrumented with strain gauges and accelerometers showed that these interactions respectively interior collisions can be considerable relating to high forces acting on cask lids, lid bolts and the content. Of course the real quantitative consequences of the interactions depend upon different conditions, among others the drop orientation, the design characteristics of the impact limiters, the dimensions of the gaps, the material characteristics of the contents, etc. . In order to investigate more precisely the phenomenon of interactions BAM carried out finite element calculations for the named casks using the ABAQUS/ Standard and ABAQUS/ Explicit computer code comparing them with results obtained from experiments. Additionally, tests with a simplified model instrumented with accelerometers were carried out accompanied by finite element calculations and analytical calculations using MATHEMATICA. The investigations on the mentioned phenomena of interaction

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

    Energy Technology Data Exchange (ETDEWEB)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

    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

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

    International Nuclear Information System (INIS)

    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

  13. Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

    International Nuclear Information System (INIS)

    To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

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

    International Nuclear Information System (INIS)

    The structural properties of spent nuclear fuel shipping containers vary as a function of the cask wall temperature. An analysis is performed to determine the effect of a realistic, though bounding, hot day environment on the thermal behavior of spent fuel shipping casks. These results are compared to those which develop under a steady-state application of the prescribed normal thermal conditions of 10CFR71. 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 using the steady-state application of the regulatory boundary conditions. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the regulatory condition. This is due to the conservative assumptions present in the ambient conditions used. The analysis demonstrates that diurnal temperature variations which penetrate the cask wall have maxima substantially less than the corresponding temperatures obtained when applying the steady-state regulatory boundary conditions. Therefore, it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the steady-state interpretation of the 10CFR71 normal conditions

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

    Energy Technology Data Exchange (ETDEWEB)

    Manson, S.J. [Texas Univ., Austin, TX (United States). Coll. of Engineering; Gianoulakis, S.E. [Sandia National Labs., Albuquerque, NM (United States)

    1994-02-01

    The structural properties of spent nuclear fuel shipping containers vary as a function of the cask wall temperature. An analysis is performed to determine the effect of a realistic, though bounding, hot day environment on the thermal behavior of spent fuel shipping casks. These results are compared to those which develop under a steady-state application of the prescribed normal thermal conditions of 10CFR71. 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 using the steady-state application of the regulatory boundary conditions. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the regulatory condition. This is due to the conservative assumptions present in the ambient conditions used. The analysis demonstrates that diurnal temperature variations which penetrate the cask wall have maxima substantially less than the corresponding temperatures obtained when applying the steady-state regulatory boundary conditions. Therefore, it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the steady-state interpretation of the 10CFR71 normal conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nizeyiman, Fidele; Alami, Aatif; Issard, Herve; Bellenger, Veronique [TN International, 1 rue des herons, Montigny le Bretonneux, 78054 Saint Quentin en Yvelines (France); Laboratoire PIMM, Arts and Metiers ParisTech, 151 Bd de l' Hopital, 75013 Paris (France)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    International Nuclear Information System (INIS)

    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

  19. Thermal Environment evaluation in Commercial kitchens

    DEFF Research Database (Denmark)

    Simone, Angela; Olesen, Bjarne W.

    2012-01-01

    The indoor climate in commercial kitchens is often unsatisfactory and the working conditions can have a significant effect on employees’ comfort and productivity. The type (fast food, casual, etc.) and climatic zone can influence the thermal conditions in the kitchens. Moreover, size...... and arrangement of the kitchen zones, appliances, etc., complicate further an evaluation of the indoor thermal environment in kitchens. In general, comfort criteria are expressed in international standards such as ASHRAE 55 or ISO EN7730. But are these standardised methods applicable for such environments...... as commercial kitchens? There is therefore a need to study the indoor environment in commercial kitchens and to establish standardized methods and procedures for setting criteria that have to be met for the design and operation of kitchens. The present paper introduces a data collection protocol based...

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

    International Nuclear Information System (INIS)

    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 fuer 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

  1. 78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc.... Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  4. Database of refractories for explosive and fire resistant steel cask for packaging and transportation of radioactive and hazardous materials

    International Nuclear Information System (INIS)

    This paper contains the results of mechanical and thermophysical properties investigations of the dense and porous refractory concretes (silicate (building), chamotte (metallurgical), alumina, zirconia (including ceramics)). Porosities of these materials were 20 - 50 %. Compression strength, thermal conductivity, thermal expansion, heat capacity and operation temperature for this refractories are discussed. The split-Hopkinson bar method was used for investigation of the strain rate about 1000 sec-1. For damage assessment of the severe events connected with overheating of the metal and oxides contents of cask and terrorist attack by means of the anti-tank weapons to cask we discussed resistance of a zirconia ceramics(concrete) to melted mixture Zr, UO2, Fe2O3 and Monroe jet. Our results testify that the porous zirconia ceramics can use in the impact limiter system of casks under mechanical, thermal and chemical attacks. (authors)

  5. A structural analysis on the KN-12 spent nuclear fuel transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-08-15

    In this study, safety of the spent nuclear fuel cask KN-12 which is developed in 2000 is evaluated for hypothetical accidents conditions such as free drop, puncture, fire accident and water immersion. Finite element code ABAQUS/Explicit is used to compare with safety analysis report of the GNB in which analysis is performed with LS-DYNA3D for hypothetical accident conditions. Through this study, the safety of KN-12 is evaluated by comprehensive structural analysis. The capability and technological advancement of Korean community on the analysis and structural assessment of the cask will be improved. Also people's anxiety about radioactive dangers will be eliminated.

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

    Science.gov (United States)

    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...

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  8. Ultrasonic inspection techniques for two weld closures proposed for RSSF waste storage casks

    International Nuclear Information System (INIS)

    One method being considered for interim storage of high-level radioactive waste materials is to place these materials in large sealed stainless steel canisters and subsequently store these canisters in a second sealed steel storage cask. Weld procedures are proposed as the closure or seal for these vessels. Inspection of these closures to assure initial and long-term integrity of the closure welds presents a challenge to nondestructive testing. The environment is thermally (400 to 10000F) and radioactively (105 R/hr) hot necessitating remote inspection procedures. As a result of research work, ultrasonic test techniques were developed for inspecting the final weld closure of the waste cask. Special transducers, coupling techniques and fixturing were developed and demonstrated in a mockup test facility by remotely examining a 2-in. full penetration weld closure. The examination was performed at room ambient and at a temperature of 2000F. Testing at the desired temperature of 4000F was not completed due to a loss in transducer performance at temperatures in excess of 2000F. Upon completion of the mockup test demonstration, the cask was subjected to a drop test. The ultrasonic results of the pre- and post-examination of two weld closures (the 2-in. full penetration weld and the threaded plug with seal weld) are presented. After the completion of the drop test, both weld closures were radiographed. The radiographs verified the ultrasonic examination and the presence of weld defects in the same areas. Sectioning of the cask closure welds with metallographic verification was not completed at the time of this writing. As a result of the experience gained from the Retrievable Surface Storage Facility (RSSF) storage cask program, recommendations pertaining to the nondestructive engineering development program for Spent Unreprocessed Fuel (SURF) storage casks are presented

  9. Final version dry cask storage study

    International Nuclear Information System (INIS)

    This report was prepared in response to Section 5064 of the Nuclear Waste Policy Amendments Act of 1987 (the Amendments Act--Public Law 100-203), which directs the Secretary of Energy to conduct a study of the use of dry-cask-storage technology for storing spent fuel at the sites of civilian nuclear reactors until a geologic repository is available. In conducting this study, whose results are being reported to the Congress, the Secretary was to consider such factors as costs, effects on human health and the environment, and the extent to which the Nuclear Waste Fund can and should be used to provide funds for at-reactor storage. In addition, the Secretary was to consult with the Nuclear Regulatory Commission (NRC), include NRC comments in the report, and solicit the views of State and local governments and the public. The study performed in response to these requirements was based largely on data published by the DOE or the NRC or included in documents issued by the DOE. Among the DOE documents are the 1987 MRS proposal to the Congress and a subsequent report, prepared to supply the Congress with additional information on the MRS facility. Because in evaluating dry storage at reactor sites it is necessary to take into account other options for meeting storage needs, this study covered all forms of at-reactor storage. 107 refs., 15 figs., 10 tabs

  10. Initial version, dry cask storage study

    International Nuclear Information System (INIS)

    This report was prepared to study the use of dry cask storage for storing spent fuel at the sites of civilian nuclear reactors until a geologic repository is available. In conducting this study, whose results are to be reported to the Congress, the Secretary is to consider such factors as costs, effects on human health and the environment, and the extent to which the Nuclear Waste Fund can and should be used to provide funds for at-reactor storage. In addition, the Secretary is to consult with the Nuclear Regulatory Commission (NRC), include NRC comments in the report, and solicit the views of State and local governments and the public. The study performed in response to these requirements was based largely on data published by the NRC or included in documents issued by the DOE. Among the DOE documents are the MRS proposal to the Congress and a subsequent report, prepared to supply the Congress with additional information on the MRS facility. Because in evaluating dry storage at reactor sites it is necessary to take into account other options for meeting storage needs, this study covered all forms of at-reactor storage. 85 refs., 5 figs, 12 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Nickell, R.E.; James, R.J. (ANATECH Research Corp., San Diego, CA (United States))

    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.

  12. ITER Upper Port Plug handling cask system assessment and design proposals

    International Nuclear Information System (INIS)

    The current design of the ITER cask for Upper Port Plugs has been evaluated. Careful reduction of the number of mechanical degrees of freedom is an opportunity to relax the tolerances in the design, resulting in cost reduction and reliability increase. A new kinematical design for the tractor module has a higher stiffness to weight ratio, reduces actuator forces by a factor four and minimizes cross-talk between lift and rotation motion. Non-cantilevered handling is recommended to reduce wheel loads on the tractor by a factor six and to simplify guidance. At the system level the tubular guide (TG) is proposed, a semi-permanent 3.5 m long tube which is an extension of the Upper Port. Cask docking is simplified and the risk of the cask tilting is prevented. Redesigning the system concept is recommended and the TG looks promising. Since a system level redesign impacts the external interfaces, overall feasibility has to be investigated.

  13. 78 FR 73379 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... of approved spent fuel storage casks in 10 CFR 72.214 (65 FR 25241; May 1, 2000). The environmental... 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

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

    International Nuclear Information System (INIS)

    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

  17. The state of the Primary Degradation Factors and Models of Concrete Cask in Spent Fuel Dry Storage System

    International Nuclear Information System (INIS)

    In South Korea, a total of twenty nuclear reactors are in operation; the cumulative amount of spent fuel is estimated to be 10,490 MTU in 2009. The full capacity of the waste storage is expected to be saturated in around 2016. However, a national strategy for spent fuel management has not yet been set down and high level waste (HLW) such as spent fuel will have to be stored at-reactor (AR) by re-racking. Recently an worldwide interest on the dry storage has increased especially around U.S. With a perspective of the material of the spent fuel dry storage cask, the system can be divided into two types of metal and concrete casks. The concrete type cask is a very attractive option because of the cost competitiveness of concrete material and its relatively long-term durability. Although the type of metal cask is chosen, the use of cementitious material is inevitable at least for the cask foundation and the facilities for the protection of dry storage structures. Upon being placed, the performance of concrete begins to deteriorate from the intrinsic change of cement and the physical/ chemical environmental conditions. Thus it is necessary to evaluate the durability of a concrete for the increase of reliability and safety of the whole system during the designed life time. Considering the dry storage system of spent fuel is the item which can create a lot of added value, the development of a dry storage cask is usually initiated by private enterprises among developed countries. The detail research results and specific design criteria for the safety assessment of a concrete cask have not been revealed to the public well. In this paper, the major expected degradation factors and related degradation models of concrete casks were investigated as part of the safety assessment by taking account of the site where Korea industrial nuclear power plants are located

  18. 40 CFR 90.427 - Catalyst thermal stress resistance evaluation.

    Science.gov (United States)

    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... catalyst conversion efficiency for Phase 1 engines. The thermal stress is imposed on the test catalyst...

  19. Micro-Climate Evaluation System in Thermal Mines

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    A fuzzy evaluation method was used to evaluate the microclimate in thermal mines. A theoretical model of a microclimate evaluation system was designed and membership functions of the evaluation indices in the system were established. An analytical hierarchy process (AHP) was used to analyze the weight of the evaluation indices and their methods of calculation. Software for this evaluation system was developed and used for the evaluation of the microclimate of 714 sections in a mine. It is shown that the evaluation results correspond completely with the actual situation. This evaluation system and the software can be applied in thermal mines.

  20. Behavior of spent fuel and cask components after extended periods of dry storage

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) promulgated 10 CFR Part 72, Title 10, for the independent storage of spent nuclear fuel and high-level radioactive waste outside reactor spent fuel pools. Part 72 currently limits the license term for an independent spent fuel storage installation to 20 years from the date of issuance. Licenses may be renewed by the Commission at or before the expiration of the license term. Applications for renewal of a license should be filed at least two years prior to the expiration of the existing license. In preparation for possible license renewal, the NRC Office of Nuclear Material and Safeguards, Spent Fuel Project Office, is developing the technical basis for renewals of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level radioactive waste at independent spent fuel storage installation sites. An analysis of past performance of selected components of these systems is required as part of that technical basis. In the years 1980 through the early 1990, the Department of Energy (DOE) procured four prototype dry storage casks for testing at the Idaho National Engineering and Environmental Laboratory (INEEL): Castor-V/21, MC-10, TN-24P, and VSC-17. The primary purpose of the testing was to benchmark thermal and radiological codes and to determine the thermal and radiological characteristics of the casks. A series of examinations in 1999 and early 2000 to investigate the integrity of the Castor V/21 cask were undertaken. There is no evidence of significant degradation of the Castor V/21 cask systems important to safety from the time of initial loading of the cask in 1985 up to the time of testing in 1999. (author)

  1. Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jae Won; Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1998-12-31

    Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous`s empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing. 6 refs., 8 figs. (Author)

  2. Performance of CASTOR{sup R} HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsmeier, Marco; Vossnacke, Andre [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany)

    2008-07-01

    On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR{sup R} HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR{sup R} HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR{sup R} HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR{sup R} HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out

  3. Building high-accuracy thermal simulation for evaluation of thermal comfort in real houses

    OpenAIRE

    Nguyen, Hoaison; Makino, Yoshiki; Lim, Azman Osman; TAN, Yasuo; Shinoda, Yoichi

    2013-01-01

    Thermal comfort is an essential aspect for the control and verification of many smart home services. In this research, we design and implement simulation which models thermal environment of a smart house testbed. Our simulation can be used to evaluate thermal comfort in various conditions of home environment. In order to increase the accuracy of the simulation, we measure thermal-related parameters of the house such as temperature, humidity, solar radiation by the use of sensors and perform p...

  4. Systems evaluation of thermal bus concepts

    Science.gov (United States)

    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.

  5. Dynamic Response Analysis of Storage Cask Lid Structure Subjected to Lateral Impact Load of Aircraft Engine Crash

    Energy Technology Data Exchange (ETDEWEB)

    Almomania, Belal; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Lee, Sanghoon [Keimyung Univ., Daegu (Korea, Republic of)

    2015-10-15

    Several numerical methods and tests have been carried out to measure the capability of storage cask to withstand extreme impact loads. Testing methods are often constrained by cost, and difficulty in preparation for several impact conditions with different applied loads, and areas of impact. Instead, analytic method is an acceptable process that can easily apply different impact conditions for the evaluation of cask integrity. The aircraft engine impact is considered as one of the most critical impact accidents on the storage cask that significantly affects onto the lid closure system and may cause a considerable release of radioactive materials. This paper presents a method for evaluating the dynamic responses of one upper metal cask lid closure without impact limiters subjected to lateral impact of an aircraft engine with respect to variation of the impact velocity. An assessment method to predict damage response due to the lateral engine impact onto metal storage cask has been studied by using computer code LS-DYNA. The dynamic behavior of the lid movements was successfully calculated by utilizing a simplified finite element cask model, which showed a good agreement with the previous research. The simulation analyses results showed that no significant plastic deformation for bolts, lid, and the cask body. In this study, the lid opening and sliding displacements are considered as the major factors in initiating the leakage path. This analysis may be useful for evaluating the instantaneous leakage rates in a connection with the sliding and opening displacements between the lid and the flange to ensure that the radiological consequences caused by an aircraft engine crash accident during the storage phase are within the permissible level.

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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.

  9. Evaluation of properties and thermal stress field for thermal barrier coatings

    Institute of Scientific and Technical Information of China (English)

    王良; 齐红宇; 杨晓光; 李旭

    2008-01-01

    In order to get thermal stress field of the hot section with thermal barrier coating (TBCs), the thermal conductivity and elastic modulus of top-coat are the physical key properties. The porosity of top-coat was tested and evaluated under different high temperatures. The relationship between the microstructure (porosity of top-coat) and properties of TBCs were analyzed to predict the thermal properties of ceramic top-coat, such as thermal conductivity and elastic modulus. The temperature and stress field of the vane with TBCs were simulated using two sets of thermal conductivity data and elastic modulus, which are from literatures and this work, respectively. The results show that the temperature and stress distributions change with thermal conductivity and elastic modulus. The differences of maximum temperatures and stress are 6.5% and 8.0%, respectively.

  10. Validation of elastic-plastic computer analyses for use in nuclear waste shipping cask design

    International Nuclear Information System (INIS)

    GA Technologies designed the Defense High Level Waste (DHLW) Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The DHLW cask has a thick-walled stainless steel body and incorporates integral stainless steel impact limiters that protect the two ends of the cask during the hypothetical accident condition 30-ft free drop. These integral impact limiters absorb the drop energy through gross plastic deformations. GA used elastic-plastic computer codes developed at Los Alamos and Lawrence Livermore Laboratories, HONDOII and DYNA3D, to analyze for this non-linear behavior. In order to evaluate the analyses, GA developed elastic-plastic stress criteria that were adapted from the ASME Boiler and Pressure Vessel Code, Division I, Section III. This innovative design and analytical approach required test verification. Therefore, SNL performed 30-ft drop and puncture tests on a half-scale model of the DHLW cask. The testing confirmed that the analytical approach works and results in a safe, conservative design

  11. Interfacing the existing cask fleet with the MRS

    International Nuclear Information System (INIS)

    This paper reports that the Department of Energy (DOE) is considering the possibility of using the existing fleet of casks to achieve spent fuel receipt at the Monitored Retrievable Storage (MRS) facility. The existing cask fleet includes the NLI-1/2, the NAC-LWT, the TN-8 (and TN-8L), the TN-9, and the IF-300 casks. Other casks may be available, but their status is not certain. Use of the existing cask fleet at the MRS places additional design requirements on the system, and specifically affects the cask-to-MRS interface. The decision to use the existing cask fleet also places additional demands on training needs and operator certification, and the configuration management system. Some existing cask designs may not be able to mate with a bottom opening hot cell MRS. Use of the existing cask fleet also greatly increases the number of shipments that must be received, to the point that a facility larger than originally envisioned may be required

  12. Studies and research concerning BNFP: cask handling equipment standardization

    International Nuclear Information System (INIS)

    This report covers the activities of one of the sub-tasks within the Spent LWR Fuel Transportation Receiving, Handling, and Storage program. The sub-task is identified as Cask Handling Equipment Standardization. The objective of the sub-task specifies: investigate and identify opportunities for standardization of cask interface equipment. This study will examine the potential benefits of standardized yokes, decontamination barriers and special tools, and, to the extent feasible, standardized methods and software for handling the variety of casks presently available in the US fleet. The result of the investigations is a compilation of reports that are related by their common goal of reducing cask turnaround time

  13. Rail tiedown tests with heavy casks for radioactive shipments

    International Nuclear Information System (INIS)

    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

  14. Drop test of transportable storage cask

    International Nuclear Information System (INIS)

    It is being planned to transport the transportable storage casks again after their storage period of several decades, so metal gaskets are used as seal material in their lids in place of rubber o-rings which deteriorate during the storage period. Since the slightest dislocation of the lids causes seal performance deterioration in the metal gaskets, it is necessary to establish a simulation technology which accurately estimates the dislocation in order to design a rigid lid structure to protect against the impact loads under 9 m drop condition. A 1:3 scale model of the transportable storage cask developed by Hitz for BWR spent fuel rods were manufactured and 9 m drop tests were performed. Measured dislocations of the lids were confirmed within the allowable limit and they were found to be accurately simulated. (author)

  15. MCO loading and cask loadout technical manual

    Energy Technology Data Exchange (ETDEWEB)

    PRAGA, A.N.

    1998-10-01

    A compilation of the technical basis for loading a multi-canister overpack (MCO) with spent nuclear fuel and then placing the MCO into a cask for shipment to the Cold Vacuum Drying Facility. The technical basis includes a description of the process, process technology that forms the basis for loading alternatives, process control considerations, safety considerations, equipment description, and a brief facility structure description.

  16. Concrete Spent Fuel Cask Criticality Calculation

    International Nuclear Information System (INIS)

    A preliminary analysis of the concrete cask for the intermediate dry storage of the spent fuel of NPP Krsko should include an estimation of the effective multiplication factor. Assuming 16x16 fuel elements, 4.3% initial enrichment, 45 GWd/tU burnup and 10 years cooling time, a concrete spent fuel capacity of 10 spent fuel assemblies is proposed. Fuel assemblies are placed inside inner cavity in a 'basket' - a boron (1%) doped steel structure. Heavy concrete (25% Fe), 45 cm thick, is enclosed in a carbon steel shell. There is also a stainless steel (SS304) lining of the storage cavity. Isotope inventory of the spent fuel after a 10 years cooling time is calculated using ORIGEN-S functional module of the SCALE-4.2 code package. The effective multiplication factor keff of dry (helium filled) and wet (water filled) cask for fresh and used fuel is calculated using CSAS4 Monte Carlo method based control module of the same SCALE-4.2 code package. The obtained results of keff of the dry cask for fresh and spent fuel are well below the required 0.95 value, but those for the water filled cask are above this value. Therefore, several additional calculations of the keff varying the thickness of a boral basket structure which had replaced the stainless steel one were done. It turned out that at least a 1.5 cm thick boral wall was needed to meet the required 0.95 value for keff. (author)

  17. Experimental evaluation of thermal energy storage

    Science.gov (United States)

    Asbury, J. G.; Hersh, H. N.

    1980-01-01

    The technical performance of commercially available thermal energy storage (TES) residential heating units under severe weather conditions is discussed. The benefits and costs of TES to the user and utility companies were assessed. The TES issues, research and development needs, and barriers to commercialization were identified. The field tests which determined the performance characteristics for the TES are described and the TES systems, which included both ceramic and hydronic systems, are compared.

  18. Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias

    International Nuclear Information System (INIS)

    Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)

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

    Science.gov (United States)

    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.

  20. Thermal aging evaluation of casting stainless steel under BWR environment

    International Nuclear Information System (INIS)

    Effect of thermal aging under BWR condition on material properties of casting stainless steel were evaluated by such as Charpy impact test, using replaced BWR component material. Solution heat treatment was performed to the same material and the material properties were obtained. Comparing each material test results, impact value of thermal aging material was lower than solution heat treatment material. By the results, thermal aging effect on material properties under BWR condition was confirmed. The material properties were compared with model equation using PLM evaluation and conservativeness of model equation was confirmed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    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.

  2. Evaluating local and overall thermal comfort in buildings using thermal manikins

    Energy Technology Data Exchange (ETDEWEB)

    Foda, E.

    2012-07-01

    Evaluation methods of human thermal comfort that are based on whole-body heat balance with its surroundings may not be adequate for evaluations in non-uniform thermal conditions. Under these conditions, the human body's segments may experience a wide range of room physical parameters and the evaluation of the local (segmental) thermal comfort becomes necessary. In this work, subjective measurements of skin temperature were carried out to investigate the human body's local responses due to a step change in the room temperature; and the variability in the body's local temperatures under different indoor conditions and exposures as well as the physiological steady state local temperatures. Then, a multi-segmental model of human thermoregulation was developed based on these findings to predict the local skin temperatures of individuals' body segments with a good accuracy. The model predictability of skin temperature was verified for steady state and dynamic conditions using measured data at uniform neutral, cold and warm as well as different asymmetric thermal conditions. The model showed very good predictability with average absolute deviation ranged from 0.3-0.8 K. The model was then implemented onto the control system of the thermal manikin 'THERMINATOR' to adjust the segmental skin temperature set-points based on the indoor conditions. This new control for the manikin was experimentally validated for the prediction of local and overall thermal comfort using the equivalent temperature measure. THERMINATOR with the new control mode was then employed in the evaluation of localized floor-heating system variants towards maximum energy efficiency. This aimed at illustrating a design strategy using the thermal manikin to find the optimum geometry and surface area of a floor-heater for a single seated person. Furthermore, a psychological comfort model that is based on local skin temperature was adapted for the use with the model of human

  3. Thermal Protection Systems Nondestructive Evaluation Tool Project

    Data.gov (United States)

    National Aeronautics and Space Administration — To address NASA's need for evaluation of bondline and in-depth integrity for lightweight rigid and/or flexible ablative materials, Physical Optics Corporation (POC)...

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

    Science.gov (United States)

    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.

  5. Thermal fatigue evaluation of piping system tee-connections (THERFAT)

    International Nuclear Information System (INIS)

    Thermal fatigue is one significant long-term degradation mechanism in nuclear power plants (NPP), in particular, as operating plants become older and life time extension activities have been initiated. In general, the common thermal fatigue issues are understood and controlled by plant instrumentation systems. However, incidents in some plants indicate that certain piping system Tees are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentation. The THERFAT project (Thermal fatigue evaluation of piping system Tee-connections) had been initiated to advance the accuracy and reliability of thermal fatigue load determination and to outline a science based but still practical methodology for managing thermal fatigue risks in Tee-connections susceptible to high cyclic temperature fluctuations. The THERFAT project was carried out by 16 international Partners (utilities, plants vendors/manufacturers, consultant engineers, research institutes) as a cost shared action funded by the European Commission. (orig.)

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  8. 78 FR 73456 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI.... SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage... Approved Spent Fuel Storage Casks'' to include Amendment No. 9 to Certificate of Compliance (CoC) No....

  9. 78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI... public comment period. The document proposed to amend the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the ``List of Approved Spent...

  10. A conceptual redesign of an Inter-Building Fuel Transfer Cask

    International Nuclear Information System (INIS)

    The Inter-Building Fuel Transfer Cask, referred to as the IBC, is a lead shielded cask for transporting subassemblies between buildings on the Argonne National Laboratory-West site near Idaho Falls, Idaho. The cask transports both newly fabricated and spent reactor subassemblies between the Experimental Breeder Reactor-II (EBR-II), the Fuel Cycle Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The IBC will play a key role in the Integral Fast Reactor (IFR) fuel recycling demonstration project. This report discusses a conceptual redesign of the IBC which has been performed. The objective of the conceptual design was to increase the passive heat removal capabilities, reduce the personnel radiation exposure and incorporate enhanced safety features into the design. The heat transfer, radiation and thermal-hydraulic properties of the IBC were analytically modelled to determine the principal factors controlling the desip. The scoping studies that were performed determined the vital physical characteristics (i.e., size, shielding, pumps, etc.) of the MC conceptual design

  11. Evaluation of Structure Influence on Thermal Conductivity of Thermal Insulating Materials from Renewable Resources

    Directory of Open Access Journals (Sweden)

    Jolanta VĖJELIENĖ

    2011-07-01

    Full Text Available The development of new thermal insulation materials needs to evaluate properties and structure of raw material, technological factors that make influence on the thermal conductivity of material. One of the most promising raw materials for production of insulation material is straw. The use of natural fibres in insulation is closely linked to the ecological building sector, where selection of materials is based on factors including recyclable, renewable raw materials and low resource production techniques In current work results of research on structure and thermal conductivity of renewable resources for production thermal insulating materials are presented. Due to the high abundance of renewable resources and a good its structure as raw material for thermal insulation materials barley straw, reeds, cattails and bent grass stalks are used. Macro- and micro structure analysis of these substances is performed. Straw bales of these materials are used for determining thermal conductivity. It was found that the macrostructure has the greatest effect on thermal conductivity of materials. Thermal conductivity of material is determined by the formation of a bale due to the large amount of pores among the stalks of the plant, inside the stalk and inside the stalk wall.http://dx.doi.org/10.5755/j01.ms.17.2.494

  12. Numerical models for evaluation thermal conductivity of coatings

    Directory of Open Access Journals (Sweden)

    Švantner M.

    2008-12-01

    Full Text Available This paper is dealing with simulation and model development for the evaluation of thermal conductivity of coatings by the Laser Quasistatic Thermography (LQT method. The main principles of the measurement method are introduced and the process of thermal conductivity evaluation based on numerical simulation is presented. The evaluation requires special procedure to simulate thermal process induced by laser pulse in coating on some substrate. The thickness of the coating is manifold less than the thickness of the substrate and total sample surface. In numerical system Cosmos/M there are created two suitable models: "Shell-Clink-Solid" model and model based on physical similarity. In this paper there are also described characteristics of both models and their comparison with classical axisymmetric and volume models.

  13. US cask requirements and industry capability survey

    International Nuclear Information System (INIS)

    The objectives of this paper are to provide an estimate of spent fuel shipping cask requirements for reactor to away-from-reactor (AFR) storage facility shipments from the present time until late in this century and to determine and document the willingness and capability of private industry to provide required future transportation services. In order to meet this objective, the Transportation Technology Center at Sandia National Laboratories sponsored Teledyne Energy Systems to conduct a survey of US industry. Results of tasks completed to carry out the objectives are reviewed

  14. Considerations on the construction testing of the CASTOR registered HAW 28M cask with respect to the traffic law in the view of the responsible authority BAM

    International Nuclear Information System (INIS)

    The authors reflect the construction testing of the CASTOR registered HAW 28M cask with respect to the traffic law in the view of the responsible authority BAM. The test procedures are based on the recommendations of the IAEA and the respective national and international legal regulations for the transport of radioactive materials. BAM is performing mechanical and thermal tests to investigate the safety of the containers in case of a severe accident. The radionuclide release has to be restricted to a defined limiting value, the radiation shielding and the nuclear safety have to be ensured. The component test is performed using prototypes of model containers combined with calculations or transferability considerations. The safety evaluation is usually based on experimental tests and numerical analyses.

  15. Considerations for Disposition of Dry Cask Storage System Materials at End of Storage System Life

    International Nuclear Information System (INIS)

    Dry cask storage systems are deployed at nuclear power plants for used nuclear fuel (UNF) storage when spent fuel pools reach their storage capacity and/or the plants are decommissioned. An important waste and materials disposition consideration arising from the increasing use of these systems is the management of the dry cask storage systems' materials after the UNF proceeds to disposition. Thermal analyses of repository design concepts currently under consideration internationally indicate that waste package sizes for the geologic media under consideration may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded into the dry storage canisters currently in use. In the United States, there are already over 1650 of these dry storage canisters deployed and approximately 200 canisters per year are being loaded at the current fleet of commercial nuclear power plants. There is about 10 cubic meters of material from each dry storage canister system that will need to be dispositioned. The concrete horizontal storage modules or vertical storage overpacks will need to be reused, re-purposed, recycled, or disposed of in some manner. The empty metal storage canister/cask would also have to be cleaned, and decontaminated for possible reuse or recycling or disposed of, likely as low-level radioactive waste. These material disposition options can have impacts of the overall used fuel management system costs. This paper will identify and explore some of the technical and interface considerations associated with managing the dry cask storage system materials. (authors)

  16. Development of scaling laws of heat removal and CFD assessment in concrete cask air path

    International Nuclear Information System (INIS)

    Highlights: • Vertical concrete cask was studied for PWR spent fuel dry storage. • Scaling laws were derived for facilities between prototype and half-scale model. • Computational Fluid Dynamics analysis was performed with 3D mesh generation. • Thermal radiation was considered with heat conduction and natural convection. - Abstract: This study investigates heat transfer in a concrete cask such as one used at intermediate storage facilities of PWR spent fuels. Sufficient removal of decay heat is necessary not to damage fuel cladding that functions as a radioactive materials barrier. The experimental design parameters were derived in the half-scale model for the assessment of the design analysis methodology including a CFD tool. The scaling methodology was developed to design the half-scale model of the concrete cask in the spent fuel dry storage through scaling analysis. As one of the most important scaling laws, the requirement of similarity was selected for the temperature rise between the inlet and the exit in the air path. Based on the natural circulation in the channel, the scaling law was derived for total canister power maintaining the similarity of the temperature rise. Then, the temperature calculation and the flow analysis were performed in concrete cask facilities for the prototype and the half scale model using Computational Fluid Dynamics code. Through the CFD simulations, the similarity of the temperature rise was demonstrated well between the inlet and the exit, and the exit temperature was well maintained between the prototype and the half scale model. Also the scaling ratios of air mass flow rate and exit velocity obtained by the scaling analysis were in good agreement with those predicted by CFD analysis

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

    International Nuclear Information System (INIS)

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

  18. An assessment methodology for thermal energy storage evaluation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Frequency response function method for evaluation of thermal striping phenomena

    International Nuclear Information System (INIS)

    A rational analysis method of thermal stress induced by fluid temperature fluctuation is developed, by utilizing frequency response characteristics of structures. High frequency components of temperature fluctuation are attenuated in the transfer process from fluids to structures. Low frequency components hardly induce thermal stress since temperature homogenization in structures. Based on investigations of frequency response mechanism of structures of fluid temperature, a frequency response function of structures was derived, which can predict stress amplitudes on structural surfaces from fluid temperature amplitudes and frequencies. This function is formulated by separation of variables, and is composed of an effective heat transfer function and an effective thermal stress one. The frequency response function method appears to evaluate thermal stress rationally and to give information on damageable frequency range of structures. (author)

  20. Development of structural response diagram approach to evaluation of thermal stress caused by thermal striping

    Energy Technology Data Exchange (ETDEWEB)

    Kasahara, Naoto; Yacumpai, Apisara [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Takasho, Hideki

    1999-02-01

    At incomplete mixing area of high temperature and low temperature fluids near the surface of structures, temperature fluctuation of fluid gives thermal fatigue damage to wall structures. This thermohydraulic and thermomechanical coupled phenomenon is called thermal striping, which has so complex mechanism and sometimes causes crack initiation on the structural surfaces that rational evaluation methods are required for screening rules in design codes. In this study, frequency response characteristics of structures and its mechanism were investigated by both numerical and theoretical methods. Based on above investigation, a structural response diagram was derived, which can predict stress amplitude of structures from temperature amplitude and frequency of fluids. Furthermore, this diagram was generalized to be the Non-dimensional structural response diagram by introducing non-dimensional parameters such as Biot number, non-dimensional frequency, and non-dimensional stress. The use of the Non-dimensional structural response diagram appears to evaluate thermal stress caused by thermal striping, rapidly without structural analysis, and rationally with considering attenuation by non-stationary heat transfer and thermal unloading. This diagram can also give such useful information as sensitive frequency range to adjust coupled thermohydraulic and thermomechanical analysis models taking account of four kinds of attenuation factors: turbulent mixing, molecular diffusion, non-stationary heat transfer, and thermal unloading. (author)

  1. DOE procurement activities for spent fuel shipping casks

    International Nuclear Information System (INIS)

    This paper discusses the DOE cask development program established to satisfy the requirements of the NWPA. The program is designed to provide safe efficient casks on a timely schedule. The casks will be certified by the NRC in compliance with the 1987 amendment to NWPA. Private industry will be used to the maximum extent. DOE will encourage use of present cask technology, but will not hesitate to advance the state-of-the-art to improve efficiency in transport operations, provided that safety is not compromised. DOE will support the contractor's efforts to advance the state-of-the-art by maintaining a technical development effort that will respond to the common needs of all the contractors. DOE and the cask contractors will develop comprehensive and well integrated programs of test and analysis for cask certification. Finally, the DOE will monitor the cask development program within a system that fosters early identification of improvement opportunities as well as potential problems, and is sufficiently flexible to respond quickly yet rationally to assure a fully successful program

  2. Decontamination of transport casks and of spent fuel storage facilities

    International Nuclear Information System (INIS)

    The present document provides an analysis of the technical papers presented at the meeting as well as a summary of the panel discussion. Conclusions and Recommendations: The meeting agreed that the primary source of contamination of transport casks is the production of radioactive isotopes in nuclear fuel and activation products of fuel components in nuclear reactors. The type, amount of mechanism for the release of these isotopes depend on the reactor type and fuel handling process. The widespread use of pools for the storage and handling of fuel provides an easy path for the transfer of contamination. Control of pool water conditions is essential for limiting the spread of contamination. For plants where casks are immersed in pools for loading, the immersion times should be minimised. Casks should be designed for ease of decontamination. The meeting discussed the use of stainless steel and suitable paints for coating casks. Designers should consider the appropriate coating for specific applications. The use of pressurized water for decontamination is recommended whenever possible. A number of commercially available reagents exist for decontaminating cask external surfaces. More work, however, is needed to cope with Pressurized Water Reactor crud within casks. Leaking fuel should be identified and isolated before storage in pools. Basic studies of the uptake and release of contamination from cask surfaces should be initiated. Standardization of methods of contamination measurement and instrumentation should be instituted. Refs, figs and tabs

  3. Differences of Technical Requirements Between Transportation and Storage Metal Casks

    International Nuclear Information System (INIS)

    The worldwide demand of storage facilities for spent fuels discharged from nuclear power stations is increasing to maintain sustainable operation of the nuclear power stations. The spent fuels are stored at first in the fuel pools (wet storage). When the spent fuels exceed the pool storage capacity, the fuels are transferred to the other storage facility located at reactor or away from reactor, which often adopts a dry storage technology. To use metal casks is one of the options for the dry storage facilities, and some storage facilities have already utilized large metal casks, whose original design concept were developed to transport the spent fuels from nuclear power stations to a reprocessing plant by trains, trucks or by sea-going vessels. It is widely understood that the technology of transportation casks developed up to now is able to apply to the storage casks without any significant design changes. Technical requirements on the design are discussed between the storage cask and the transportation cask to confirm of the understanding based on the assumption that the large metal cask is used for transportation and storage respectively. (author)

  4. Robotic radiation survey and analysis system for radiation waste casks

    International Nuclear Information System (INIS)

    Sandia National Laboratories (SNL) and the Hanford Engineering Development Laboratories have been involved in the development of remote systems technology concepts for handling defense high-level waste (DHLW) shipping casks at the waste repository. This effort was demonstrated the feasibility of using this technology for handling DHLW casks. These investigations have also shown that cask design can have a major effect on the feasibility of remote cask handling. Consequently, SNL has initiated a program to determine cask features necessary for robotic remote handling at the waste repository. The initial cask handling task selected for detailed investigation was the robotic radiation survey and analysis (RRSAS) task. In addition to determining the design features required for robotic cask handling, the RRSAS project contributes to the definition of techniques for random selection of swipe locations, the definition of robotic swipe parameters, force control techniques for robotic swipes, machine vision techniques for the location of objects in 3-D, repository robotic systems requirements, and repository data management system needs

  5. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each casks neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  6. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. [Audin (Lindsay), Ossining, NY (United States)

    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.

  7. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. (Audin (Lindsay), Ossining, NY (United States))

    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.

  8. Nuclear cask testing films misleading and misused

    International Nuclear Information System (INIS)

    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

  9. Validation of CFD-methods to predict heat transfer and temperatures during the transport and storage of casks under a cover

    Energy Technology Data Exchange (ETDEWEB)

    Leber, A. [WTI Wissenschaftlich-Technische-Ingenieurberatung GmbH (Germany); Graf, W. [GNS Gesellschaft fuer Nuklear-Service mbH (Germany); Hueggenberg, R. [GNB Gesellschaft fuer Nuklear-Behaelter mbH (Germany)

    2004-07-01

    With respect to the transport of casks for radioactive material, the proof of the safe heat removal can be accomplished by validated calculation methods. The boundary conditions for thermal tests for type B packages are specified in the ADR based on the regulations defined by the International Atomic Energy Agency. The varying boundary conditions under transport or storage conditions are based on the varying thermal conditions true for different cask types. In most cases the cask will be transported in lying position under a cover (e.g. canopy or tarpaulin) and stored in standing position in an array with other casks. The main heat transport mechanisms are natural convection and thermal radiation. The cover or the storage building are furnished with vents that create an air flow, which will improve the natural convection. Depending on the thermal boundary conditions, the cask design and the heat power, about 50 - 95% of the heat power will be removed from the finned cask surface by natural convection. Consequently the convection by air flow is the main heat transport mechanism. The air flow can be approximated with analytical methods by solving the integral heat and flow balances for the domain. In a stationary state the overpressure due the buoyancy and the pressure loss in the flow resistances are equal. Based on the air flow, the relevant temperatures of the cask can be calculated in an iterative process. Due to the fast development of numerical calculation methods and computer hardware, the use of Computational- Fluid-Dynamics(CFD) calculations plays an important role. CFD-calculations are based on solving the equations of conservation (Navier-Stokes equations) using a finite element mesh or a finite volume mesh of the model. For a finned cask lying under a cover, where the main contributing element for heat removal is natural convection in combination with the thermal radiation, a CFD-calculation can be the most appropriate method. Common CFD-Codes are FLUENT

  10. Validation of CFD-methods to predict heat transfer and temperatures during the transport and storage of casks under a cover

    International Nuclear Information System (INIS)

    With respect to the transport of casks for radioactive material, the proof of the safe heat removal can be accomplished by validated calculation methods. The boundary conditions for thermal tests for type B packages are specified in the ADR based on the regulations defined by the International Atomic Energy Agency. The varying boundary conditions under transport or storage conditions are based on the varying thermal conditions true for different cask types. In most cases the cask will be transported in lying position under a cover (e.g. canopy or tarpaulin) and stored in standing position in an array with other casks. The main heat transport mechanisms are natural convection and thermal radiation. The cover or the storage building are furnished with vents that create an air flow, which will improve the natural convection. Depending on the thermal boundary conditions, the cask design and the heat power, about 50 - 95% of the heat power will be removed from the finned cask surface by natural convection. Consequently the convection by air flow is the main heat transport mechanism. The air flow can be approximated with analytical methods by solving the integral heat and flow balances for the domain. In a stationary state the overpressure due the buoyancy and the pressure loss in the flow resistances are equal. Based on the air flow, the relevant temperatures of the cask can be calculated in an iterative process. Due to the fast development of numerical calculation methods and computer hardware, the use of Computational- Fluid-Dynamics(CFD) calculations plays an important role. CFD-calculations are based on solving the equations of conservation (Navier-Stokes equations) using a finite element mesh or a finite volume mesh of the model. For a finned cask lying under a cover, where the main contributing element for heat removal is natural convection in combination with the thermal radiation, a CFD-calculation can be the most appropriate method. Common CFD-Codes are FLUENT

  11. Evaluating transport coefficients in real time thermal field theory

    CERN Document Server

    Mallik, S

    2012-01-01

    Transport coefficients in a hadronic gas have been calculated earlier in the imaginary time formulation of thermal field theory. The steps involved are to relate the defining retarded correlation function to the corresponding time-ordered one and to evaluate the latter in the conventional perturbation expansion. Here we carry out both the steps in the real time formulation.

  12. Application of liquid crystals in thermal nondestructive evaluation

    International Nuclear Information System (INIS)

    In recent years, thermal nondestructive evaluation using Cholestric liquid crystals have found wide applications in industry. Thermography using Cholesteric liquid crystals can be used for detection of nonbonds in metallic composites, hot spots in electronic circuits and preliminary examination of welded pressure vessels. This paper presents the results of experiments on thermography of components using encapsulated liquid crystals. (author)

  13. What drives Greek consumer preferences for cask wine?

    DEFF Research Database (Denmark)

    Chrysochou, Polymeros; Corsi, A. M.; Krystallis Krontalis, Athanasios

    2012-01-01

    a sustainable eco-friendly positioning. Originality/value – This study contributes to the understanding of what drives consumers’ preferences for cask wine, something that few studies have done until now. Moreover, this is the first study to use the BWS method for this type of product.......Purpose – Cask wine (bag-in-box, soft pack) has not received considerable attention in wine marketing research, but interest among winemakers and consumers has been increasing steadily. However, little is known about what drives consumer preferences for cask wine and, furthermore, what the profile...

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

    Energy Technology Data Exchange (ETDEWEB)

    Qu, Jianmin [Northwestern Univ., Evanston, IL (United States); Bazant, Zdenek [Northwestern Univ., Evanston, IL (United States); Jacobs, Laurence [Georgia Inst. of Technology, Atlanta, GA (United States); Guimaraes, Maria [Electrical Power Research Institute, Palo Alto, CA (United States)

    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

  15. Development of dual-purpose metal cask for interim storage of spent nuclear fuel (1). Outline of cask structure

    International Nuclear Information System (INIS)

    Spent fuels discharged from nuclear power plants in Japan are planed to be reprocessed at the nuclear fuel recycle plant under construction at Rokkasho-mura. Since the amount of the spent fuels exceeds that of recycled fuel, the spent fuels have to be properly stored and maintained as recycle fuel resource until the beginning of the reprocessing. For that sake, interim storage installations are being constructed outside the nuclear power plants by 2010. The storage dry casks have been practically used as the interim storage in the nuclear power plants. From this reason, the storage system using the storage dry casks is promising as the interim storage installations away form the reactors, which are under discussion. In the interim storage facilities, the storage using the dry cask of the storage metal cask with business showings, having the function of transportation is now under discussion. By employing transportation and storage dual-purpose cask, the repack equipments can be exhausted, and the reliability of the interim storage installations can be increased. Hitachi, Ltd. has been developing the high reliable and economical transportation and storage dry metal cask. In this report, the outline of our developing transportation and storage dry cask is described. (author)

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

    International Nuclear Information System (INIS)

    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

  17. Country report France [Operation and maintenance of spent fuel storage and transportation casks/containers

    International Nuclear Information System (INIS)

    Transportation from Electricite de France and other foreign utilities to COGEMA La Hague reprocessing plant is performed with one family of casks in the 100 ton class. The experience gained in transport cask design and operation has resulted in design of transport/storage and storage only systems. There are 6 cask types for transportation only and 10 cask types for dual purpose storage and transportation. French authorities approve each cask design. Cask vendors provide training and assistance to users as well as a transportation file containing all actions and recording inspections of the cask. Maintenance frequencies are determined according to design an experience and maintenance specifications prepared. The extent of maintenance is at three levels: inspections on arrival and departure, every 3 years or 15 transports and every 6 years or 60 transports. According to French experience the cask maintenance costs over lifetime are the same as the cost of the cask itself. (author)

  18. Development of tipping-over analysis of cask subjected to earthquake strong motion

    International Nuclear Information System (INIS)

    Since a cask is vertically oriented during loading in cask-storage, it is necessary to investigate the integrity of the cask against tipping-over during strong earthquakes. The rocking and sliding behavior of the cask during strong earthquakes can be analyzed as a dynamic vibration problem for a rigid cylinder. In this paper, in order to clarify the tipping-over characteristics of a cask during strong earthquakes, the authors applied the Distinct Element Method (DEM) to the seismic response analysis of the cask. DEM was introduced by Cundall P.A. in 1971. It is based on the use of an explicit numerical scheme. The cask was considered to be a rigid polygonal element, which satisfied the equation of motion and the law of action and reaction. They examined the applicability of this code by comparison with experimental results obtained from shaking table tests using scale model casks considering the dimension of a 100 ton class full-scale cask

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  2. Evaluation of infrared collimators for testing thermal imaging systems

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

  4. An Evaluation on the Smart Composite Damaged by Thermal Shock

    Science.gov (United States)

    Lee, Jin Kyung; Lee, Sang Pill; Park, Young Chul; Lee, Joon Hyun

    A shape memory alloy (SMA) as part of some products and system has been used to keep their shape at any specified temperature. By using this characteristic of the shape memory alloy it can be solved the problem of the residual stress by difference of coefficients of thermal expansion between reinforcement and matrix within composite. In this study, TiNi/Al6061 shape memory alloy composite was fabricated through hot press method, and the optimal fabrication condition was created. The bonding effect of the matrix and the reinforcement within the SMA composite was strengthened by cold rolling. The SMA composite can be applied as the part of airplane and vessel, and used under tough condition of repetitive thermal shock cycles of high and low temperatures. Therefore, the thermal shock test was performed for the SMA composite, and mechanical properties were evaluated. The tensile strength of the SMA composite showed a slight decline with the thermal shock cycles. In addition, acoustic emission (AE) technique was used to quantify the microscopic damage behavior of cold rolled TiNi/Al6061 shape memory alloy composite that underwent thermal shock cycles. The damage degree on the specimen that underwent thermal shock cycles was discussed. Actually AE parameters such as AE event, count and energy was analyzed, and these parameters was useful to evaluate the damage behavior and degree of the SMA composite. The waveform of the signal caused by debonding was pulse type, and showed the frequency range of 160 kHz, however, the signal by the fiber fracture showed the pulse type of high magnitude and frequency range of 220 kH.

  5. Operating Experience in Spent Fuel Storage Casks

    International Nuclear Information System (INIS)

    A safe storage of spent fuels has been considered as one of the inevitable tasks for TEPCO for the last few decades. In order to increase flexibility for the fuel storage measures, TEPCO has been storing spent fuels in an on-site dry storage facility at Fukushima-Daiichi Nuclear Power Station. Since 1995, more than 400 fuel assemblies have been safely store. Integrity of storage casks and fuels were carefully checked by periodical inspections, which were conducted in 2000 and 2005. The next investigation will be held within a few years in order to verify the safety conditions even after a 15-year storage. These series of inspections will give plenty of useful data for the design and operation of the Mutsu facility, which will be the first off-site interim spent fuel storage facility away from any reactor site in Japan. (author)

  6. Fatigue evaluation of piping connections under thermal transients

    International Nuclear Information System (INIS)

    In designing nuclear power plant piping, thermal transients, caused by non-steady operation conditions, should be considered. These events may reduce considerably the lifetime of the pipes, creating the necessity of using structural elements designed in such a way to minimize the acting thermal stresses. Typical examples of the usage of these elements are the connections between pipes of small and large diameters, in which it is usually used a weldolet. Nevertheless, in some situations, the thermal stresses caused by the transients are greater than the allowable limits, being, in this case, an alternative for best results, the introduction of a special fitting replacing the weldolet. Such a fitting is designed in a way to permit a better distribution of the stresses, reducing its maximum value to acceptable levels. This paper intends to present a fatigue evaluation of a connection, using the above mentioned fitting, when subjected to a load expressed in terms of a step thermal gradient, varying from 263 deg to 40 deg C. Two different methodologies are used in this analysis: (a) Determination of the temperature distribution from the heat transfer equations for piping, being the stresses calculated according to ASME III NB-3600. (b) Thermal and stress analyses using axisymmetric elements, according to the rules presented at ASME III NB-3200. In the first case, named simplified analysis, the computer code used is the PIPESTRESS, while in the second case, the ANSYS program was adopted

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

    Science.gov (United States)

    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

  8. Impact velocity vs. target hardness relationships for equivalent response of cask structures

    International Nuclear Information System (INIS)

    In this paper, impact velocity vs. target hardness relationships for cask structures are reviewed. The relationships are based on equivalent cask responses in terms of equal deceleration or similar cask damages. By examining several past cask or container tests as well as some analytical results, some conclusions can be drawn about the relationship between target hardness and equivalent impact velocities. This relationship clearly shows that the cask response to impact is cask-dependent and that the rigid sphere impact model results in an unconservative estimate of equivalent velocity

  9. Experience with certifying borated stainless steel as a shipping cask basket material

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, D.G. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Nickell, R.E. (Applied Science and Technology, Inc., Poway, CA (USA))

    1990-01-01

    The original cask designs for a cask demonstration project featured fuel baskets constructed of borated stainless steel (bss) as a structural material. The project is intended to demonstrate casks that can be used for both shipping and storing spent nuclear fuel assemblies. The baskets were intended to maintain the fuel assemblies in a subcritical array for both normal and accident conditions. The Nuclear Regulatory Commission, however, judged bss to be unacceptable as a structural material. The cask designs were subsequently modified. The knowledge gained during this cask demonstration project may be applicable to development of bss as a basket material in future cask design. 6 refs., 2 figs., 2 tabs.

  10. Experience with certifying borated stainless steel as a shipping cask basket material

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, D.G. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Nickell, R.E. [Applied Science and Technology, Poway, CA (United States)

    1990-10-01

    This paper discusses the original cask designs for a cask demonstration project that has featured fuel baskets constructed of borated stainless steel (bss) as a structural material. The project is intended to demonstrate casks that can be used for both shipping and storing spent nuclear fuel assemblies. The baskets were intended to maintain the fuel assemblies in a subcritical array for both normal and accident conditions. The Nuclear Regulatory Commission, judged bss to be unacceptable as a structural material. The cask designs were subsequently modified. The knowledge gained during this cask demonstration project may be applicable to development of bss as a basket material in future cask design.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

  14. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.

  15. Thermal Comfort While Sitting on Office Chairs – Subjective Evaluations

    Directory of Open Access Journals (Sweden)

    Zoran Vlaović

    2012-12-01

    Full Text Available Thermal comfort is related to human physiological reactions. In order to maintain a constant internal temperature, the human body must dissipate heat in a warm climate, and prevent heat losses in a cold climate. The overall sensation of comfort accompanies the warmest part of the body in a warm environment and the coldest one in a cold environment. Chair design and clothing may affect the difference in sensitivity between certain parts of the body, that is, they may affect thermal comfort. This research focused on subjective sensation of warmth and moisture while sitting on offi ce chairs. The subjective method of evaluating thermal discomfort is based on ISO 7730:2005 standard, according to which a questionnaire was made for this research. Six subjects took part in the research. They were sitting on five different office chairs as they performed their usual jobs in controlled conditions. From the point of view of the evaluation of the sensation of warmth, all chairs were evaluated neutrally. The sensation under the buttocks and thighs was reported to be somewhat warmer, while the sensation on the back was reported to be somewhat colder, which was affected by the design of the back of the chair. No correlation has been proven between the actual temperature and moisture measurements and subjective evaluations of thermal comfort, in spite of a number of direct links. The use of the present method offers the possibility of further research into this subject, which would prove more thoroughly a correlation between design and construction solutions of office chairs and the comfort perceived by sitting persons.

  16. Preliminary safety analysis of criticality for dual-purpose metal cask under dry storage conditions in South Korea

    International Nuclear Information System (INIS)

    Highlights: • DPC is under development led by Korea Radioactive Waste Agency in South Korea. • The results of criticality analysis with respect to design requirements. • The keff under normal and off-normal conditions were 0.36 and 0.46, respectively. • In addition, the keff under a postulated accident condition was evaluated to be 0.94. - Abstract: A dual-purpose metal cask is under development led by Korea Radioactive Waste Agency (KORAD) in Korea, for the dry interim storage and long-distance transportation. This cask comprises a main body made of carbon steel and a stainless steel Dry Shielded Canister (DSC), with stainless steel baskets inside to contain spent fuel assemblies. In this study, nuclear criticality safety analysis was conducted as a part of safety assessment of the metal cask. Analysis to show criticality safety in accordance with regulatory requirements of PWR spent fuel storage was carried out. 10CFR72.124 “Criteria for nuclear criticality safety” and the Regulatory Guide of the American Nuclear Society, ANSI/ANS-57.9 “Design Criteria for an Independent Spent Fuel” and US NRC's “Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility” were employed as regulatory standard and criteria. This paper shows results of criticality analysis with respect to each designated criterion with modeling of a virtual nuclear fuel assembly and a cask body that induces the maximum reactivity among various design basis fuels of the metal cask. In addition, the sensitivity analysis of nuclear criticality taking into account the various modeling deviation such as manufacturing tolerance and modeling assumptions of conventional models was carried out to ensure the reliability of the analysis result. The criticality evaluation result of the metal cask and the maximum keff under normal and off-normal conditions were 0.36884 and 0.46255, respectively. The maximum keff under a postulated accident condition triggering

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

  19. A conceptual redesign of an inter-building fuel transfer cask

    International Nuclear Information System (INIS)

    The Inter-Building Fuel Transfer Cask, referred to as the IBC, is a lead shielded cask for transporting subassemblies between buildings on the Argonne National Laboratory-West site near Idaho Falls, Idaho. The cask transports both newly fabricated and spent reactor subassemblies between the Experimental Breeder Reactor-2 (EBR-2), the Fuel Cycle Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The IBC will play a key role in the Integral Fast Reactor (IFR) fuel recycling demonstration project. The existing IBC technology, designed and fabricated in the late fifties, is outdated and is a source of personnel exposure at ANL-W. The current IBC system requires forced argon cooling and has extremely limited passive cooling capabilities due to existing design features. A conceptual redesign of the IBC has been performed. The objective of the conceptual design was to increase the passive heat removal capabilities, reduce the personnel radiation exposure and incorporate enhanced safety features into the design. The heat transfer, radiation and thermal-hydraulic properties of the IBC were analytically modeled to determine the principal factors controlling the design. The scoping studies that were performed determined the vital physical characteristics (i.e., size, shielding, pumps, etc.) of the IBC conceptual design. The conceptual design for the IBC allows subassemblies with up to 800 Watts of decay heat to be passively cooled, a significant increase over the existing system. The new design which incorporates better passive cooling mechanisms will prevent inadvertent damage to the subassembly during postulated loss-of-power and loss-of-flow accident scenarios. The new design also decreases the radiation hazard to personnel by having fewer external systems, a better shield plug design, and surfaces that are easier to decontaminate. The control and monitoring system will also be state-of-the-art technology

  20. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising... Spent Fuel Storage Casks'' to include Amendment No. 8 to Certificate of Compliance (CoC) No....

  1. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.230 Procedures for spent fuel storage cask...

  2. 75 FR 33736 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150--AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System... Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the NAC International, Inc. (NAC), MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

  3. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    Energy Technology Data Exchange (ETDEWEB)

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  4. Response of spent fuel transportation casks to explosive loadings

    International Nuclear Information System (INIS)

    Casks for the transportation of spent power reactor fuel can be exposed to explosive loadings from several causes. Exposure can come from an accident involving a propane or other hydrocarbon tanker, from an accident involving military or industrial explosives, or from deliberate sabotage. The regulations for the design of these casks do not specifically include requirements for resistance to blast loadings, but the hypothetical accident sequence that the casks are required to survive assure some measure of blast resistance. To perform accurate risk and security assessments, this blast resistance must be quantified. This paper will discuss the methodology used to determine the blast resistance of a representative rail and a representative truck spent fuel transportation cask. The methodology discussed in this paper can be used to determine the response to explosive loadings other than the one discussed in this paper or to determine the effect of explosive loadings on other casks. Due to the sensitive nature of this topic, this paper is intentionally vague on a number of parameters used in the analyses

  5. Safety Tests of Concrete Storage Cask for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    In preparation for the timely installation of interim storage facility for spent nuclear fuel (SF), KORAD is developing domestic models of SF storage systems and the concrete storage cask is one of them. A concrete cask consists of a metallic canister which confines SF with welded closure and a concrete overpack which provides radiation shielding and physical protection to the canister. The safety requirements for a SF storage cask is well established in US and summarized in regulatory guides such as NUREG-1536. KAERI has been performing tests of the concrete cask to demonstrate its safety and compliance to the regulatory requirements with high priority stipulated in NUREG-1536. The test program includes the structural performance tests under tip-over and earthquake and decay heat removal test under normal, off-normal and accident conditions. In this paper, brief introduction to the structural tests and their results are provided. Safety tests to demonstrate the safety of KORAD21C concrete storage cask were successfully performed. The structural integrity during tip-over and earthquake were demonstrated with scale model tests and the results are analyzed in comparison with safety analysis results

  6. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

    Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1/2 truck cask, which has been relicensed to carry pressurized water reactor (PWR) fuel with 56,000 MWd/ton U burnup at 450 days of cooling time. New cask designs should consider the effects of high burnup for future use, even though it is not commercially desirable to include currently unneeded capability. In conclusion, the increased heat and gamma radiation of high-burnup fuels can be accommodated by additional cooling time, but the increased neutron radiation source cannot be accommodated unless the balance of neutron and gamma contributions to the overall dose rate is properly chosen in the initial cask design. Criticality control of high-burnup fuels is possible with heavily poisoned baskets, but burnup credit in licensing is a much more direct means of demonstrating criticality safety

  7. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

    CANDU spent fuel discharged from the reactor core contains Pu, so, a special attention must be focussed into two directions: tracing for the fuel reactivity in order to prevent critical mass formation and personnel protection during the spent fuel manipulation. Shielding analyses, an essential component of the nuclear safety, take into account the difficulties occurred during the manipulation, transport and storage of spent fuel bundles, both for personnel protection and impact on the environment. The main objective here consists in estimations on radiation doses in order to reduce them under specified limit values. In order to perform the shielding calculations for the spent fuel transport cask three different codes were used: XSDOSE code and MORSE-SGC code, both incorporated in the SCALE4.4a system, and PELSHIE-3 code, respectively. As source of radiation one spent standard CANDU fuel bundle was used. All the geometrical and material data, related to the transport casks, were considered according to the shipping cask type B model, whose prototype has been realized and tested in the Institute for Nuclear Research Pitesti. The radial gamma dose rates estimated to the cask wall and in air, at different distances from the cask, are presented together with a comparison between the dose rates values obtained by all three recipes of shielding calculations. (authors)

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, A. [New York Power Authority, White Plains, NY (United States)

    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.

  11. Vestibule and Cask Preparation Mechanical Handling Calculation

    Energy Technology Data Exchange (ETDEWEB)

    N. Ambre

    2004-05-26

    The scope of this document is to develop the size, operational envelopes, and major requirements of the equipment to be used in the vestibule, cask preparation area, and the crane maintenance area of the Fuel Handling Facility. This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAIC Company L.L.C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC28-01R W12101'' (Ref. 167124). This correspondence was appended by further correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC28-01R W12101; TDL No. 04-024'' (Ref. 16875 1). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process.

  12. Production of casks acceptable for final storage by subsequent treatment of prefilled casks

    International Nuclear Information System (INIS)

    During the operation and the decommissioning of nuclear facilities also radioactive waste material which cannot be encompassed under the general standard waste categories arises. To transfer these types of waste material to interim/final repositories a conditioning/treatment is necessary in most cases. The acceptance conditions of the interim and final repositories require a conditioning considering the type of waste, the specific activities, and the casks to be used. A possible way of conditioning e. g. liquid waste (resins, filter aid, etc.) is to fill the waste into thick-wall casks, if necessary with additional shielding and subsequent drying res. draining. This presentation shall show the experiences and the results gained from the conditioning of these types of middle and higher activated waste. In the NPP Neckar (GKN) 14 ea. 200-I-rolling hoop drums and in the NPP Brokdorf (KBR) 83 ea. mouldings filled with granular resins were stored. 32 200-I-drums with higher activated filters, sludge, as well as mixed waste were located in shielded areas of the drum storage. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Sola, A., E-mail: a.sola@inrim.it; Kuepferling, M.; Basso, V.; Pasquale, M. [Istituto Nazionale di Ricerca Metrologica, Strada delle Cacce 91, 10135 Turin (Italy); Kikkawa, T. [Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan); Uchida, K. [Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan); PRESTO, Japan Science and Technology Agency, Saitama 332-0012 (Japan); Saitoh, E. [Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan); WPI Advanced Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan); CREST, Japan Science and Technology Agency, Tokyo 102-0076 (Japan); Advanced Science Research Center, Japan Atomic Energy Agency, Tokai 319-1195 (Japan)

    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}.

  14. AREVA NP Inc next generation fresh UO2 fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    International Nuclear Information System (INIS)

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries including USA (France), Germany, Belgium, Sweden, China, and South Africa... and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: -Preferential flooding - Fuel rod lattice pitch expansion for full length of fuel assemblies - Neutron absorber penalty -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - BoralTM as neutron absorber - High density polyethylene (HDPE) or Nylon as neutron moderator - Foam as thermal and mechanical protection The cask is designed to handle the complete

  15. AREVA NP next generation fresh UO2 fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    International Nuclear Information System (INIS)

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries including USA, France, Germany, Belgium, Sweden, China, and South Africa - and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up-to-date shipping cask (so called MAP project) gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: - Preferential flooding; - Fuel rod lattice pitch expansion for full length of fuel assemblies; - Neutron absorber penalty; -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - BoralTM as neutron absorber; - High density polyethylene (HDPE) or Nylon as neutron moderator; - Foam as thermal and mechanical protection. The cask is

  16. Standard Practice for Evaluating Solar Absorptive Materials for Thermal Applications

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice covers a testing methodology for evaluating absorptive materials used in flat plate or concentrating collectors, with concentrating ratios not to exceed five, for solar thermal applications. This practice is not intended to be used for the evaluation of absorptive surfaces that are (1) used in direct contact with, or suspended in, a heat-transfer liquid, (that is, trickle collectors, direct absorption fluids, etc.); (2) used in evacuated collectors; or (3) used in collectors without cover plate(s). 1.2 Test methods included in this practice are property measurement tests and aging tests. Property measurement tests provide for the determination of various properties of absorptive materials, for example, absorptance, emittance, and appearance. Aging tests provide for exposure of absorptive materials to environments that may induce changes in the properties of test specimens. Measuring properties before and after an aging test provides a means of determining the effect of the exposure. 1.3 Th...

  17. REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)

    Energy Technology Data Exchange (ETDEWEB)

    CHASTAIN, S.A.

    2005-10-24

    Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The

  18. Standard review plan for dry cask storage systems. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    The Standard Review Plan (SRP) For Dry Cask Storage Systems provides guidance to the Nuclear Regulatory Commission staff in the Spent Fuel Project Office for performing safety reviews of dry cask storage systems. The SRP is intended to ensure the quality and uniformity of the staff reviews, present a basis for the review scope, and clarification of the regulatory requirements. Part 72, Subpart B generally specifies the information needed in a license application for the independent storage of spent nuclear fuel and high level radioactive waste. Regulatory Guide 3.61 {open_quotes}Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask{close_quotes} contains an outline of the specific information required by the staff. The SRP is divided into 14 sections which reflect the standard application format. Regulatory requirements, staff positions, industry codes and standards, acceptance criteria, and other information are discussed.

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

    Science.gov (United States)

    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.

  20. Dry Spent Fuel Cask Transporter equipment design, testing, and operational features

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE) has established a program for the testing of a variety of dry spent fuel storage casks. The program is being conducted at the Idaho National Engineering Laboratory (INEL) by EG and G Idaho Inc. Testing of storage casks at INEL requires that large storage casks (max. gross wt. 127.1 Mg) be moved and positioned from/to an indoor loading location to an outdoor storage pad. A Dry Spent Fuel Cask Transporter has been developed to safely, conveniently, and economically transport/handle a variety of storage casks within and around the confines of nuclear sites and facility

  1. Validation and benchmarking of calculation methods for photon and neutron transport at cask configurations

    International Nuclear Information System (INIS)

    The reliability of calculation tools to evaluate and calculate dose rates appearing behind multi-layered shields is important with regard to the certification of transport and storage casks. Actual benchmark databases like SINBAD do not offer such configurations because they were developed for reactor and accelerator purposes. Due to this, a bench-mark-suite based on own experiments that contain dose rates measured in different distances and levels from a transport and storage cask and on a public benchmark to validate Monte-Carlo-transport-codes has been developed. The analysed and summarised experiments include a 60Co point-source located in a cylindrical cask, a 252Cf line-source shielded by iron and polyethylene (PE) and a bare 252Cf source moderated by PE in a concrete-labyrinth with different inserted shielding materials to quantify neutron streaming effects on measured dose rates. In detail not only MCNPTM (version 5.1.6) but also MAVRIC, included in the SCALE 6.1 package, have been compared for photon and neutron transport. Aiming at low deviations between calculation and measurement requires precise source term specification and exact measurements of the dose rates which have been evaluated carefully including known uncertainties. In MAVRIC different source-descriptions with respect to the group-structure of the nuclear data library are analysed for the calculation of gamma dose rates because the energy lines of 60Co can only be modelled in groups. In total the comparison shows that MCNPTM fits very wall to the measurements within up to two standard deviations and that MAVRIC behaves similarly under the prerequisite that the source-model can be optimized. (author)

  2. Impact Analyses and Tests of Concrete Overpacks of Spent Nuclear Fuel Storage Casks

    International Nuclear Information System (INIS)

    A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches, those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters

  3. Thermoelectric Powered Wireless Sensors for Dry-Cask Storage

    Science.gov (United States)

    Carstens, Thomas Alan

    This study focuses on the development of self-powered wireless sensors. These sensors can be used to measure key parameters in extreme environments; e.g., temperature monitoring for spent nuclear fuel during dry-cask storage. This study has developed a design methodology for these self-powered monitoring systems. The main elements that constitute this work consist of selecting and testing a power source for the wireless sensor, determination of the attenuation of the wireless signal, and testing the wireless sensor circuitry in an extreme environment. OrigenArp determined the decay heat and gamma/neutron source strength of the spent fuel throughout the service life of the dry-cask. A first principles analysis modeled the temperatures inside the dry-cask. A finite-element heat transfer code calculated the temperature distribution of the thermoelectric and heat sink. The temperature distributions determine the power produced by the thermoelectric. It was experimentally verified that a thermoelectric generator (HZ-14) with a DC/DC converter (Linear Technology LTC3108EDE) can power a transceiver (EmbedRF) at condition which represent prototypical conditions throughout and beyond the service life of the dry-cask. The wireless sensor is required to broadcast with enough power to overcome the attenuation from the dry-cask. It will be important to minimize the attenuation of the signal in order to broadcast with a small transmission power. To investigate the signal transmission through the dry-cask, CST Microwave Studio was used to determine the scattering parameter S2,1 for a horizontal dry-cask. Important parameters that can influence the transmission of the signal are antenna orientation, antenna placement, and transmission frequency. The thermoelectric generator, DC/DC converter, and transceiver were exposed to 60Co gamma radiation (exposure rate170.3 Rad/min) at the University of Wisconsin Medical Radiation Research Center. The effects of gamma radiation on the

  4. Quality assurance in a cask fleet parts control system

    International Nuclear Information System (INIS)

    This paper discusses applicable portions of the eighteen Quality Assurance criteria of Subpart H, 10 CFR 71 which are incorporated into a relational data base system which has been designed to manage the spare parts control system for a fleet of spent nuclear fuel casks. The system includes not only parts in warehouse storage but parts in use in the field plus casks, ancillary equipment, test equipment, support devices, and even personnel. It provides a high degree of assurance that any device for which a condition for certification has expired will be flagged for recertification testing or removal from service well before the critical date

  5. IMPACLIB: a material property data library for impact analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-12-01

    The paper describes the structural data library and graphical program for impact and stress analyses of radioactive material transport casks. Four kinds of material data, structure steels, stainless steels, leads and woods are compiled. These materials are main structural elements of casks. Structural data such as, coefficient of thermal expansion, modulus of longitudinal elasticity, modulus of transverse elasticity, Poisson`s ratio and stress-strain relationship have been tabulated. Main features of IMPACLIB are as follows: (1) data have been tabulated against temperature or strain rate, (2) thirteen kinds of polynominal fitting for stress-strain curve are available, (3) it is capable of graphical representations for structural data and (4) the IMPACLIB is able to be used on not only main frame computers but also work stations (OS UNIX) and personal computers (OS Windows 3.1). In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides a user`s guide for computer program and input data for the IMPACLIB. (author)

  6. IMPACLIB: a material property data library for impact analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    The paper describes the structural data library and graphical program for impact and stress analyses of radioactive material transport casks. Four kinds of material data, structure steels, stainless steels, leads and woods are compiled. These materials are main structural elements of casks. Structural data such as, coefficient of thermal expansion, modulus of longitudinal elasticity, modulus of transverse elasticity, Poisson's ratio and stress-strain relationship have been tabulated. Main features of IMPACLIB are as follows: (1) data have been tabulated against temperature or strain rate, (2) thirteen kinds of polynominal fitting for stress-strain curve are available, (3) it is capable of graphical representations for structural data and (4) the IMPACLIB is able to be used on not only main frame computers but also work stations (OS UNIX) and personal computers (OS Windows 3.1). In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides a user's guide for computer program and input data for the IMPACLIB. (author)

  7. Spent fuel storage cask testing and operational experience at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Spent-fuel storage cask research, development, and demonstration activities are being performed for the U.S. Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) as a part of the storage cask testing program. The cask testing program at federal sites and other locations supports the Nuclear Waste Policy Act (NWPA) and DOE objectives for cooperative demonstrations with the cask vendors and utilities for development of at-reactor dry cask storage capabilities for spent nuclear fuel assemblies. One research and development program for the storage cask performance testing of metal storage cask was initiated through a cooperative agreement between Virginia Power and DOE in 1984. The performance testing was conducted for the DOE and the Electric Power Research Institute by the Pacific Northwest laboratory, operated for DOE by Battelle Memorial Institute, and the Idaho National Engineering Laboratory (INEL), operated for DOE by EG ampersand G Idaho, Inc. In 1988 a cooperative agreement was entered into by DOE with Pacific Sierra Nuclear Associates (PSN) for performance testing of the PSN concrete Ventilated Storage Cask. Another closely related activity involving INEL is a transportable storage cask project identified as the Nuclear Fuel Services Spent-Fuel Shipping/Storage Cask Demonstration Project. The purpose of this project is to demonstrate the feasibility of packing, transporting, and storing commercial spent fuel in dual-purpose transport/storage casks

  8. The interim storage facility with dry storage casks and its safeguards activity

    International Nuclear Information System (INIS)

    Recyclable-Fuel Storage Company (RFS) is constructing an interim storage facility of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel from nuclear power plants and to serve for about 50 years in RFSC. Metallic dry casks have already been used for dry cask storage facility at Tokai No.2 power station of Japan Atomic Power Company. But, RFSC is not exactly the same as the dry cask storage facility at Tokai No.2 power station, for example, cask transportation between facilities and no hot cells. Therefore, additional safeguards activities are necessary. The outline of the design and handling of metallic dry casks at RFSC and the currently developing status of safeguards activity such as containment and surveillance for the cask receipt and storage at RFSC, etc are described. (author)

  9. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    CERN Document Server

    Poulson, D; Guardincerri, E; Morris, C L; Bacon, J D; Plaud-Ramos, K; Morley, D; Hecht, A

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, the cask contents can be confirmed with high confidence in less than two days exposure. Similar results can be obtained by moving a smaller detector to view the cask from multiple angles.

  10. Thermal evaluation of vertical greenery systems for building walls

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Nyuk Hien; Kwang Tan, Alex Yong; Chen, Yu. [Department of Building, School of Design and Environment, National University of Singapore, 4 Architecture Drive, Singapore 117566 (Singapore); Sekar, Kannagi [Department of Building, School of Design and Environment, National University of Singapore, 4 Architecture Drive, Singapore 117566 (Singapore); National Parks Board, Singapore Botanic Garden, 1 Cluny Road, Singapore 259569 (Singapore); Tan, Puay Yok; Chan, Derek; Chiang, Kelly [National Parks Board, Singapore Botanic Garden, 1 Cluny Road, Singapore 259569 (Singapore); Wong, Ngian Chung [Building and Construction Authority, 5 Maxwell Road, Singapore 069110 (Singapore)

    2010-03-15

    This research involves the study of 8 different vertical greenery systems (VGSs) installed in HortPark to evaluate the thermal impacts on the performance of buildings and their immediate environment based on the surface and ambient temperatures. VGSs 3 and 4 have the best cooling efficiency according to the maximum temperature reduction of the wall and substrate surfaces. These results point to the potential thermal benefits of vertical greenery systems in reducing the surface temperature of buildings facades in the tropical climate, leading to a reduction in the cooling load and energy cost. In terms of the lowest diurnal range of average wall surface temperature fluctuation, VGSs 4 and 1 show the highest capacities. No vertical greenery system performs well in term of the diurnal range of average substrate temperature fluctuation. By limiting the diurnal fluctuation of wall surface temperatures, the lifespan of building facades is prolonged, slowing down wear and tear as well as savings in maintenance cost and the replacement of facade parts. The effects of vertical greenery systems on ambient temperature are found to depend on specific vertical greenery systems. VGS 2 has hardly any effect on the ambient temperature while the effects of VGS 4 are felt as far as 0.60 m away. Given the preponderance of wall facades in the built environment, the use of vertical greenery systems to cool the ambient temperature in building canyons is promising. Furthermore, air intakes of air-conditioning at a cooler ambient temperature translate into saving in energy cooling load. (author)

  11. Logistics management for storing multiple cask plug and remote handling systems in ITER

    International Nuclear Information System (INIS)

    Highlights: ► We model the logistics management problem in ITER, taking into account casks of multiple typologies. ► We propose a method to determine the best position of the casks inside a given storage area. ► Our method obtains the sequence of operations required to retrieve or store an arbitrary cask, given its storage place. ► We illustrate our method with simulation results in an example scenario. -- Abstract: During operation, maintenance inside the reactor building at ITER (International Thermonuclear Experimental Reactor) has to be performed by remote handling, due to the presence of activated materials. Maintenance operations involve the transportation and storage of large, heavyweight casks from and to the tokamak building. The transportation is carried out by autonomous vehicles that lift and move beneath these casks. The storage of these casks face several challenges, since (1) the cask storage area is limited in space, and (2) all casks have to be accessible for transportation by the vehicles. In particular, casks in the storage area may block other casks, so that the former has to be moved to a temporary position to give way to the latter. This paper addresses the challenge of managing the logistics of cask storage, where casks may have different typologies. In particular, we propose an approach to (1) determine the best position of the casks inside the storage area, and to (2) obtain the sequence of operations required to retrieve and store an arbitrary cask from/to a given storage place. A combinatorial optimization approach is used to obtain solutions to both these problems. Simulation results illustrate the application of the proposed method to a simple scenario

  12. Evaluation of antioxidants stability by thermal analysis and its protective effect in heated edible vegetable oil

    OpenAIRE

    Seme Youssef Reda

    2011-01-01

    In this work, through the use of thermal analysis techniques, the thermal stabilities of some antioxidants were investigated, in order to evaluate their resistance to thermal oxidation in oils, by heating canola vegetable oil, and to suggest that antioxidants would be more appropriate to increase the resistance of vegetable oils in the thermal degradation process in frying. The techniques used were: Thermal Gravimetric (TG) and Differential Scanning Calorimetry (DSC) analyses, as well as an a...

  13. Evaluation of thermal resistance of building insulations with reflective surfaces

    Science.gov (United States)

    Št'astník, S.

    2012-09-01

    The thermal resistance of advanced insulation materials, applied namely in civil engineering, containing reflective surfaces and air gaps, cannot be evaluated correctly using the valid European standards because of presence of the dominant nonlinear radiative heat transfer and other phenomena not included in the recommended computational formulae. The proper general physical analysis refers to rather complicated problems from classical thermodynamics, whose both existence theory and numerical analysis contain open questions and cannot be done in practice when the optimization of composition of insulation layers is required. This paper, coming from original experimental results, demonstrates an alternative simplified computational approach, taking into account the most important physical processes, useful in the design of modern insulation systems.

  14. Studies and research concerning BNFP: advanced cask handling studies

    International Nuclear Information System (INIS)

    Cask turnaround times at loading and unloading sites can be improved by providing better working conditions, improved safety, reduced decontamination time, training, and where practical to do so, improved facility design. This report consists of treatments of several of these topics with the common goal of improving operational efficiency

  15. Separator assembly for use in spent nuclear fuel shipping cask

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  16. Implementation of response function concept for spent fuel cask analyses

    International Nuclear Information System (INIS)

    Due to the uncertain schedule about the first disposal of the large quantity of spent nuclear fuel (SNF) accumulated at the US commercial nuclear power plants, and due to the wide range of burnups and cooling times of the SNF, it is urgent to develop a quick and realistic method for analyzing an interim-storage or shipping package of SNF. The existing method uses design-basis SNF, and it is unnecessarily conservative and therefore uneconomic. This paper demonstrates the use of response-function concept for shielding and criticality analysis for a commercial SNF shipping cask. A PC-based computer code is written for this purpose. The program allows users to perform accurate shielding and criticality analyses for any selected cask payload on real-time basis. The results are less conservative, but more realistic than that of the design-basis analyses. One must be noted, however, that the response function is cask-specific. Therefore, the concept is most beneficial to the major cask type which is to be repeatedly used for a large number of SNF shipments

  17. Monitored Retrievable Storage conceptual system study: cask-in-trench

    International Nuclear Information System (INIS)

    This report provides a description of the Cask-in-Trench Storage Concept which meets a specified set of requirements; an estimate of the costs of construction, operation and decommissioning of the concept; the costs required to expand the facility throughput and storage capability; and the life cycle costs of the facility. 22 figures, 34 tables

  18. Dry Storage Casks Monitoring by Means of Ultrasonic Tomography

    Science.gov (United States)

    Salchak, Y.; Bulavinov, A.; Pinchuk, R.; Lider, A.; Bolotina, I.; Sednev, D.

    Spent nuclear fuel (SNF) is one of the most hazardous types of nuclear power plant waste. This fact emphasizes the importance of careful handling and storage of SNF. There are two current state-of-the art technologies of SNF storage facility: wet and dry. It is important to mention that IAEA does not determine which kind of handling strategy should be chosen, however it is noted that dry storage of SNF could be used for one hundred years. Mining and Chemical Enterprise (MCE) is one of the leading Russian companies that deals exclusively with the dry storage of SNF. This company has implemented a long-term storage scheme. At the same time MCE faced the challenge of nondestructive monitoring of the degradation process of structural material of cask and its sealing with weld seam. Currently, X-ray testing is used for this purpose but in order to provide an effective nonradioactive method of monitoring MCE has initiated a collaborative R&D project with TPU supported by the Russian Government. Ultrasonic industrial tomography technique was proposed as the solution. The method is based on application of phased and sparse arrays transducer with real-time visualization algorithm. Received acoustic data is processed and realized by means of Sampling Phased Array technology which is a collaborative development of TPU and I-Deal Technology, GmbH. The multichannel ultrasonic set-up of immersion control was assembled for performing testing of seven experimental specimens with representative defects (side drill-holes, notches, natural welding flaws). X-ray tomography of high-resolution was chosen as the reference method. All indications were successfully reconstructed in B and C-scans and 3D image. The next step is to automate the monitoring procedure completely and to introduce an evaluation tool for current flaw state and prediction of its further behavior.

  19. Design of casks: incorporating operational feedback from maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Bimet, F.; Hartenstein, M. [COGEMA Logistics, Saint Quentin (France)

    2004-07-01

    Casks are designed to conform to regulations and to client specifications. Essential areas such as easy operation, low costs of maintenance, low operation and maintenance doses, limited waste, are not explicitly covered. Notwithstanding, COGEMA LOGISTICS uses all feedback available, so that casks are designed to be easy, safe and economical to operate and maintain. Maintenance is an activity where you do spot items that old-time designers could have made better, and things that users should not have done. Thanks to quality assurance, there are a number of data available, waiting to be collected and exploited; they have to be identified, located, retrieved, and analysed. Using information such as wear, damage, use of spare parts, access problems helps to make casks ever better. It leads to more efficient concepts, and to upgrades on existing designs; it also allows to integrate environmental considerations, inter alia in the choice of materials and in maintenance methods. It is necessary for the designer to interact with the users, the cask owners, the maintenance providers, in order to gather all relevant information and events. This is made easier when all these actors are ''under one roof'', or have very close ties. This paper presents COGEMA LOGISTICS methods for collecting and analysing all these experiences waiting to be used. It explains our process for analysing data, preparing yearly reports that are made available to our designers. It describes how each new design is subject to a maintainability study, using this feedback, so that the cask safety is always assured, that radiological doses are kept to a minimum, and that operating and maintenance costs will remain as low as possible.

  20. Design of casks: incorporating operational feedback from maintenance

    International Nuclear Information System (INIS)

    Casks are designed to conform to regulations and to client specifications. Essential areas such as easy operation, low costs of maintenance, low operation and maintenance doses, limited waste, are not explicitly covered. Notwithstanding, COGEMA LOGISTICS uses all feedback available, so that casks are designed to be easy, safe and economical to operate and maintain. Maintenance is an activity where you do spot items that old-time designers could have made better, and things that users should not have done. Thanks to quality assurance, there are a number of data available, waiting to be collected and exploited; they have to be identified, located, retrieved, and analysed. Using information such as wear, damage, use of spare parts, access problems helps to make casks ever better. It leads to more efficient concepts, and to upgrades on existing designs; it also allows to integrate environmental considerations, inter alia in the choice of materials and in maintenance methods. It is necessary for the designer to interact with the users, the cask owners, the maintenance providers, in order to gather all relevant information and events. This is made easier when all these actors are ''under one roof'', or have very close ties. This paper presents COGEMA LOGISTICS methods for collecting and analysing all these experiences waiting to be used. It explains our process for analysing data, preparing yearly reports that are made available to our designers. It describes how each new design is subject to a maintainability study, using this feedback, so that the cask safety is always assured, that radiological doses are kept to a minimum, and that operating and maintenance costs will remain as low as possible

  1. Peanut Seed Vigor Evaluation Using a Thermal Gradient

    Directory of Open Access Journals (Sweden)

    Timothy L. Grey

    2011-01-01

    Full Text Available Experiments conducted from 2007 to 2009 evaluated germination of 11 peanut runner-type cultivars. Germination was evaluated in Petridishes incubated over a thermal gradient ranging from 14 to 30°C at 1.0 C increments. Beginning 24 hr after seeding, peanut was counted as germinated when radicles were greater than 5 mm long, with removal each day. Germination was counted daily for seven days after seeding. Growing-degree day (GDD accumulation for each temperature increment was calculated based on daily mean temperature for that Petri dish. Two indices were obtained from a logistic growth curve used to elucidate seed germination by cultivar: (1 maximum indices of germination and (2 GDD value at 80% germination (Germ80, an indication of seed vigor the lower the Germ80 value, the greater the seed lot vigor. Based on the two indices, seed lots “AT 3081R”, “AP-3”, “GA-06G”, and “Carver” had the strongest seed vigor (Germ80 26 to 47 GDD and a high maximum incidence of germination rate (80 to 94%. Seed lots of “C99-R”, “Georgia-01R”, “Georgia-02C”, and “Georgia-03L” had inconsistent seed performance, failing to achieve 80% germination in at least two of three years.

  2. Opportunities to increase the productivity of spent fuel shipping casks in the United States

    International Nuclear Information System (INIS)

    Trends indicate that future transportation requirements for spent fuel will be different from those anticipated when the current generation of casks and vehicles was designed. Increased storage capacity at most reactors will increase the average post irradiation age of the spent fuel to be transported. A scenario is presented which shows the 18 casks currently available should be sufficient until approximately 1983. Beyond this time, it appears that an adequate transportation system can be maintained by acquiring, as needed, casks of current designs and new casks currently under development. Spent fuel transportation requirements in the post-1990 period can be met by a new generation of casks specifically designed to transport long-cooled fuel. In terms of the number of casks needed, productivity may be increased by 19% if rail cask turnaround time is reduced to 4 days from the current range of 6.5 to 8.5 days. Productivity defined as payloads per cask year could be increased 62% if the turnaround time for legal weight truck casks were reduced from 12 hours to 4 hours. On a similar basis, overweight truck casks show a 28% increase in productivity

  3. Scoping study of casks shipped from the MRS facility to various repository sites

    International Nuclear Information System (INIS)

    The objective of this study was to determine the maximum number of specialized repository waste packages that could be shipped from the Monitored Retrievable Storage (MRS) facility in Pb-, Fe-, and U-shielded casks weighing 200,000 or 300,000 lbs. The study included 18 different waste packages designed for the Salt, Tuff, and Basalt repositories. Nine of these contained consolidated PWR fuel pins, and nine contained consolidated BWR fuel pins. Discrete ordinates calculations were performed to determine the neutron and gamma shield thicknesses that would ensure a dose rate of 10 millirem/hr, 10 ft from the centerline of the cask(s). Over 100 casks of particular interest have been identified, while preliminary design information is also given for 522 casks of potential interest. Relative to the 200,000-lb casks, 50 to 100% more fuel may be shipped in the larger 300,000-lb casks. Placing the spent fuel canisters in overpacks prior to shipment from the MRS will reduce the net payload by 30 to 50%. The highest-capacity cask/waste package combination studied corresponds to a 300,000-lb U-shielded cask containing 84 consolidated PWR fuel assemblies in 21 canisters, or 171 consolidated BWR fuel assemblies in 19 canisters. Criticality analyses have shown these high-capacity casks to be safely subcritical - even if all the canisters were loaded with unirradiated LWR fuel containing 3.4 wt % U-235

  4. Evaluation of Erosion Resistance of Advanced Turbine Thermal Barrier Coatings

    Science.gov (United States)

    Zhu, Dongming; Kuczmarski, Maria A.; Miller, Robert A.; Cuy, Michael D.

    2007-01-01

    The erosion resistant turbine thermal barrier coating system is critical to aircraft engine performance and durability. By demonstrating advanced turbine material testing capabilities, we will be able to facilitate the critical turbine coating and subcomponent development and help establish advanced erosion-resistant turbine airfoil thermal barrier coatings design tools. The objective of this work is to determine erosion resistance of advanced thermal barrier coating systems under simulated engine erosion and/or thermal gradient environments, validating advanced turbine airfoil thermal barrier coating systems based on nano-tetragonal phase toughening design approaches.

  5. Thermal spray deposition and evaluation of low-Z coatings

    Energy Technology Data Exchange (ETDEWEB)

    Seals, R.D.; Swindeman, C.J.; White, R.L.

    1996-09-01

    Thermally sprayed low-Z coatings of B{sub 4}C on Al substrates were investigated as candidate materials for first-wall reactor protective surfaces. Comparisons were made to thermally sprayed coatings of B, MgAl{sub 2}O{sub 4}, Al{sub 2}O{sub 3}, and composites. Graded bond layers were applied to mitigate coefficient of thermal expansion mismatch. Microstructures, thermal diffusivity before and after thermal shock loading, steel ball impact resistance, CO{sub 2} pellet cleaning and erosion tolerance, phase content, stoichiometry by Rutherford backscattering spectroscopy, and relative tensile strengths were measured.

  6. Operational assessment of the transnuclear TN-9 truck spent fuel shipping cask: studies and research concerning BNFP

    International Nuclear Information System (INIS)

    This report presents the results of an operational assessment of the Transnuclear Inc., TN-9 spent fuel cask. This packaging system transports seven current generation boiling-water-reactor nuclear fuel assemblies in a truck shipping mode. The studies were performed at the Barnwell Nuclear Fuel Plant by employees of Allied-General Nuclear Services. The work was funded by the Department of Energy during fiscal year 1981. The cooperation of Transnuclear in this effort is gratefully acknowledged. The study is based on repeated simulated unloading runs of TN-9. Specific tasks and areas of study included: (1) sequential dry-run handling operations under simulated unloading conditions, (2) detailed time and manpower studies, (3) estimates of operator radiation exposure, (4) a general evaluation of the cask system capabilities as they relate to unloading and loading facility operations, and (5) preparation of operating procedures for both unloading (confirmed by practice runs) and loading (yet to be confirmed). Also included is general information on the cask, auxiliary equipment, and the Certificate of Compliance

  7. Transportation package thermal and shielding response to a regulatory fire

    International Nuclear Information System (INIS)

    The objective of this work is to evaluate the effect of neutron shield charring due to a regulatory fire on the thermal response and shielding effectiveness of a Multi-Purpose Canister (MPC) and transportation cask. A thermal response model which includes the effect of neutron shield charring is developed. The model is solved using a time dependent finite element code. The maximum fuel temperature-time history and the extent of shield charring are determined. This is used to estimate the primary dose rate from the package in the post-fire condition. It is determined that charring has an insignificant effect on the thermal response of the fuel. Furthermore, while charring increases dose rates, these rates remain below Nuclear Regulatory Commission limits for accident conditions

  8. Dynamic evaluation of thermal comfort environment of air-conditioned buildings

    Energy Technology Data Exchange (ETDEWEB)

    Hanqing, Wang; Zhiyong, Wang [Department of Civil Engineering, Zhuzhou Institute of Technology, Zhuzhou Hunan 412008 (China); Chunhua, Huang; Yingyun, Liu [Nanhua University, Hengyang 421001 (China); Zhiqiang, Liu [School of Energy and Power Engineering, Central South University, Changsha 410083 (China); Guangfa, Tang [School of Civil Engineering, Hunan University, Changsha 410082 (China)

    2006-11-15

    In this paper, based upon Fanger's thermal comfort concept, several concepts, which utilize computing results obtained from the large eddy simulation (LES), are put forward, such as thermal comfort index based on time-averaged parameters, instantaneous thermal index, time-averaged thermal comfort index and time-averaged thermal comfort index along walking routes. Also their discrepancies and calculation methods are discussed in the paper. Apart from these, we have calculated PD value as an example, whose results indicate that the distributions of four indices are obviously different. Therefore, it is suggested to distinguish different cases and select correspondingly thermal comfort evaluation indices.

  9. Thermal Grease Evaluation for ATLAS Upgrade Micro-Strip Detector.

    CERN Document Server

    Barbier, G; The ATLAS collaboration; Clark, A; Ferrère, D; Pernecker, S; Perrin, E; Streit, KP; Weber, M

    2010-01-01

    The ATLAS upgrade detector foreseen at the phase 2 upgrade of LHC requires a complete new inner detector using silicon pixel and strip detectors. For both technologies, a specific mechanical and thermal design is required. Such a design may use soft thermal interfaces such as grease between the various parts. One foreseeable use would be between the cooling pipe and the thermal block allowing the strip modules to be decoupled from the mechanical and cooling structure. This note describes the technique used and the results obtained when characterizing a few grease samples. The results have been compared with thermal FEA simulations. A thermal conductivity measurement for each sample could be extracted from the measurements, with a systematic uncertainty of less than 6%. Some samples were irradiated to the expected fluence at sLHC and their resulting thermal conductivity compared with the non-irradiated samples.

  10. Evaluation of Candidate In-Pile Thermal Conductivity Techniques

    Energy Technology Data Exchange (ETDEWEB)

    B. Fox; H. Ban; J. Daw; K. Condie; D. Knudson; J. Rempe

    2009-05-01

    Thermophysical properties of materials must be known for proper design, test, and application of new fuels and structural properties in nuclear reactors. In the case of nuclear fuels during irradiation, the physical structure and chemical composition change as a function of time and position within the rod. Typically, thermal conductivity changes, as well as other thermophysical properties being evaluated during irradiation in a materials and test reactor, are measured out-of-pile in “hot-cells.” Repeatedly removing samples from a test reactor to make out-of-pile measurements is expensive, has the potential to disturb phenomena of interest, and only provide understanding of the sample's end state at the time each measurement is made. There are also limited thermophysical property data for advanced fuels. Such data are needed for the development of next generation reactors and advanced fuels for existing nuclear plants. Having the capacity to effectively and quickly characterize fuels and material properties during irradiation has the potential to improve the fidelity of nuclear fuel data and reduce irradiation testing costs.

  11. NONDESTRUCTIVE EVALUATION OF WOOD STRENGTH USING THERMAL CONDUCTIVITY

    Directory of Open Access Journals (Sweden)

    Türker Dündar,

    2012-06-01

    Full Text Available Relationships between the coefficient of thermal conductivity (CTC and the strength properties of wood were investigated. Small clear test specimens were prepared from beech, fir, and pine wood. CTC values of the test specimens were measured based on the ASTM C 1113-99 hot-wire method. Wood density and some mechanical properties were then determined according to related ISO standards. In order to designate relationships between the CTC and mechanical properties, linear regression analysis was performed. Significant linear correlations were found between the CTC and the specific gravity, the modulus of rupture, the modulus of elasticity, and the impact bending strength of the wood from all tree species. However, there was a weak and non-significant relationship between the CTC and the compression strength of the specimens from each tree species. As a consequence, the CTC has a considerable potential in nondestructive evaluation of wood density and strength. However, the reciprocal correlations among the MC-strength, MC-CTC, temperature-strength, and temperature-CTC appear to be most significant limitations for using CTC as a NDE method for wood. Further detailed investigations are needed.

  12. Thermal analysis of spent nuclear fuel shipping cas

    International Nuclear Information System (INIS)

    In this study, a computational fluid dynamics (CFD) thermal analysis was performed for the TN-24P cask. For the analysis, ANSYS Fluent as a CFD tool was selected since it has the proper finite volume methods to realistically simulate the thermal behavior of shipping casks. For the analysis, spent fuels discharged from pressurized water reactors (PWRs) were modeled. In the model, there are 24 PWR spent fuel assemblies loaded in the TN-24P cask. The fuel design is assumed to be similar to standard Westinghouse 15x15 rod design. Total heat (decay) generated in the cask was estimated to be 20.6 kW. To input the axial power profile required to calculate the heat flux, a User Defined Function was generated. Fuel storage space (canister) is filled with Helium gas to cool spent nuclear fuel. In the cask, heat transfer occurs through the heat conduction by helium and basket, natural circulation driven by gravity, and thermal radiation in the complex geometry. In the canister region, laminar flow model with Boussinesq approximation is used to simulate the natural circulation. The helium domain was assumed symmetric in the model. For thermal radiation, the Discrete Ordinates (DO) model was chosen in the presented study due to its accuracy and capability of parallel processing. In typical vertical TN-24P dry storage cask system consist of two nested cask. Between inner and outer cask is in the air. Air inlet section is at the bottom side of cask and outlet ventilation is at top of cask. At this region, turbulence regime occurs and turbulence is modeled by using k-epsilon model. The analysis include small scaled and full scaled model. In small scale model, geometry is defined rectangular to make mesh generation easy and to validate the analysis tools using the experimental data. In the full-scale simulation, the results of analysis and experimental data for peak clad temperature (PCT) were compared. Key Words: TN-24P dry storage cask, CFD, thermal analysis, PCT, air blockage

  13. Treatment of stainless steel cladding in pressurized thermal shock evaluation: deterministic analyses

    International Nuclear Information System (INIS)

    Fracture mechanics is one of the major areas of the pressurized thermal shock (PTS) evaluation. To evaluate the reactor pressure vessel integrity associated with PTS, PFM methodology demands precise calculation of temperature, stress, and stress intensity factor for the variety of PTS transients. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property, at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, treatment schemes to evaluate stress and resulting stress intensity factor for RPV with stainless steel clad are introduced. For a reference transient, the effects of clad thermal conductivity and thermal expansion coefficients on deterministic fracture mechanics analysis are examined

  14. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  15. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP)

  16. Size and transportation capabilities of the existing U.S. cask fleet

    International Nuclear Information System (INIS)

    This paper investigates the current spent nuclear fuel cask fleet capability in the United States. 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

  17. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    Science.gov (United States)

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

  18. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    OpenAIRE

    Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A.

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first applicati...

  19. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    Energy Technology Data Exchange (ETDEWEB)

    Muhamad, Shalina Sheik [Prototype and Plant Development Center, Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia); Hamzah, Mohd Arif Arif B. [Prototype and Plant Development Center, Technical Support Division Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia)

    2014-02-12

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP)

  20. Castor transport and storage casks for VVER and RBMK fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Gartz, R.; Gobler, A.; John, R.; Diersch, R. [GNB Gesellschaft fur Nuklear-Behalter mbH, Essen (Germany); Nemec, P. [Skoda Nuclear Machinery Plzen (Czech Republic)

    1998-12-31

    CASTOR casks have been successfully developed, manufactured and delivered for Russian type reactor fuel assemblies. These casks fulfill both the requirements for type B packages according to IAEA regulations and the requirements covering different accident situations to be assumed at the storage site. In the following, the CASTOR casks CASTOR 440/84, CASTOR RBMK and CASTOR VVER 1000 are described, the nuclear content is characterized and an overview about the status of licensing, manufacturing and delivery is given. (authors) 3 refs.

  1. Comparison of temperature measurement methods for evaluation of the thermal environment in vehicles

    DEFF Research Database (Denmark)

    Rosendahl, J..; Olesen, Bjarne W.

    2006-01-01

    A new standard, ISO/DIS 14505 is under preparation, dealing with the assessment of the thermal environment in vehicles, based on the equivalent temperature. The scope of this paper is to demonstrate the different results obtained when evaluating the thermal environment inside a vehicle, using the...... during each test. Based on the measurement results, the time used to reach the level of thermal comfort is evaluated and compared....

  2. Pattern Recognition of Thermal Analysis Cooling Curves and Quality Evaluation of Melt Cast Alloys

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The development of thermal analysis techniques for evaluation of cast alloy melt quality and its current applications in the foundry were reviewed. The characteristics of the current thermal analysis techniques were analyzed. A new comprehensive method for cooling curve recognition has been proposed. The evaluation of cast alloy melt quality was realized.

  3. The NINO [No Inspector, No Operator system] cask-loading safeguards system

    International Nuclear Information System (INIS)

    It is, in general difficult to determine by means of camera-surveillance techniques what is loaded into spent-fuel casks being prepared for shipment from light-water reactors to other reactors, reprocessing facilities, or long-term storage. Furthermore, the expected high frequency of cask loadings in the coming years would place too great a burden on the IAEA and Euratom inspectorates if each had to be observed by an inspector. For the case of shipment to other reactors and reprocessing facilities, the casks are soon opened and, in principle, their contents could be ascertained by direct inspection. In the case of long-term-storage facilities, the casks would stay sealed for years, thereby requiring the IAEA to know positively how many spent-fuel assemblies were loaded at the reactor and to have a continuity of knowledge of the cask's contents. It has been proposed instead that the facility operator place the cask seal on the cask within the field of view of a surveillance system linked to the cask seal. This solution, however, may not provide enough credibility for acceptance by the safeguards community. This paper presents an alternative to both inspector presence at cask loading and operator assistance in applying seals; this alternative is called the No Inspector, No Operator system (NINO)

  4. Experimental studies of free-standing spent fuel storage cask subjected to strong earthquakes

    International Nuclear Information System (INIS)

    Concrete cask spent fuel storage system is considered to essentially have an economical advantage and becoming widely used. For vertically free-standing concrete cask on the floor pad in the cask storage facility, its tipping-over and sliding behavior during earthquake is one of the technical key issues to guarantee its safe performance. In this paper, the experimental studies are reported by performing the excitation test with a scale model concrete cask using two-dimensional shaking table and the applicability of the energy spectrum approach is discussed. (author)

  5. Beneficial uses shipping system (BUSS) cask, safety analysis report for packaging: Volumes 1 and 2

    International Nuclear Information System (INIS)

    The Beneficial Uses Shipping System (BUSS) cask Safety Analysis Report for Packaging (SARP) was originally prepared by Sandia National Laboratory (SNL). After the certification process was completed, the ownership of the BUSS cask and associated SARP was transferred from SNL to the DOE Hanford site in Richland, Washington. During timely renewal of the BUSS cask certificate of compliance, the SARP was revised to (1) respond to the timely renewal questions, (2) consolidate the previous revision made by SNL, and (3) bring the SARP into compliance with the 1996 version of 10 CFR 71. Since the BUSS cask is now the responsibility of RL, the SARP was reissued as a Hanford document

  6. Fuel Element Transfer Cask Modelling Using MCNP Technique

    Science.gov (United States)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  7. Stress analysis of closure bolts for shipping casks

    International Nuclear Information System (INIS)

    This report specifies the requirements and criteria for stress analysis of closure bolts for shipping casks containing nuclear spent fuels or high level radioactive materials. The specification is based on existing information conceming the structural behavior, analysis, and design of bolted joints. The approach taken was to extend the ASME Boiler and Pressure Vessel Code requirements and criteria for bolting analysis of nuclear piping and pressure vessels to include the appropriate design and load characteristics of the shipping cask. The characteristics considered are large, flat, closure lids with metal-to-metal contact within the bolted joint; significant temperature and impact loads; and possible prying and bending effects. Specific formulas and procedures developed apply to the bolt stress analysis of a circular, flat, bolted closure. The report also includes critical load cases and desirable design practices for the bolted closure, an in-depth review of the structural behavior of bolted joints, and a comprehensive bibliography of current information on bolted joints

  8. Evaluation of the of thermal shock resistance of a castable containing andalusite aggregates by thermal shock cycles

    International Nuclear Information System (INIS)

    The thermal shock resistance of refractory materials is one of the most important characteristics that determine their performance in many applications, since abrupt and drastic differences in temperature can damage them. Resistance to thermal shock damage can be evaluated based on thermal cycles, i.e., successive heating and cooling cycles followed by an analysis of the drop in Young's modulus occurring in each cycle. The aim of this study was to evaluate the resistance to thermal shock damage in a commercial refractory concrete with andalusite aggregate. Concrete samples that were sintered at 1000 deg C and 1450 deg C for 5 hours to predict and were subjected to 30 thermal shock cycles, soaking in the furnace for 20 minutes at a temperature of 1000 deg C, and subsequent cooling in circulating water at 25 deg C. The results showed a decrease in Young's modulus and rupture around 72% for samples sintered at 1000 ° C, and 82% in sintered at 1450 ° C. The refractory sintered at 1450 deg C would show lower thermal shock resistance than the refractory sintered at 1000 deg C. (author)

  9. Certification of a spent fuel cask for storage and transportation

    International Nuclear Information System (INIS)

    This paper addresses the US Nuclear Regulatory Commission's requirements for the dry storage and transportation of spent fuel, focusing on how the performance standards differ between storage and transportation. The paper also discusses the NRC cask review process, and some current issues in each area of certification. In addition, some of the issues associated with the US Department of Energy's proposed multi-purpose canister are discussed

  10. Evaluation of thermal sprayed coating using ultrasonic inspection by means of bottom echo back reflection

    Institute of Scientific and Technical Information of China (English)

    Toshifumi KUBOHORI; Toru ITO; Wahidullah WAHI; Yasuyuki INUI; Toshiro IKUTA

    2009-01-01

    Thermal spraying technique is widely used in various mechanical parts as a surface reforming technique. However, as demand to maintain superior mechanical performance in harsh operating environment increases, the need for non-destructive evaluation method for thermal spray coating becomes more important. For this purpose, we thinned the thickness of the thermal sprayed coating by abrasion with blasting and used ultrasonic inspection by means of bottom echo reflection for effective measurement of abrasion quantity in thermal sprayed coating. The results obtained are summarized as follows. When the thickness of thermal sprayed coating becomes thin, the echo height increases. This is because thermal sprayed coatings absorb ultrasonic energy. Ultrasonic energy absorbed by Al2O3 is smaller compared with Fe-13Cr coating. Thermal sprayed coatings submerged in water have a lower echo height compared with air. As mentioned above, the thermal sprayed coating thickness can be estimated using ultrasonic inspection by means of bottom echo back reflection.

  11. Evaluation of Defects of Thermal Barrier Coatings by Thermal Shock Test Using Eddy Current Testing

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Tae Hoon; Cho, Youn Ho; Lee, Joon Hyun [Pusan National University, Busan (Korea, Republic of); Oh, Jeong Seok; Lee, Koo Hyun [KIMM, Daejeon (Korea, Republic of)

    2009-10-15

    Periodical thermal shock can introduce defects in thermal barrier coating made by layers of CoNiCrAlY bond coating(BC) and ZrO{sub 2}-8wt%Y{sub 2}O{sub 3} ceramic top coating(TC) on Inconel-738 substrate using plasma spraying. Thermal shock test is performed by severe condition that is to heat until 1000 .deg. C and cool until 20 .deg. C. As the number of cycle is increased, the fatigue by thermal shock is also increased. After test, the micro-structures and mechanical characteristics of thermal barrier coating were investigated by SEM, XRD. The TGO layer of is Al{sub 2}O{sub 3} formed between BC and TC by periodical thermal shock test, and its change in thickness is inspected by eddy current test(ECT). By ECT test, it is shown that TGO and micro-crack can be detected and it is possible to predict the life of thermal barrier coating

  12. Evaluation of solar thermal storage for base load electricity generation

    OpenAIRE

    Adinberg R.

    2012-01-01

    In order to stabilize solar electric power production during the day and prolong the daily operating cycle for several hours in the nighttime, solar thermal power plants have the options of using either or both solar thermal storage and fossil fuel hybridization. The share of solar energy in the annual electricity production capacity of hybrid solar-fossil power plants without energy storage is only about 20%. As it follows from the computer simulations performed for base load electricity dem...

  13. An analysis of contingencies for making casks available for use during the early years of Federal Waste Management System operations

    International Nuclear Information System (INIS)

    A study has been performed to examine the contingencies that could be pursued by the Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) for shipping spent fuel beginning in 1998. OCRWM's current plan is to initiate operations using early production units of Initiative I truck and rail/barge casks that are presently being designed. Contingencies to this plan were considered in case some unforeseen event occurs that precludes the Initiative I casks from entering into service early in 1998 in sufficient quantities (both numbers and types) to satisfy DOE's shipping needs. Specifically, the study addressed the potential availability of cask systems, selected several cask usage scenarios, determined the requirements for casks under these scenarios, generically assessed different strategies for acquiring casks or the use of casks, and generically assessed cask fabrication capabilities. Issues concerning both domestic and foreign resources were addressed with a focus on the first five years of Federal Waste Management System (FWMS) operation

  14. Thermal hazard evaluation of lauroyl peroxide mixed with nitric acid.

    Science.gov (United States)

    Tsai, Lung-Chang; You, Mei-Li; Ding, Mei-Fang; Shu, Chi-Min

    2012-01-01

    Many thermal runaway incidents have been caused by organic peroxides due to the peroxy group, -O-O-, which is essentially unstable and active. Lauroyl peroxide (LPO) is also sensitive to thermal sources and is incompatible with many materials, such as acids, bases, metals, and ions. From the thermal decomposition reaction of various concentrations of nitric acid (HNO3) (from lower to higher concentrations) with LPO, experimental data were obtained as to its exothermic onset temperature (T0), heat of decomposition (ΔHd), isothermal time to maximum rate (TMRiso), and other safety parameters exclusively for loss prevention of runaway reactions and thermal explosions. As a novel finding, LPO mixed with HNO3 can produce the detonation product of 1-nitrododecane. We used differential scanning calorimetry (DSC), thermal activity monitor III (TAM III), and gas chromatography/mass spectrometer (GC/MS) analyses of the reactivity for LPO and itself mixed with HNO3 to corroborate the decomposition reactions and reaction mechanisms in these investigations. PMID:22763742

  15. Thermal Hazard Evaluation of Lauroyl Peroxide Mixed with Nitric Acid

    Directory of Open Access Journals (Sweden)

    Chi-Min Shu

    2012-07-01

    Full Text Available Many thermal runaway incidents have been caused by organic peroxides due to the peroxy group, –O–O–, which is essentially unstable and active. Lauroyl peroxide (LPO is also sensitive to thermal sources and is incompatible with many materials, such as acids, bases, metals, and ions. From the thermal decomposition reaction of various concentrations of nitric acid (HNO3 (from lower to higher concentrations with LPO, experimental data were obtained as to its exothermic onset temperature (T0, heat of decomposition (ΔHd, isothermal time to maximum rate (TMRiso, and other safety parameters exclusively for loss prevention of runaway reactions and thermal explosions. As a novel finding, LPO mixed with HNO3 can produce the detonation product of 1-nitrododecane. We used differential scanning calorimetry (DSC, thermal activity monitor III (TAM III, and gas chromatography/mass spectrometer (GC/MS analyses of the reactivity for LPO and itself mixed with HNO3 to corroborate the decomposition reactions and reaction mechanisms in these investigations.

  16. Evaluation of thermal conditions inside a vehicle cabin

    Directory of Open Access Journals (Sweden)

    Orzechowski Tadeusz

    2016-01-01

    Full Text Available There are several important factors influencing road accidents. Temperature inside the vehicle is ranked third after alcohol and seat belts. For this reason, maintaining thermal comfort in the passenger compartment is essential. Thermal comfort is provided by the air conditioning system, which consumes much energy. In the case of electrically powered vehicles, this results in a shorter range. Optimization of such systems is therefore required. This paper proposes a set of equations describing the thermal conditions inside the vehicle, which are the result of appropriate energy balances for air, interior elements, and glass. Variable transmission conditions are included for transparent materials exposed to short and long wave radiation. The study focused on unsteady air-conditioning of the vehicle interior. The measurement data was compared with the results obtained through numerical solutions of the proposed set of equations.

  17. Evaluation of thermal conditions inside a vehicle cabin

    Science.gov (United States)

    Orzechowski, Tadeusz; Skrobacki, Zbigniew

    2016-03-01

    There are several important factors influencing road accidents. Temperature inside the vehicle is ranked third after alcohol and seat belts. For this reason, maintaining thermal comfort in the passenger compartment is essential. Thermal comfort is provided by the air conditioning system, which consumes much energy. In the case of electrically powered vehicles, this results in a shorter range. Optimization of such systems is therefore required. This paper proposes a set of equations describing the thermal conditions inside the vehicle, which are the result of appropriate energy balances for air, interior elements, and glass. Variable transmission conditions are included for transparent materials exposed to short and long wave radiation. The study focused on unsteady air-conditioning of the vehicle interior. The measurement data was compared with the results obtained through numerical solutions of the proposed set of equations.

  18. 76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-06-08

    ... 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear... the NRC's spent fuel storage regulations to add the Holtec HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks'' as Certificate of Compliance Number 1032. DATES:...

  19. 76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI90 List of Approved Spent Fuel Storage Casks: HI.... Nuclear Regulatory Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations to add the HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage...

  20. 78 FR 22411 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections

    Science.gov (United States)

    2013-04-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No... direct final rule amending its spent fuel storage regulations by revising the Holtec International, Inc. (Holtec) HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks''...

  1. Standard Practice for Evaluating Thermal Insulation Materials for Use in Solar Collectors

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1994-01-01

    1.1 This practice sets forth a testing methodology for evaluating the properties of thermal insulation materials to be used in solar collectors with concentration ratios of less than 10. Tests are given herein to evaluate the pH, surface burning characteristics, moisture adsorption, water absorption, thermal resistance, linear shrinkage (or expansion), hot surface performance, and accelerated aging. This practice provides a test for surface burning characteristics but does not provide a methodology for determining combustibility performance of thermal insulation materials. 1.2 The tests shall apply to blanket, rigid board, loose-fill, and foam thermal insulation materials used in solar collectors. Other thermal insulation materials shall be tested in accordance with the provisions set forth herein and should not be excluded from consideration. 1.3 The assumption is made that elevated temperature, moisture, and applied stresses are the primary factors contributing to the degradation of thermal insulation mat...

  2. Evaluation of hot corrosion behavior of thermal barrier coatings

    Science.gov (United States)

    Hodge, P. E.; Miller, R. A.; Gedwill, M. A.

    1980-01-01

    Calcium silicate and yttria stabilized zirconia/MCrAlY thermal barrier coating systems on air-cooled specimens were exposed to sodium plus vanadium doped Mach 0.3 combustion gases. Thermal barrier coating endurance was determined to be a strong inverse function of ceramic coating thickness. Coating system durability was increased through the use of higher Cr + Al NiCrAl and CoCrAlY bond coatings. Chemical and electron microprobe analyses supported the predictions of condensate compositions and the determination of their roles in causing spalling of the ceramic coatings.

  3. High Performance Flat Plate Solar Thermal Collector Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Rockenbaugh, Caleb [National Renewable Energy Lab. (NREL), Golden, CO (United States); Dean, Jesse [National Renewable Energy Lab. (NREL), Golden, CO (United States); Lovullo, David [National Renewable Energy Lab. (NREL), Golden, CO (United States); Lisell, Lars [National Renewable Energy Lab. (NREL), Golden, CO (United States); Barker, Greg [National Renewable Energy Lab. (NREL), Golden, CO (United States); Hanckock, Ed [National Renewable Energy Lab. (NREL), Golden, CO (United States); Norton, Paul [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-09-01

    This report was prepared for the General Services Administration by the National Renewable Energy Laboratory. The Honeycomb Solar Thermal Collector (HSTC) is a flat plate solar thermal collector that shows promising high efficiencies over a wide range of climate zones. The technical objectives of this study are to: 1) verify collector performance, 2) compare that performance to other market-available collectors, 3) verify overheat protection, and 4) analyze the economic performance of the HSTC both at the demonstration sites and across a matrix of climate zones and utility markets.

  4. Evaluation of thermal storage materials for solar cooker

    OpenAIRE

    Abate, Solomon

    2014-01-01

    The performance of a solar thermal energy storage system using Lapland granite rock fragments 2-4 cm in diameter were assessed using a scaled-down model. The thesis deals with a selected medium that absorbs and stores solar heat during the day time and releases it when the sun was not shining. A storage rock bed of 5.89 kg with 30 cm x 30 cm base area and 6 cm thickness was placed at the bottom of a solar cooker and painted with black color to increase thermal absorption. The overall performa...

  5. Simulation and evaluation of latent heat thermal energy storage

    Science.gov (United States)

    Sigmon, T. W.

    1980-01-01

    The relative value of thermal energy storage (TES) for heat pump storage (heating and cooling) as a function of storage temperature, mode of storage (hotside or coldside), geographic locations, and utility time of use rate structures were derived. Computer models used to simulate the performance of a number of TES/heat pump configurations are described. The models are based on existing performance data of heat pump components, available building thermal load computational procedures, and generalized TES subsystem design. Life cycle costs computed for each site, configuration, and rate structure are discussed.

  6. Melt Quality Evaluation of Ductile Iron by Pattern Recognition of Thermal Analysis Cooling Curves

    Institute of Scientific and Technical Information of China (English)

    LI Zhenhua; LI Yanxiang; ZHOU Rong

    2008-01-01

    The melt quality of ductile iron can be related to the melt's thermal analysis cooling curve. The freezing zone of the thermal analysis cooling curve was found to indicate the melt quality of the ductile iron. A comprehensive difference parameter, Ω, of the thermal analysis cooling curves was found to be related to the properties of ductile iron melts such as composition, temperature, and graphite morphology. As Ω ap- proached O, the thermal analysis cooling curves were found to come together with all the properties indicat- ing melt quality about the same. A database of thermal analysis cooling curves related to the properties of the ductile iron melts was set up as a basis for a method to accurately evaluate the melt quality of ductile iron by pattern recognition of thermal analysis cooling curves. The quality of a ductile iron melt can then be immediately determined by comparing its thermal analysis cooling curve freezing zone shape to those in the database.

  7. Thermal fatigue damage evaluation of a PWR NPP steam generator injection nozzle model subjected to thermal stratification phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Leite da Silva, Luiz, E-mail: silvall@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear, CDTN/CNEN, Av. Presidente Antonio Carlos, 6627, Campus UFMG, Pampulha Belo Horizonte, MG CEP 31.270-901 (Brazil); Rodrigues Mansur, Tanius, E-mail: tanius@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear, CDTN/CNEN, Av. Presidente Antonio Carlos, 6627, Campus UFMG, Pampulha Belo Horizonte, MG CEP 31.270-901 (Brazil); Cimini Junior, Carlos Alberto, E-mail: cimini@demec.ufmg.b [Departamento de Engenharia Mecanica da Universidade Federal de Minas Gerais, DEMEC/UFMG, Av. Presidente Antonio Carlos, 6627, Pampulha, Belo Horizonte, MG CEP 31.270-901 (Brazil)

    2011-03-15

    Thermal stratification phenomenon with the same thermodynamic steam generator (SG) injection nozzle parameters was simulated. After 41 experiments, the experimental section was dismantled; cut and specimens were made of its material. Other specimens were made of the preserved pipe material. By comparing their fatigue tests results, the pipe material damage was evaluated. The water temperature layers and also the outside pipe wall temperatures were measured at the same level. Strains outside the pipe in 7 positions were measured. The experimental section develops thermal stratified flows, stresses and strains caused enlargement of material grain size and reduction in fatigue life.

  8. Sensitivity Analysis Applied to the Validation of the 10 B Capture Reaction in Nuclear Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S

    2004-03-18

    Boron has commonly been used in nuclear fuel casks to ensure a sufficient margin of subcriticality. The amount of boron used in most casks far exceeds the amount of boron present in any of the available benchmark experiments. Such heavy loadings of boron in the casks may result in considerable spectral differences as compared to the benchmarks, resulting in boron sensitivities that are very different from those of the benchmarks. Before the calculations to determine the nuclear safety margin for various fuel loadings are deemed acceptable, as part of the safety basis, the computer code and cross sections must be validated against experimental benchmarks that cover the area of applicability of the proposed cask design. Therefore, this study was performed to determine if these available benchmarks can be used to validate a criticality code and neutron cross sections for the fuel casks. The sensitivity/uncertainty methodology has been applied to several application cask systems with different boron areal densities. Although, the sensitivities of the nuclear fuel cask applications are not completely covered by the set of benchmarks that were used in this study with regard to the 10B capture cross section, the effect of this lack of coverage on the keff is minimal. Thus, the experimental biases are determined to be appropriate for the cask systems, and no additional bias (penalty) due to high boron loading need be imposed.

  9. Modelling of RBMK-1500 SNF storage casks activation during very long term storage.

    Science.gov (United States)

    Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Ragaisis, Valdas

    2016-09-01

    Existing interim spent nuclear fuel storage facility (SNFSF) at Ignalina nuclear power plant in Lithuania is fully loaded with CASTOR(®)RBMK-1500 and CONSTOR(®)RBMK-1500 storage casks. The planned lifetime of these casks is 50 years and the first loaded cask was moved to the SNFSF in 1999. The start of operation of disposal facility in Lithuania is foreseen later than the planned interim storage ends. So, the possibilities to extend the storage period over 50 years should be considered. Therefore, the casks decommissioning issues should be taken into account, as due to prolonged neutron irradiation casks materials could became activated. This paper presents modelling results of storage casks neutron activation during 300 year storage period. Modelling results show, that after 50 years of storage, side-wall and bottom of CASTOR(®)RBMK-1500 cask are activated above clearance criteria. However, for 100-300 year storage period all of the casks components could be free released. PMID:27344524

  10. 78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks... Regulatory Commission (NRC) is proposing to amend its spent fuel storage regulations by revising the NAC... within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 3 to Certificate...

  11. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181... spent fuel storage cask designs. The NRC subsequently issued a final rule on November 21, 2008 (73 FR... COMMISSION 10 CFR Part 72 RIN 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System,......

  12. 75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150--AI88 List of Approved Spent Fuel Storage Casks: NAC.... Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by... 72. PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL,...

  13. 75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision... Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

  14. 77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2012-01-27

    ... 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 2 AGENCY: Nuclear... amended the NRC's spent fuel storage regulations by revising the NAC International, Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 2...

  15. Regulation of dopamine release by CASK-β modulates locomotor initiation in Drosophila melanogaster

    Directory of Open Access Journals (Sweden)

    Justin eSlawson

    2014-11-01

    Full Text Available CASK is an evolutionarily conserved scaffolding protein that has roles in many cell types. In Drosophila, loss of the entire CASK gene or just the CASK-β transcript causes a complex set of adult locomotor defects. In this study, we show that the motor initiation component of this phenotype is due to loss of CASK-β in dopaminergic neurons and can be specifically rescued by expression of CASK-β within this subset of neurons. Functional imaging demonstrates that mutation of CASK-β disrupts coupling of neuronal activity to vesicle fusion. Consistent with this, locomotor initiation can be rescued by artificially driving activity in dopaminergic neurons. The molecular mechanism underlying this role of CASK-β in dopaminergic neurons involves interaction with Hsc70-4, a molecular chaperone previously shown to regulate calcium-dependent vesicle fusion. These data suggest that there is a novel CASK-β-dependent regulatory complex in dopaminergic neurons that serves to link activity and neurotransmitter release.

  16. Licensing and safety issues associated with dry cask storage update. Panel Discussion

    International Nuclear Information System (INIS)

    Full text of publication follows: Panelists from the nuclear industry, cask vendors, the U.S. Department of Energy (DOE), and the U.S. Nuclear Regulatory Commission will speak to the current status of licensing casks for interim storage and shipping to the DOE permanent site and alternate interim private storage initiatives. Subject coverage will include a broad range of relevant issues. (authors)

  17. Preliminary investigation of aluminium foam as an energy absorber for nuclear transportation cask

    Energy Technology Data Exchange (ETDEWEB)

    Rajendran, R. [BARC Facilities, Bhabha Atomic Research Centre, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: rajurajendr@yahoo.co.in; Prem Sai, K.; Chandrasekar, B. [BARC Facilities, Bhabha Atomic Research Centre, Kalpakkam, Tamilnadu 603 102 (India); Gokhale, A. [Defence Metallurgical Research Laboratory, Kanchanbagh, Hyderabad 500 058 (India); Basu, S. [BARC Facilities, Bhabha Atomic Research Centre, Kalpakkam, Tamilnadu 603 102 (India)

    2008-10-15

    Closed cell aluminum foam is investigated for its impact energy absorption characteristics. For this purpose, a drop hammer of 106 kg was fabricated. A free-fall drop tower was used for the experiments. The hammer was impacted on the rigid foundation with and without aluminium foam at its bottom. Acceleration-time history was recorded for each drop. Deflection of the foam undergoing impact was measured. Compression test was carried out on a foam cylinder to obtain the representative stress-strain diagram from which energy-deflection diagram was derived. Gibson-Ashby's plateau stress-density relation was applied to evaluate the energy-deflection characteristics of foams of different densities, which were eventually applied to the theoretical predictions. Force reduction factor offered by foam is attractive enough to considering, it as the candidate for sacrificial member of the transportation cask.

  18. U.S. Regulatory Research Program for Implementation of Burnup Credit in Transport Casks

    International Nuclear Information System (INIS)

    In 1999 the U.S. Nuclear Regulatory Commission (U.S. NRC) initiated a research program to support the development of technical bases and guidance that would facilitate the implementation of burnup credit into licensing activities for transport and dry cask storage. This paper reviews the following major areas of investigation: (1) specification of axial burnup profiles, (2) assumption on cooling time, (3) allowance for assemblies with fixed and removable neutron absorbers, (4) the need for a burnup margin for fuel with initial enrichments over 4 wt %, and (5) evaluation of assay data and critical experiments. The capabilities of a new computational tool that facilitates the performance and coupling of the depletion and criticality analyses needed for burnup credit are also discussed

  19. Computational fluid dynamics evaluation of liquid food thermal process in a brick shaped package

    Directory of Open Access Journals (Sweden)

    Pedro Esteves Duarte Augusto

    2012-03-01

    Full Text Available Food processes must ensure safety and high-quality products for a growing demand consumer creating the need for better knowledge of its unit operations. The Computational Fluid Dynamics (CFD has been widely used for better understanding the food thermal processes, and it is one of the safest and most frequently used methods for food preservation. However, there is no single study in the literature describing thermal process of liquid foods in a brick shaped package. The present study evaluated such process and the influence of its orientation on the process lethality. It demonstrated the potential of using CFD to evaluate thermal processes of liquid foods and the importance of rheological characterization and convection in thermal processing of liquid foods. It also showed that packaging orientation does not result in different sterilization values during thermal process of the evaluated fluids in the brick shaped package.

  20. Evaluation of hot spot factors for thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950degC in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly. (author)

  1. Fabrication, characterization, and thermal property evaluation of silver nanofluids

    Science.gov (United States)

    Noroozi, Monir; Radiman, Shahidan; Zakaria, Azmi; Soltaninejad, Sepideh

    2014-11-01

    Silver nanoparticles were successfully prepared in two different solvents using a microwave heating technique, with various irradiation times. The silver nanoparticles were dispersed in polar liquids (distilled water and ethylene glycol) without any other reducing agent, in the presence of the stabilizer polyvinylpyrrolidone (PVP). The optical properties, thermal properties, and morphology of the synthesized silver particles were characterized using ultraviolet-visible spectroscopy, photopyroelectric technique, and transmission electron microscopy. It was found that for the both solvents, the effect of microwave irradiation was mainly on the particles distribution, rather than the size, which enabled to make stable and homogeneous silver nanofluids. The individual spherical nanostructure of self-assembled nanoparticles has been formed during microwave irradiation. Ethylene glycol solution, due to its special properties, such as high dielectric loss, high molecular weight, and high boiling point, can serve as a good solvent for microwave heating and is found to be a more suitable medium than the distilled water. A photopyroelectric technique was carried out to measure thermal diffusivity of the samples. The precision and accuracy of this technique was established by comparing the measured thermal diffusivity of the distilled water and ethylene glycol with values reported in the literature. The thermal diffusivity ratio of the silver nanofluids increased up to 1.15 and 1.25 for distilled water and ethylene glycol, respectively.

  2. Evaluation of the Thermodynamic Models for the Thermal Diffusion Factor

    DEFF Research Database (Denmark)

    Gonzalez-Bagnoli, Mariana G.; Shapiro, Alexander; Stenby, Erling Halfdan

    2003-01-01

    Over the years, several thermodynamic models for the thermal diffusion factors for binary mixtures have been proposed. The goal of this paper is to test some of these models in combination with different equations of state. We tested the following models: those proposed by Rutherford and Drickamer...

  3. Fabrication, characterization, and thermal property evaluation of silver nanofluids

    OpenAIRE

    Noroozi, Monir; Radiman, Shahidan; Zakaria, Azmi; Soltaninejad, Sepideh

    2014-01-01

    Silver nanoparticles were successfully prepared in two different solvents using a microwave heating technique, with various irradiation times. The silver nanoparticles were dispersed in polar liquids (distilled water and ethylene glycol) without any other reducing agent, in the presence of the stabilizer polyvinylpyrrolidone (PVP). The optical properties, thermal properties, and morphology of the synthesized silver particles were characterized using ultraviolet-visible spectroscopy, photopyro...

  4. Evaluation of the underground soil thermal storage properties in Libya

    Energy Technology Data Exchange (ETDEWEB)

    Nassar, Y.; ElNoaman, A.; Abutaima, A.; Yousif, S.; Salem, A. [Solar Energy Laboratory, Faculty of Engineering and Technology, Sebha University, P.O. Box 68, Brack (Libya)

    2006-04-15

    Experimental investigation was conducted of temperature distribution through the underground soil of Tripoli (Capital of Libya). The aim of the experiment is to monitor the temperature variation of the underground soil under a depth of 4m and around the year, in order to know the thermal capacity ability of the soil to be used as a seasonal thermal storage. The measurements covered two types of systems: the first one is dry soil and the second is dry soil covered by a glass sheet. The measurements indicate that, at a depth of 4m, the average temperatures for the dry and dry-glass covered systems are 21, 46{sup o}C, with maximum temperatures of 21.5 and 47{sup o}C during December and January, and the minimum temperatures occurred in May and June, are reached values of 19, 44{sup o}C, respectively. The temperatures for the two systems were almost constant through the year and fluctuating with a monthly period of 2p/12. Results show that, the underground thermal capacity can be used as a source of heating and cooling of buildings leading to reduce the energy consumption in this application. Furthermore, for industrial and domestic heating processes, one can utilize the dry-glass covered system to cover a significant part of the heating load. Anyhow, the experimental study may not applicable everywhere, so an analytical presentation for the system will be necessary to save money and efforts. The first step to put the analytical model in reality is to get the thermal properties of the underground soil, and this is the aim of the present study. The paper described the followed procedure during theoretical-heat transfer approach. The thermal properties were presented as a function of the ground depth, furthermore, the paper presented the measured temperatures of the two systems for Tripoli underground soil. [Author].

  5. Effectively meeting spent fuel storage needs with a family of dry storage casks

    International Nuclear Information System (INIS)

    During 1988--89, a number of nuclear utilities have announced their intent of developing supplemental spent fuel storage. These on-site facilities are to be operable by 1991--93. This paper discusses how the Castor ductile cast iron (DCI) storage casks is a tested and licensed means of meeting this fuel storage need. Since 1986, a total of 14 casks have been sold to the Virginia Power Co. (V.P.). Eight casks are now loaded and in storage at the V.P. Surry Nuclear Station. These casks are directly pool loaded and moved to a storage pad using straight forward handling operations. Once on the pad, there is no further need for cask operation or maintenance with this sealed and passive storage system

  6. The dry storage cask in interim storage facility and safeguards activity

    International Nuclear Information System (INIS)

    The Japan Atomic Power Company (JAPC) is preparing for interim storage of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel and to serve for about 50 years in RFSC. Metallic dry casks have already been used for spent fuel dry storage at Tokai No.2 power station. But, RFSC is not exactly the same as the dry storage facility in Tokai No.2 power station, for example, casks are transported out side of the reactor site and RFSC has no fuel handling system. Therefore, additional implementation of safeguards is necessary. This report introduces the design and handling of metallic dry casks for RFSC and the currently developing status of the safeguards activity such as containment and surveillance for the fuel loading at the power station, the cask receipt and storage at RFSC, etc. (author)

  7. Thermal Mechanical Stability of Single-Crystal-Oxide Refractive Concentrators Evaluated for High-Temperature Solar-Thermal Propulsion

    Science.gov (United States)

    Jacobson, Nathan S.; Jacobson, Nathan S.; Miller, Robert A.

    1999-01-01

    Recently, refractive secondary solar concentrator systems were developed for solar thermal power and propulsion (ref. 1). Single-crystal oxides-such as yttria-stabilized zirconia (Y2O3-ZrO2), yttrium aluminum garnet (Y3Al5O12, or YAG), magnesium oxide (MgO), and sapphire (Al2O3)-are candidate refractive secondary concentrator materials. However, the refractive concentrator system will experience high-temperature thermal cycling in the solar thermal engine during the sun/shade transition of a space mission. The thermal mechanical reliability of these components in severe thermal environments is of great concern. Simulated mission tests are important for evaluating these candidate oxide materials under a variety of transient and steady-state heat flux conditions. In this research at the NASA Lewis Research Center, a controlled heat flux test approach was developed for investigating the thermal mechanical stability of the candidate oxide. This approach used a 3.0-kW continuous-wave (wavelength, 10.6 mm) carbon dioxide (CO2) laser (ref. 2). The CO2 laser is especially well-suited for single-crystal thermal shock tests because it can directly deliver well-characterized heat energy to the oxide surfaces. Since the oxides are opaque at the 10.6-mm wavelength of the laser beam, the light energy is absorbed at the surfaces rather than transmitting into the crystals, and thus generates the required temperature gradients within the specimens. The following figure is a schematic diagram of the test rig.

  8. A Facile Approach to Evaluate Thermal Insulation Performance of Paper Cups

    Directory of Open Access Journals (Sweden)

    Yudi Kuang

    2015-01-01

    Full Text Available Paper cups are ubiquitous in daily life for serving water, soup, coffee, tea, and milk due to their convenience, biodegradability, recyclability, and sustainability. The thermal insulation performance of paper cups is of significance because they are used to supply hot food or drinks. Using an effective thermal conductivity to accurately evaluate the thermal insulation performance of paper cups is complex due to the inclusion of complicated components and a multilayer structure. Moreover, an effective thermal conductivity is unsuitable for evaluating thermal insulation performance of paper cups in the case of fluctuating temperature. In this work, we propose a facile approach to precisely analyze the thermal insulation performance of paper cups in a particular range of temperature by using an evaluation model based on the MISO (Multiple-Input Single-Output technical theory, which includes a characterization parameter (temperature factor and a measurement apparatus. A series of experiments was conducted according to this evaluation model, and the results show that this evaluation model enables accurate characterization of the thermal insulation performance of paper cups and provides an efficient theoretical basis for selecting paper materials for paper cups.

  9. Thermal hydraulic evaluation of advanced wire-wrapped assemblies

    International Nuclear Information System (INIS)

    The thermal-hydraulic analyses presented in this report are based on application of the subchannel concept in association with the use of bulk parameters for coolant velocity and coolant temperature within a subchannel. The interactions between subchannels are due to turbulent interchange, pressure-induced diversion crossflow, directed sweeping crossflow induced by the helical wire wrap, and transverse thermal conduction. The FULMIX-II computer program was successfully developed to perform the steady-state temperature predictions for LMFBR fuel assemblies with the reference straight-start design and the advanced wire-wrap designs. Predicted steady-state temperature profiles are presented for a typical CRBRP 217-rod wire-wrapped assembly with the selected wire-wrap designs

  10. Evaluation of hand applied naled thermal fog for Wyeomyia control.

    Science.gov (United States)

    Curtis, G A; Carlson, D B

    1990-09-01

    Tests on the effect of hand applied naled thermal fog, both as a single treatment on one day/week and a single treatment on 3 successive days, did not control Wyeomyia vanduzeei and Wy. mitchellii. Five-min landing/biting counts in a native oak/palm woodland demonstrated that single applications produced an average landing rate decrease of 13%. Treatments 3 days in succession did not suppress the landing rate. PMID:1977876

  11. Interpretation of FAENA and TIFFSS experiment : Comparison of temperature evaluation methods on thermal striping

    OpenAIRE

    笠原 直人; Yves LEJEAIL

    2000-01-01

    Since thermal striping is a coupled thermohydraulic and thermomechanical phenomenon, sodium mock-up tests were usually required to confirm structural integrity. CEA and JNC have developed evaluation procedures of thermal striping to establish design-by-analysis methodology for this phenomenon. In order to compare and to validate these methods, two benchmark problems were planned under EJCC contract. One of benchmarks provided by CEA is temperature and fatigue evaluation of tubes and plates te...

  12. Thermal performance evaluation of a conical solar water heater integrated with a thermal storage system

    International Nuclear Information System (INIS)

    Highlights: • To track the critical mass flow rate for maximal CSWH efficiency. • A high flow rate deteriorated the thermal stratification in the attached tank. • The CSWH operates more efficiently if the fluid is heated at a critical flow rate. • The higher temperature rise with vacuum glass was observed for all flow rates. - Abstract: In the present research, a conical solar water heater (CSWH) with an attached thermal storage tank, with or without a vacuum glass absorber, was analyzed under different operating conditions. For maximum solar radiation of the system, the collector was equipped with a dual-axis tracking system and sun sensor, which kept the system oriented towards the sun at every instant during its operation. A forced cooling system circulated fluid to remove the solar heat from the absorber surface. Performance analyses with and without the vacuum glass absorber were conducted at different mass flow rates, inlet temperatures, and solar irradiation values. The influence of the vacuum glass cover and all operational parameters on the collector efficiency, outlet temperature, and thermal stratification were investigated. The efficiency increased with increasing inlet flow rate, and the maximum efficiency was obtained at a critical flow rate of 6 L/min. When the flow rate was increased beyond this critical value, the efficiency began to decrease. The temperature rise of the working fluid with vacuum glass at a high rate of insolation was considerably higher than without a vacuum glass for all flow rates. Use of a high flow rate deteriorated the thermal stratification process in the storage tank, while it increased the efficiency of the conical solar water-heating system. It can be concluded that the CSWH operates more efficiently if the fluid is heated slightly at a critical flow rate

  13. CLASSIFICATION OF THE MGR CARRIER/CASK HANDLING SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    J.A. Ziegler

    2001-02-08

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) carried cask handling system structures, systems and components (SSCs) performed by the MGR Preclosure Safety and Systems Engineering Section. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 2000). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 2000).

  14. Numerical evaluation of the thermal performances of roof-mounted radiant barriers

    CERN Document Server

    Miranville, Frédéric; Lucas, Franck; Johan, Seriacaroupin

    2014-01-01

    This paper deals with the thermal performances of roof-mounted radiant barriers. Using dynamic simulations of a mathematical model of a whole test cell including a radiant barrier installed between the roof top and the ceiling, the thermal performance of the roof is calculated. The mean method is more particularly used to assess the thermal resistance of the building component and lead to a value which is compared to the one obtained for a mass insulation product such as polyurethane foam. On a further stage, the thermal mathematical model is replaced by a thermo-aeraulic model which is used to evaluate the thermal resistance of the roof as a function of the airflow rate. The results shows a better performance of the roof in this new configuration, which is widely used in practice. Finally, the mathematical relation between the thermal resistance and the airflow rate is proposed.

  15. Spent fuel storage and transport cask decontamination and modification. An overview of management requirements and applications based on practical experience

    International Nuclear Information System (INIS)

    A large increase in the number of casks required for transport and/or storage of spent fuel is forecast into the next century. The principal requirement will be for increased number of storage and dual purpose (transport/storage) casks for interim storage of spent fuel prior to reprocessing or permanent disposal in both on-site and off-site storage facilities. Through contact with radioactive materials spent fuel casks will be contaminated on both internal and external surfaces. In broad terms, cask contamination management can be defined by three components: minimisation, prevention and decontamination. This publication is a compilation of international experience with cask contamination problems and decontamination practices. The objective is to present current knowledge and experience as well as developments, trends and potential for new applications in this field. Furthermore, the report may assist in new design or modification of existing casks, cask handling systems and decontamination equipment

  16. Evaluation of thermal perception in schoolyards under Mediterranean climate conditions

    Science.gov (United States)

    Antoniadis, D.; Katsoulas, N.; Papanastasiou, D.; Christidou, V.; Kittas, C.

    2016-03-01

    The aim of this paper was to study qualitatively and quantitatively the thermal perception and corresponding heat stress conditions that prevail in two schoolyards in a coastal city in central Greece. For this purpose, meteorological parameters (i.e., wind speed, temperature, relative humidity, solar radiation) were recorded at 70 and 55 measuring points in the schoolyards, from 14:00 to 15:30 local time, during May and June of 2011. The measuring points were distributed so as to get measurements at points (a) directly exposed to the sun, (b) under the shadow of trees and building structures, and (c) near building structures. Cluster analysis was applied to group observations and revealed places that are microclimatically homogeneous. Thermal perception and heat stress conditions were assessed by means of the physiologically equivalent temperature (PET, °C), and the results are presented in relevant charts. The impact of material's albedo, radiation's reflection by structures and obstacles, and different tree species on thermal perception and heat stress conditions was also assessed. The analysis showed that trees triggered a reduction of incident solar radiation that ranged between 79 and 94 % depending on tree's species, crown dimension, tree height, and leaf area. PET values were mainly affected by solar radiation and wind speed. Trees caused a reduction of up to 37 % in PET values, while a 1-m s-1 increase in wind speed triggered a reduction of 3.7-5.0 °C in PET value. The effective shading area in the two schoolyards was small, being 27.5 and 11 %. The results of this study could be exploited by urban planning managers when designing or improving the outdoor environment of a school complex.

  17. Structural evaluation of candidate space shuttle thermal protection systems

    Science.gov (United States)

    Burns, A. B.

    1972-01-01

    The characteristics and development of a lightweight reusable thermal protection system for the space shuttle are discussed. The test articles consisted of metallic substrates with upper surfaces covered with all-silica, reusable, surface insulation material. The material is processed in the form of tiles. The external surfaces of the tiles are provided with a coating system which consists of a borosilicate coating with a silicon carbide emittance agent and impregnation with a hydrophobic agent. The finished tiles are attached to the metal substrate by adhesive bonding. Charts and graphs of the properties of the material are provided.

  18. Evaluation of the thermal ageing of duplex stainless steels

    International Nuclear Information System (INIS)

    Three methods have been investigated to follow up the thermal ageing of duplex stainless steels: microhardness tests, instrumented ultramicrohardness tests and Small Angle Neutron Scattering (SANS) techniques. The values measured with these methods have been correlated with pertinent parameters of the metallurgical ageing phenomenon determined by Atom-Probe. These methods seem to be sensitive and reproducible enough to detect and follow up the ageing of duplex stainless steels. They can be applied on small samples (chips) drawn from in-service components. (authors). 10 refs., 9 figs., 3 tabs

  19. Evaluation of in-situ thermal energy storage for lunar based solar dynamic systems

    Science.gov (United States)

    Crane, Roger A.

    1991-01-01

    A practical lunar based thermal energy storage system, based on locally available materials, could significantly reduce transportation requirements and associated costs of a continuous, solar derived power system. The concept reported here is based on a unique, in-situ approach to thermal energy storage. The proposed design is examined to assess the problems of start-up and the requirements for attainment of stable operation. The design remains, at this stage, partially conceptional in nature, but certain aspects of the design, bearing directly on feasibility, are examined in some detail. Specifically included is an engineering evaluation of the projected thermal performance of this system. Both steady state and start-up power requirements are evaluated and the associated thermal losses are evaluated as a basis for establishing potential system performance.

  20. Analysis and design of dry cask storage pads for plant hatch Isfsi

    International Nuclear Information System (INIS)

    An independent spent fuel storage installation (ISFSI) at Southern Company's Edwin I. Hatch Nuclear Plant (HNP) was completed, licensed, and put in service in the summer of 2000. Currently this dry cask on-site storage facility provides a temporary spent fuel storage for three Holtec HI-STAR 100 system casks. After re-racking and rod consolidation efforts, the HNP ISFSI was necessary to maintain a full core discharge capacity of its spent nuclear fuel pools and also to temporarily delay a need for a permanent off-site spent nuclear fuel repository. The HNP ISFSI was carried out to meet the following three main criteria established at the beginning of the HNP Spent Fuel Storage Project. These three criteria were 1) to use the general license approach which utilizes the license of the cask vendor rather than obtaining a site-specific license, 2) to select only dry cask products that are intended for dual purpose licensing, and 3) to acquire sufficient dry cask storage capacity to fully meet the plant's need. This paper describes the major steps of analysis and design of dry cask storage pads for Plant Hatch ISFSI. Results showed that HNP ISFSI met the applicable codes, regulatory and cask vendor requirements. (author)

  1. Two decades of experience with more than 750 CASTOR and CONSTOR transport and storage casks

    International Nuclear Information System (INIS)

    In 1983 the world-wide first dual purpose transport and storage cask - a CASTOR registered Ic-DIORIT - was loaded in Wuerenlingen/ Switzerland. Meanwhile CASTOR registered casks are used at 24 sites on four continents. Spent fuel assemblies of PWR, BWR, VVER, RBMK, FBR, MTR and THTR as well as vitrified high active waste canisters are transported and/or stored in these kinds of monolithic metal casks. MOX spent fuel of PWR and BWR has been loaded, too. Starting in the mid of the 90s, GNB developed the new CONSTOR registered cask concept, which is based on a double liner technology with a layer of heavy concrete as shielding material inbetween. This CONSTOR registered cask concept fulfils all design criteria for transport and for storage given by the IAEA recommendations and by national authorities. Up to now, more than 750 CASTOR registered and CONSTOR registered casks have been used for transports or/and loaded for longterm interim storage. More than two decades of storage experience attest to the excellent behavior of the casks including the metallic gaskets and the tightness monitoring system. Detailed measurements of temperatures and of gamma and neutron dose rates have shown in each case that the safety requirements have been fulfilled. These measurements allowed to reduce unnecessary safety margins to optimize the benefit for the user

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

    International Nuclear Information System (INIS)

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

  3. Contamination transfers during fuel transport cask loading. A concrete situation

    Energy Technology Data Exchange (ETDEWEB)

    Fournel, B.; Turchet, J.P.; Faure, S.; Allinei, P.G. [DEN/DED Centre d' Etudes de Cadarache, 13 - Saint Paul lez Durance (France); Briquet, L. [EDF GENV, 93 - Saint Denis (France); Baubet, D. [SGS Qualitest Industrie, 30 - Pont Saint Esprit (France)

    2002-07-01

    In 1998, a number of contamination cases detected during fuel shipments have been pointed out by the french nuclear safety authority. Wagon and casks external surfaces were partly contaminated upon arrival in Valognes railway terminal. Since then, measures taken by nuclear power plants operators in France and abroad solved the problem. In Germany, a report analyzing the situation in depth has been published in which correctives actions have been listed. In France, EDF launched a large cleanliness program (projet proprete radiologique) in order to better understand contamination transfers mechanisms during power plants exploitation and to list remediation actions to avoid further problems. In this context, CEA Department for Wastes Studies at Cadarache (CEA/DEN/DED) was in charge of a study about contamination transfers during fuel elements loading operations. It was decided to lead experiments for a concrete case. The loading of a transport cask at Tricastin-PWR-1 was followed in november 2000 and different analysis comprising water analysis and smear tests analysis were carried out and are detailed in this paper. Results are discussed and qualitatively compared to those obtained in Philippsburg-BWR, Germany for a similar set of tests. (authors)

  4. Pilot study dismantlement of 20 lead-lined shipping casks

    International Nuclear Information System (INIS)

    This report describes a pilot study conducted at the INEL to dismantle lead-lined casks and shielding devices, separate the radiologically contaminated and hazardous materials, and recycle resultant scrap lead. The facility areas where the work was performed, dismantlement methods, and process equipment are described. Issues and results associated with recycling the lead as a free-released scrap metal are presented and discussed. Data and results from the pilot study are summarized and presented. The study concluded that cask dismantlement at the INEL can be performed as a legitimate recycling activity for scrap lead. Ninety-one percent of the lead recovered passed free-release criteria. The value of the 50,375 lb of recovered lead is approximately $0.45/lb. Resultant waste streams can be satisfactorily treated and disposed. Only very low levels of bulk radiological contamination (47 picocuries/gram of 137 Cs and 3.2 picocuries/gram of 6OCo) were detected in the lead rejected for free release

  5. Transfer cask system design activities: status and plan

    Energy Technology Data Exchange (ETDEWEB)

    Locke, D., E-mail: darren.locke@f4e.europa.eu [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Gutierrez, C. Gonzalez; Damiani, C.; Gracia, V. [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Friconneau, J.-P.; Martins, J.-P.; Blight, J. [ITER Organisation, CS 90 046, 13067St. Paul Lez Durance Cedex (France)

    2011-10-15

    The ITER Cask and Plug Remote Handling System (CPRHS), a.k.a. Transfer Cask System, is a critical element of the ITER Remote Maintenance System (IRMS) devoted to transportation of components between the Tokamak building and Hot Cell. Due to the necessary confinement of contaminated components the CPRHS is defined as Safety Importance Class 1 (SIC-1) plus the mobile nature of the CPRHS brings with it a significant number of complex interfaces with other ITER sub-systems. With a total CPRHS fleet in excess of 20 units, including seven typologies, the management of design and procurement needs to be carefully planned and implemented to ensure compliance with ITER's requirements. Fusion for Energy (F4E) and its beneficiaries/contractors are currently working under ITER Task Agreements (ITAs) on the conceptual design of the CPRHS and, following the signing of the Procurement Arrangement (PA) in mid 2012, will take responsibility for the entire CPRHS fleet. F4E must, therefore, develop a robust strategy to meet the needs of both ITER machine assembly (for which a number of CPRHS units will be utilised) and the remote maintenance of ITER. Within this context this paper will present the status of the current CPRHS design activities, highlight some of the significant issues which will be faced during procurement and present the overall strategy which is being implemented by F4E in order to meet these challenging objectives.

  6. Non-destructive evaluation of thermal aging of cast duplex stainless steel

    International Nuclear Information System (INIS)

    Cast duplex stainless steel is frequently used in main coolant pipes and reactor coolant pump casings in nuclear power plants because of its excellent material strength, toughness and superior corrosion resistance. It is known, however, that thermal aging occurs when this material is exposed to temperatures over 300degC for extended periods of time. As a result, the material toughness decreases. It is necessary therefore to evaluate changes in the mechanical properties of this material caused by thermal aging using non-destructive methods for the maintenance and management of components made of cast duplex stainless steel. In order to develop a non-destructive technique for evaluating the toughness reduction of cast duplex stainless steel due to thermal aging, five types of non-destructive techniques were compared. These include ultrasonic sound velocity measurement, the thermoelectric power measurement the electric resistance method, the SQUID (Superconducting Quantum Interface Device) method, and the positron annihilation method. The thermal aging detectability of each technique was compared and examined in experiments using specimens on which accelerated thermal aging had been carried out. It was concluded that the thermoelectric power measurement was the most effective technique for evaluating thermal aging because the correlation coefficient between the non-destructive evaluation parameters and the mechanical properties of aged materials was high and the dispersion of measurements was small. (author)

  7. Internal Thermal Control System Hose Heat Transfer Fluid Thermal Expansion Evaluation Test Report

    Science.gov (United States)

    Wieland, P. O.; Hawk, H. D.

    2001-01-01

    During assembly of the International Space Station, the Internal Thermal Control Systems in adjacent modules are connected by jumper hoses referred to as integrated hose assemblies (IHAs). A test of an IHA has been performed at the Marshall Space Flight Center to determine whether the pressure in an IHA filled with heat transfer fluid would exceed the maximum design pressure when subjected to elevated temperatures (up to 60 C (140 F)) that may be experienced during storage or transportation. The results of the test show that the pressure in the IHA remains below 227 kPa (33 psia) (well below the 689 kPa (100 psia) maximum design pressure) even at a temperature of 71 C (160 F), with no indication of leakage or damage to the hose. Therefore, based on the results of this test, the IHA can safely be filled with coolant prior to launch. The test and results are documented in this Technical Memorandum.

  8. FIELD EVALUATION OF TERRA THERM IN SITU THERMAL DESTRUCTION (ISTD) TREATMENT OF HEXACHLOROCYCLOPENTADIENE

    Science.gov (United States)

    This report summarizes the U.S. Environmental Protection Agency (EPA) Superfund Innovative Technology Evaluation (SITE) Program evaluation of the In Situ Thermal Destruction (ISTD) technology, developed by others, was refined by TerraTherm, Inc. The demonstration was designed to ...

  9. Validation of perceptual strain index to evaluate the thermal strain in experimental hot conditions

    Directory of Open Access Journals (Sweden)

    Habibollah Dehghan

    2015-01-01

    Conclusions: The research findings showed when there is no access to other forms of methods to evaluate the heat stress, it can be used the PeSI in evaluating the strain because of its favorable correlation with the thermal strain.

  10. Operation and maintenance of spent fuel storage and transportation casks/containers

    International Nuclear Information System (INIS)

    Member States have a growing need for casks for spent fuel storage and transportation. A variety of casks has been developed and is in use at an increasing number of sites. This has resulted in an accumulation of experience that will provide valuable information for other projects in spent fuel management. This publication provides a comprehensive review of information on the cask operation and maintenance associated with spent fuel storage. It draws upon generic knowledge from industrial experience and applications and is intended to serve as a basis for better planning and implementation in future projects

  11. Material specification and quality control program for ductile iron spent fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Rehmer, B.; Frenz, H.; Weidlich, S.; Kuehn, H.D.

    1995-12-31

    In the process of testing spent fuel casks, BAM is gaining a lot of relevant data regarding the quality level of Ductile Cast Iron (DCI). This paper discusses the basic parameters governing the material behavior of ferritic and ferritic-pearlitic DCI and reviews the development of cask quality over the last years. The effect of microstructure and sample size on the fracture toughness of DCI is discussed. The results of a test program show the prominent effect of pearlite content and graphite nodule structure in the mechanical and fracture toughness characteristics of DCI. This observation is important for quality assurance programs for shipping and storage casks of radioactive materials.

  12. Thermal paint production: the techno-economic evaluation of muscovite as insulating additive.

    Directory of Open Access Journals (Sweden)

    Gabriela Fernandes Ribas

    2016-09-01

    Full Text Available Muscovite is known by its thermal and electrical insulating properties. Based on this, it was hypothesized that its addition on paints should increase the thermal resistance. The use of muscovite as mineral insulating is pointed out as advantageous due to its low cost compared to other materials used for this purpose, such as the ceramic microsphere. The use of a low cost material could open the access to the medium and low income families, implying two aspects: the life quality increase by thermal comfort and the increase of energy saving. Thus, this part of the population could open a new market to thermal paints. Aiming to contribute to this issue, this work evaluated the thermal insulation performance of commercial paints containing muscovite additions and determined the economic evaluation for its industrial production. The thermal paint was formulated by adding 10%, 20% and 40% of muscovite to the commercial paint. This was applied on steel reinforced mortar boards. Thermal insulation tests were carried out in bench scale using an adapted box. The economic evaluation of the industrial production of muscovite-based thermal paint was conducted, considering the Brazilian economic market in this activity. The results showed its ability as an insulating agent due to a reduction of 0.667 °C/mm board by the addition of 40% muscovite. The economic analysis also demonstrated the feasibility of the thermal paint industrial production. The payback is favorable to 5 years when compared to the Selic short-term lending rate, with 21.53% of internal rate return and a net present value of US$ 15,085.76.

  13. Evaluation of Thermal Anomalies in Multi-Boreholes Field Considering the Effects of Groundwater Flow

    OpenAIRE

    Shibin Geng; Yong Li; Xu Han; Huiliang Lian; Hua Zhang

    2016-01-01

    In this paper, the performance of multiple boreholes (multi-BHEs) field is evaluated by considering the groundwater flow. Optimization strategies are presented to mitigate thermal anomalies in the BHEs field. This study shows that groundwater flow greatly improves the heat transfer but causes thermal anomalies downstream. To overcome this problem, a heat transfer model is established for multi-boreholes based on temperature field superposition and moving finite line source model (MFLS). The M...

  14. Using thermal analysis to evaluate the fire effects on organic matter content of Andisols

    OpenAIRE

    J. Neris; Hernández-Moreno, J. M.; C Jiménez; M. Tejedor

    2013-01-01

    Soil organic compounds play a relevant role in aggregate stability and thus, in the susceptibility of soils to erosion. Thermal analysis (N2 and air) and chemical oxidation techniques (dichromate and permanganate oxidation) were used to evaluate the effects of a forest fire on the organic matter of Andisols. Both thermal analysis and chemical methods showed a decrease in the organic matter content and an increase in the recalcitrance of the remaining organic compounds in the burned zones. The...

  15. Evaluation of thermal conductivity of heat-cured acrylic resin mixed with A1203

    Directory of Open Access Journals (Sweden)

    Ebadian B.

    2002-08-01

    Full Text Available One of the most important characteristics of denture base is thermal conductivity. This property has a major role in secretions of salivary glands and their enzymes, taste of the food and gustatory response. Polymethyl methacrylate used in prosthodontics is relatively an insulator. Different materials such as metal fillers and ceramics have been used to solve this problem. The aim of this study was the evaluation of AI2O3 effect on thermal conductivity of heat-cured acrylic resin. Acrylic resin was mixed with AI2O3 in two different weight rates (15 and 20 % of weight. So, group 1 and 2 were divided on this basis. Samples with pure acrylic resin were considered as control group. 18 cylindrical patterns were made in 9x9 mm dimensions and thermocouple wires embedded in each sample to act as conductor. The specimens were put in water with 70±1°C thermal range for 10 minutes. Then, thermal conductivity was measured. The results were analyzed with variance analysis and Dunken test. There was significant difference between thermal conductivity of all groups in all period times. It the first seconds, thermal conductivity in groups 1 and 2 were more than control group. Therefore, for developing of thermal conductivity of acrylic resin, A1203 can be used. Certainly, other characteristic of new resin should be evaluated.

  16. Development and applicability evaluation of frequency response function of structures to fluctuations of thermal stratification

    International Nuclear Information System (INIS)

    The oscillation of a thermal stratification layer can induce thermal fatigue damage on structures with nuclear components. To evaluate the thermal stress induced by thermal stratification oscillation, a frequency response function was developed in our previous research. However, this function does not consider the thickness of the stratified layer. Thus, it is difficult to evaluate the stress generated by actual thermal stratified layers having finite thicknesses with sufficient accuracy. To clarify the effects of layer thickness on induced thermal stress, finite element simulations were conducted under various fluid conditions. As a result, it was clarified that the non-dimensional layer thickness Ht*, which is the ratio of layer thickness to layer oscillation length, can explain the thermal stress response mechanism with layer thickness. Based on the clarified mechanisms, the frequency response function was improved. Applicability of the proposed function to a closed branch pipe of a Light Water Reactor (LWR) and the upper plenum of a pressure vessel of a Fast Breeder Reactor (FBR) was validated through comparison with finite element simulations. (author)

  17. BWR-spent fuel transport and storage with the TN trademark 9/4 and TN trademark 24BH casks

    International Nuclear Information System (INIS)

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks in ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., as Muehleberg Nuclear Power Plant owner, is involved in this process and has selected to store its spent fuel, a new high capacity dual-purpose cask, the TN trademark 24BH. For the transport in a medium size cask, COGEMA LOGISTICS has developed a new cask, the TN trademark 9/4, to replace the NTL9 cask, which performed numerous transports of BWR spent fuel in the past decades. Licensed IAEA 1996, the TN trademark 9/4 is a 40 ton transport cask, for 7 BWR high burn-up spent fuel assemblies. The spent fuel assemblies can be transferred in the ZWILAG hot cell in the TN trademark 24BH cask. The first use of these casks took place in 2003. Ten TN trademark 9/4 transports were performed, and one TN trademark 24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN trademark 24BH high capacity dual purpose cask, the TN trademark 9/4 transport cask and describe in detail their characteristics and possibilities

  18. CASK stabilizes neurexin and links it to liprin-α in a neuronal activity-dependent manner.

    Science.gov (United States)

    LaConte, Leslie E W; Chavan, Vrushali; Liang, Chen; Willis, Jeffery; Schönhense, Eva-Maria; Schoch, Susanne; Mukherjee, Konark

    2016-09-01

    CASK, a MAGUK family protein, is an essential protein present in the presynaptic compartment. CASK's cellular role is unknown, but it interacts with multiple proteins important for synapse formation and function, including neurexin, liprin-α, and Mint1. CASK phosphorylates neurexin in a divalent ion-sensitive manner, although the functional relevance of this activity is unclear. Here we find that liprin-α and Mint1 compete for direct binding to CASK, but neurexin1β eliminates this competition, and all four proteins form a complex. We describe a novel mode of interaction between liprin-α and CASK when CASK is bound to neurexin1β. We show that CASK phosphorylates neurexin, modulating the interaction of liprin-α with the CASK-neurexin1β-Mint1 complex. Thus, CASK creates a regulatory and structural link between the presynaptic adhesion molecule neurexin and active zone organizer, liprin-α. In neuronal culture, CASK appears to regulate the stability of neurexin by linking it with this multi-protein presynaptic active zone complex. PMID:27015872

  19. Evaluation of high temperature superconductive thermal bridges for space borne cryogenic detectors

    Science.gov (United States)

    Scott, Elaine P.

    1996-01-01

    Infrared sensor satellites are used to monitor the conditions in the earth's upper atmosphere. In these systems, the electronic links connecting the cryogenically cooled infrared detectors to the significantly warmer amplification electronics act as thermal bridges and, consequently, the mission lifetimes of the satellites are limited due to cryogenic evaporation. High-temperature superconductor (HTS) materials have been proposed by researchers at the National Aeronautics and Space Administration Langley's Research Center (NASA-LaRC) as an alternative to the currently used manganin wires for electrical connection. The potential for using HTS films as thermal bridges has provided the motivation for the design and the analysis of a spaceflight experiment to evaluate the performance of this superconductive technology in the space environment. The initial efforts were focused on the preliminary design of the experimental system which allows for the quantitative comparison of superconductive leads with manganin leads, and on the thermal conduction modeling of the proposed system. Most of the HTS materials were indicated to be potential replacements for the manganin wires. In the continuation of this multi-year research, the objectives of this study were to evaluate the sources of heat transfer on the thermal bridges that have been neglected in the preliminary conductive model and then to develop a methodology for the estimation of the thermal conductivities of the HTS thermal bridges in space. The Joule heating created by the electrical current through the manganin wires was incorporated as a volumetric heat source into the manganin conductive model. The radiative heat source on the HTS thermal bridges was determined by performing a separate radiant interchange analysis within a high-T(sub c) superconductor housing area. Both heat sources indicated no significant contribution on the cryogenic heat load, which validates the results obtained in the preliminary conduction

  20. Neutronics and dose calculation for prospective spent nuclear fuel cask for Ghana Research Reactor - 1 facility

    International Nuclear Information System (INIS)

    Ghana Research Reactor-1 core is to be converted from highly enrich Uranium (HEU) fuel to low enriched Uranium (LEU) fuel in the near future: a storage cask will be needed to store the HEU fuel. Notwithstanding the core conversion process, It is also important for the facilitv to have a storage cask ready when the fuel is finally spent to temporarily store the fuel until permanent storage is provided. Winfrith Improved Multigroup Scheme-Argonne National Laboratory (WIMS-ANL). Reactor Burnup System (REBUS). Oak Ridge Isotope Generation (ORIGEN2) and Monte Carlo ''N'' Particle (MCNP5) codes have been used to design the cask. WIMS-ANL was used in generating cross sections for the REBUS code which was used in the burnup calculations. The REBUS code was used to estimate the core life time. An estimated core life of approximatcly 750 full-power-equivaicnt-days was obtained for reactor operation of 2hours a day. 4 days a week and 48 weeks in a year. The ORIGIN2 code recorded U-235 burnup weight percent of 2.90% whilst the result from the REBUS3 code was 2.86%. The amount of Pu-239 at the end of the irradiation period was 145 mg which is very low relative to other low power reactors. Isotopic inventory obtained from the ORIGIN2 and REBUS3 runs were used in setting up the MCNP5 input deck for the MCNP5 calculation of the criticality and dose rate. Six cask design options were investigated. The materials for the casks designs were selected based on their attenuation coefficient properties and their high removal cross section properties. The various materials were arranged in no specific order in multilayered casks. The reason for investigating six casks was to look at various arrangements of the cask layers that will optimize effective shielding. The spent nuclear fuel at discharge was used as the radioactivity source during the MCNP simulation. The multilayer cask shield comprise of serpentine concrete of density 5.14 g/cm3 and thickness 21.94cm which

  1. Long term containment performance test for spent fuel transport/storage casks

    International Nuclear Information System (INIS)

    The use of transport/storage cask for spent fuel storage is considered to be rational and economical. Since the storage duration may continue for 40 years or so, the function of sealing radioactive materials in the casks must be reliable for long-term. Long-term containment test of full-scale spent fuel transport/storage cask models have been in progress since 1990 in CRIEPI, Japan. It has been 11 years since it started. The results so far demonstrate and confirm very reliable containment performance of the cask lid structure with metal gaskets. Using the test data it is predicted by Larson-Miller Parameter (LMP) method that the containment system will keep its integrity at least for 40 years. (author)

  2. Regulators Experiences in Licensing and Inspection of Dry Cask Storage Facilities

    International Nuclear Information System (INIS)

    The United States Nuclear Regulatory Commission (NRC), through the combination of a rigorous licensing and inspection program, ensures the safety and security of dry cask storage. NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site-specific licensing and general licensing. In July 1986, the NRC issued the first site-specific license to the Surry Nuclear Power Plant in Virginia authorizing the interim storage of spent fuel in a dry storage cask configuration. Today, there are over 30 ISFSIs currently licensed by the NRC with over 700 loaded dry casks. Current projections identify over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. This paper discusses the NRC licensing and inspection experiences. (authors)

  3. Storage and transportation of spent fuel and high-level waste using dry storage casks

    International Nuclear Information System (INIS)

    This paper describes the REA 2023 dry storage cask which has been designed for on-site storage and transportation of spent fuel and high-level waste. The REA 2023 is the first domestic commercial spent fuel dry storage cask completed for the Department of Energy program for demonstration of methods to improve on site utility fuel storage capacity. A description of the operations required for on-site handling and storage is provided with illustrations and photographs of the fabricated cask. An auxiliary skid is also described which is designed for both on-site handling/storage and transportation. A description of the lifting yoke and transportation impact limiters completes the total system for storage and transportation of spent fuel and high level waste in the REA 2023 casks

  4. Regulatory body experiences in licensing and inspection of dry cask storage facilities

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission (NRC), through a rigorous licensing and inspection programme, ensures the safety and security of dry cask storage. The NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site specific licensing and general licensing. In July 1986, the NRC issued the first site specific licence to the Surry Nuclear Power Plant in Virginia, authorizing the interim storage of spent fuel in a dry storage cask configuration. Presently, there are over 40 ISFSIs licensed by the NRC, with over 800 loaded dry casks. Current projections indicate that there will be over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. The paper discusses the NRC's licensing and inspection experiences. (author)

  5. Regulators experiences in licensing and inspection of dry cask storage facilities

    International Nuclear Information System (INIS)

    The United States Nuclear Regulatory Commission (NRC), through the combination of a rigorous licensing and inspection program, ensures the safety and security of dry cask storage. NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site-specific licensing and general licensing. In July 1986, the NRC issued the first site specific license to the Surry Nuclear Power Plant in Virginia authorizing the interim storage of spent fuel in a dry storage cask configuration. Today, there are over 30 ISFSIs currently licensed by the NRC with over 700 loaded dry casks. Current projections identify over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. This paper discusses the NRC licensing and inspection experiences. (author)

  6. The application of fracture mechanics to the safety assessment of transport casks for radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    BAM is the German responsible authority for the mechanical and thermal design safety assessment of packages for the transport of radioactive materials. The assessment has to cover the brittle fracture safety proof of package components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new ''Guidelines for the Application of Ductile Cast Iron for Transport and Storage Casks for Radioactive Materials''. Based on these guidelines higher stresses than before can become permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof with the help of the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA Advisory Material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will be concluded.

  7. Atomic Oxygen Durability Evaluation of Protected Polymers Using Thermal Energy Plasma Systems

    Science.gov (United States)

    Banks, Bruce A.; Rutledge, Sharon K.; Degroh, Kim K.; Stidham, Curtis R.; Gebauer, Linda; Lamoreaux, Cynthia M.

    1995-01-01

    The durability evaluation of protected polymers intended for use in low Earth orbit (LEO) has necessitated the use of large-area, high-fluence, atomic oxygen exposure systems. Two thermal energy atomic oxygen exposure systems which are frequently used for such evaluations are radio frequency (RF) plasma ashers and electron cyclotron resonance plasma sources. Plasma source testing practices such as ample preparation, effective fluence prediction, atomic oxygen flux determination, erosion measurement, operational considerations, and erosion yield measurements are presented. Issues which influence the prediction of in-space durability based on ground laboratory thermal energy plasma system testing are also addressed.

  8. Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

    International Nuclear Information System (INIS)

    TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.1022 at/cm3 for hydrogen and 9.1020 at/cm3 for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C

  9. Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Abadie, P. [COGEMA Logistics (AREVA Group), Saint-Quentin-en-Yvelines (France)

    2004-07-01

    TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10{sup 22} at/cm{sup 3} for hydrogen and 9.10{sup 20} at/cm{sup 3} for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C.

  10. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed.

  11. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed

  12. Status of cask procurement strategy to satisfy DOE/OCRWM requirements

    International Nuclear Information System (INIS)

    The Nuclear Waste Policy Act requires the development of a safe and efficient system to transport spent nuclear fuel to and within the Federal Waste Management System. This paper describes the DOE/OCRWM strategy to develop and procure a major component of the Transportation System-the transport cask systems. The original initiative to develop high-capacity innovative designs and its current status is described. The follow-on phase to design and procure proven technology cask systems is also discussed

  13. Final report on shipping-cask sabotage source-term investigation

    International Nuclear Information System (INIS)

    A need existed to estimate the source term resulting from a sabotage attack on a spent nuclear fuel shipping cask. An experimental program sponsored by the US NRC and conducted at Battelle's Columbus Laboratories was designed to meet that need. In the program a precision shaped charge was fired through a subscale model cask loaded with segments of spent PWR fuel rods and the radioactive material released was analyzed. This report describes these experiments and presents their results

  14. Spent fuel shipping cask handling capability assessment of 27 selected light water reactors

    International Nuclear Information System (INIS)

    This report presents an assessment of the spent fuel shipping cask handling capabilities of those nuclear plants currently projected to lose full core reserve capability in their spent fuel storage basins in the near future. The purpose of this assessment is to determine which cask types, in the current fleet, each of the selected reactors can handle. The cask handling capability of a nuclear plant depends upon both external and internal conditions at the plant. The availability of a rail spur, the lifting capacity of the crane, the adequacy of clearances in the cask receiving, loading, and decontamination areas and similar factors can limit the types of casks that can be utilized at a particular plant. This report addresses the major facility capabilities used in assessing the types of spent fuel shipping casks that can be handled at each of the 27 selected nuclear plants approaching a critical storage situation. The results of this study cannot be considered to be final and are not intended to be used to force utilities to ship by a particular mode. In addition, many utilities have never shipped spent fuel. Readers are cautioned that the results of this study reflect the current situation at the selected plants and are based on operator perceptions and guidance from NRC related to the control of heavy loads at nuclear power plants. Thus, the cask handling capabilities essentially represent snap-shots in time and could be subject to change as plants further analyze their capabilities, even in the near-term. The results of this assessment indicate that 48% of the selected plants have rail access and 59% are judged to be candidates for overweight truck shipments (with 8 unknowns due to unavailability of verifiable data). Essentially all of the reactors can accommodate existing legal-weight truck casks. 12 references, 1 figure, 4 tables

  15. Operations manual for the Beneficial Uses Shipping System cask. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Bronowski, D.R.; Yoshimura, H.R.

    1993-04-01

    This document is the Operations Manual for the Beneficial Uses Shipping System (BUSS) cask. These operating instructions address requirements; for loading, shipping, and unloading, supplementing general operational information found in the BUSS Safety Analysis Report for Packaging (SARP), SAND 83-0698. Use of the BUSS cask is authorized by Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) for the shipment of special form cesium chloride or strontium flouride capsules.

  16. COBRA-SFS modifications and cask model optimization

    Energy Technology Data Exchange (ETDEWEB)

    Rector, D.R.; Michener, T.E.

    1989-01-01

    Spent-fuel storage systems are complex systems and developing a computational model for one can be a difficult task. The COBRA-SFS computer code provides many capabilities for modeling the details of these systems, but these capabilities can also allow users to specify a more complex model than necessary. This report provides important guidance to users that dramatically reduces the size of the model while maintaining the accuracy of the calculation. A series of model optimization studies was performed, based on the TN-24P spent-fuel storage cask, to determine the optimal model geometry. Expanded modeling capabilities of the code are also described. These include adding fluid shear stress terms and a detailed plenum model. The mathematical models for each code modification are described, along with the associated verification results. 22 refs., 107 figs., 7 tabs.

  17. Marginal overweight operating scenario for DOE's initiative I highway casks

    International Nuclear Information System (INIS)

    This paper assesses the potential transport of high-capacity Initiative I highway casks under development by the Office of Civilian Radioactive Waste Management (OCRWM) as permitted marginal overweight shipments that: exceed a gross vehicle weight (gvw) limit of 80,000, but weight less than 96,000 pounds; follow axle and axle group weight limits adopted by the Surface Transportation Assistance Act (STAA) of 1982; conform to dimensional restrictions to operate on most major highways; and comply with the Federal Bridge Formula. The marginal overweight tractor-trailer would operate in normal open-quotes over-the-roadclose quotes mode and comply with all laws and regulations. The vehicle would have a sleeper berth and two drivers - one to drive while the other provides escort and communications services and accumulates required off-duty time

  18. Comparative evaluation of thermal degradation for biodiesels derived from various feedstocks through transesterification

    International Nuclear Information System (INIS)

    Highlights: • TG–FTIR is employed to evaluate thermal degradation characteristics of biodiesels. • Lower content of unsaturated fatty acid is responsible for higher thermal stability. • Evolved products are alkanes, aldehyde/ketones, ethers, alkenes and CO2. - Abstract: Biodiesel is commonly derived from various feedstocks through transesterification. Since thermal degradation is concerned with a lot of scientific applications, it is essential to evaluate the thermal degradation characteristics for biodiesel. In this study, thermal degradation of biodiesel is investigated through thermogravimetric analysis (TGA) coupled with Fourier transform infrared spectroscopy (FTIR). The fatty acid composition and the characteristic functional groups of biodiesel are detected by gas chromatography (GC) and FTIR in advance. Then, TGA experiments are conducted at linear heating rates from 20 °C min−1 to 40 °C min−1 under nitrogen atmosphere in temperature range from 25 °C to 600 °C. Subsequently, the activation energy, including enthalpy, is determined by Friedman method and Flynn–Wall–Ozawa method and the reaction order is calculated through the Avrami theory. In addition, the pre-exponential factor, the Gibbs free energy, and the entropy are also calculated. Finally, to obtain a comprehensive understanding for thermal degradation of biodiesels, all evolved products are detected by FTIR in real time

  19. Using thermal analysis to evaluate the fire effects on organic matter content of Andisols

    Directory of Open Access Journals (Sweden)

    J. Neris

    2013-09-01

    Full Text Available Soil organic compounds play a relevant role in aggregate stability and thus, in the susceptibility of soils to erosion. Thermal analysis (N2 and air and chemical oxidation techniques (dichromate and permanganate oxidation were used to evaluate the effects of a forest fire on the organic matter of Andisols. Both thermal analysis and chemical methods showed a decrease in the organic matter content and an increase in the recalcitrance of the remaining organic compounds in the burned zones. Thermal analysis indicated an increase in the thermal stability of the organic compounds of fire-affected soils and a lower content of both labile and recalcitrant pools as a consequence of the fire. However, this decrease was relatively higher in the labile pool and lower in the recalcitrant one, indicative of an increase in the recalcitrance of the remaining organic compounds. Apparently, black carbon did not burn under our experimental conditions. Under N2, the results showed a lower labile and a higher recalcitrant and refractory contents in burned and some unburned soils, possibly due to the lower decomposition rate under N2 flux. Thermal analysis using O2 and the chemical techniques showed a positive relation, but noticeable differences in the total amount of the labile pool. Thermal analysis methods provide direct quantitative information useful to characterize the soil organic matter quality and to evaluate the effects of fire on soils.

  20. Thermal performance study and evaluation of comfort temperatures in vernacular buildings of North-East India

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Manoj Kumar; Atreya, S.K. [Instrument Design and Development Centre, Indian Institute of Technology Delhi, New Delhi 110016 (India); Mahapatra, Sadhan [Department of Energy, Tezpur University, Tezpur 784028, Assam (India)

    2010-02-15

    Solar passive techniques are being used in vernacular buildings throughout the world. Researchers have done extensive study on thermal performance of vernacular buildings in the different parts of the world. Vernacular architecture of North-Eastern India represents the principle of climate-responsive architecture, which still lacks experimental validation and quantitative analysis. Thermal comfort not only makes the occupants comfortable but also governs energy consumption in the building. Detailed field studies on thermal performances of typical traditional vernacular dwellings in different bioclimatic zones have been undertaken. This field study includes detailed survey of 150 vernacular dwellings, field tests and thermal sensation vote of 300 occupants on ASHRAE thermal sensation scale. Field test includes measurement of temperature, humidity, illumination level and building design parameters. Thermal performances of these vernacular dwellings were evaluated for winter, pre-summer, summer/monsoon and pre-winter months of the year 2008. This evaluation is based on 'adaptive approach', which is the outcome of the field studies and is now part of ASHRAE standard 55/2004 for predicting comfortable temperature of naturally ventilated buildings. This study also tried to find out the range of comfort temperature in these vernacular buildings for different season of the year. It has been found that these vernacular dwellings perform quite satisfactorily except in the winter months and the occupants feel comfortable in a wider range of temperature. (author)

  1. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, O.G.; Matveev, V.Z.; Morenko, A.I.; Il' kaev, R.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)

    2004-07-01

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks.

  2. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    International Nuclear Information System (INIS)

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks

  3. Use of transportable storage casks in the nuclear waste management system

    International Nuclear Information System (INIS)

    A study was performed to determine the viability of the use of transportable storage casks (TSCs), and other metal casks that are designed primarily for storage but which might be used to ship their stored contents to DOE on a one-time use basis (referred to in this study as storage only casks, or SOCs), in the combined utility/DOE spent fuel management system. The viability of the use of TSCs and SOCs was assessed in terms of the costs and savings involved in their use, the sensitivity of these costs and savings to changes in the capacity and cost of fabrication of the casks, the impacts of variation in cask design features on cost and radiation exposure of personnel, and their prospective use in connection with the transport of defense high level wastes. Estimates were developed of the costs of acquiring and handling of TSCs and SOCs at reactor sites. For comparison purposes, similar costs were developed for the use of concrete storage casks at reactor sites. Estimates of the savings involved to the DOE system as a result of receiving spent fuel in TSCs or SOCs were separately developed. A summary of the results of cost estimates developed in Section 4.0 and the Appendices to this report is shown in Tables 2-1 and 2-2 for instances in which the TSC or SOC were delivered to DOE containing intact fuel assemblies and cans of consolidated fuel, respectively. 2 figs., 14 tabs

  4. Use of transportable storage casks in the nuclear waste management system: Appendices

    International Nuclear Information System (INIS)

    A study was performed to determine the viability of the use of transportable storage casks (TSCs), and other metal casks that are designed primarily for storage but which might be used to ship their stored contents to DOE on a one-time use basis (referred to in this study as storage only casks, or SOCs), in the combined utility/DOE spent fuel management system. The viability of the use of TSCs and SOCs was assessed in terms of the costs and savings involved in their use, the sensitivity of these costs and savings to changes in the capacity and cost of fabrication of the casks, the impacts of variation in cask design features on cost and radiation exposure of personnel, and their prospective use in connection with the transport of defense high level wastes. Estimates were developed of the costs of acquiring and handling of TSCs and SOCs at reactor sites. For comparison purposes, similar costs were developed for the use of concrete storage casks at reactor sites. Estimates of the savings involved to the DOE system as a result of receiving spent fuel in TSCs or SOCs were separately developed. These costs are developed and presented in Volume 2, Appendices A through J

  5. Evaluation of thermal cameras in quality systems according to ISO 9000 or EN 45000 standards

    Science.gov (United States)

    Chrzanowski, Krzysztof

    2001-03-01

    According to the international standards ISO 9001-9004 and EN 45001-45003 the industrial plants and the accreditation laboratories that implemented the quality systems according to these standards are required to evaluate an uncertainty of measurements. Manufacturers of thermal cameras do not offer any data that could enable estimation of measurement uncertainty of these imagers. Difficulties in determining the measurement uncertainty is an important limitation of thermal cameras for applications in the industrial plants and the cooperating accreditation laboratories that have implemented these quality systems. A set of parameters for characterization of commercial thermal cameras, a measuring set, some results of testing of these cameras, a mathematical model of uncertainty, and a software that enables quick calculation of uncertainty of temperature measurements with thermal cameras are presented in this paper.

  6. Evaluation and optimization of non enzymatic browning of “cajuina” during thermal treatment

    Directory of Open Access Journals (Sweden)

    L. F. Damasceno

    2008-06-01

    Full Text Available "Cajuina" is a very popular drink in the Brazilian northeastern region and is produced by clarifying cashew apple juice. To preserve "cajuina" from chancing, the clarified cashew apple juice is submitted to thermal treatment where a desired final color should be obtained. To optimize color formation while maintaining high vitamin C and low 5-hydroxymethylfurfural (5-HMF concentrations the thermal treatment of "cajuina" needs to be studied and the non enzymatic mechanism should be better understood and controlled. In this work the effect of thermal treatment on "cajuina" (clarified cashew apple juice was studied at temperatures from 88°C to 121°C. Changes in color were measured and the variation in vitamin C, 5-hydroxymethylfurfural (5-HMF and sugar content were used to evaluate non enzymatic browning. The kinetic models were used to optimize the thermal treatment to produce "cajuina" with an absorbance at 420 nm of 0.023.

  7. Evaluation of Transverse Thermal Stresses in Composite Plates Based on First-Order Shear Deformation Theory

    Science.gov (United States)

    Rolfes, R.; Noor, A. K.; Sparr, H.

    1998-01-01

    A postprocessing procedure is presented for the evaluation of the transverse thermal stresses in laminated plates. The analytical formulation is based on the first-order shear deformation theory and the plate is discretized by using a single-field displacement finite element model. The procedure is based on neglecting the derivatives of the in-plane forces and the twisting moments, as well as the mixed derivatives of the bending moments, with respect to the in-plane coordinates. The calculated transverse shear stiffnesses reflect the actual stacking sequence of the composite plate. The distributions of the transverse stresses through-the-thickness are evaluated by using only the transverse shear forces and the thermal effects resulting from the finite element analysis. The procedure is implemented into a postprocessing routine which can be easily incorporated into existing commercial finite element codes. Numerical results are presented for four- and ten-layer cross-ply laminates subjected to mechanical and thermal loads.

  8. Interpretation of FAENA and TIFFSS experiments. Comparison of temperature evaluation methods on thermal striping

    Energy Technology Data Exchange (ETDEWEB)

    Kasahara, Naoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Lejeail, Yves [CEA-Cadarache DER/SERSI/LECC (France)

    2000-06-01

    Since thermal striping is a coupled thermohydraulic and thermomechanical phenomenon, sodium mock-up tests were usually required to confirm structural integrity. CEA and JNC have developed evaluation procedures of thermal striping to establish design-by-analysis methodology for this phenomenon. In order to compare and to validate these methods, two benchmark problems were planned under EJCC contract. One of benchmarks provided by CEA is temperature and fatigue evaluation of tubes and plates tests performed with the FAENA facility. Another problem from JNC is the same evaluation of plates tests conducted by the TIFFSS facility. This report describes the results of intercomparison of temperature evaluation methods through application to both FAENA and TIFFFSS experiments. (author)

  9. Evaluating piezo-electric transducer response to thermal shock from in-cylinder pressure data

    NARCIS (Netherlands)

    Rosseel, E.; Sierens, R.; Baert, R.S.G.

    1999-01-01

    One of the major effects limiting the accuracy of piezo-electric transducers for performing in-cyclinder pressure measurements is their sensitivity to the cyclic thermal loading effects of the intermittent combustion process. This paper compares 5 different methods for evaluating the effect of this

  10. Three-Dimensional Numerical Evaluation of Thermal Performance of Uninsulated Wall Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ridouane, El Hassan [National Renewable Energy Lab. (NREL), Golden, CO (United States); Bianchi, Marcus V.A. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2011-11-01

    This study describes a detailed 3D computational fluid dynamics model that evaluates the thermal performance of uninsulated wall assemblies. It accounts for conduction through framing, convection, and radiation and allows for material property variations with temperature. This research was presented at the ASME 2011 International Mechanical Engineering Congress and Exhibition; Denver, Colorado; November 11-17, 2011

  11. Improvement and evaluation of thermal, electrical, sealing and mechanical contacts, and their interface materials

    Science.gov (United States)

    Luo, Xiangcheng

    Material contacts, including thermal, electrical, seating (fluid sealing and electromagnetic sealing) and mechanical (pressure) contacts, together with their interface materials, were, evaluated, and in some cases, improved beyond the state of the art. The evaluation involved the use of thermal, electrical and mechanical methods. For thermal contacts, this work evaluated and improved the heat transfer efficiency between two contacting components by developing various thermal interface pastes. Sodium silicate based thermal pastes (with boron nitride particles as the thermally conductive filler) as well as polyethylene glycol (PEG) based thermal pastes were developed and evaluated. The optimum volume fractions of BN in sodium silicate based pastes and PEG based pastes were 16% and 18% respectively. The contribution of Li+ ions to the thermal contact conductance in the PEG-based paste was confirmed. For electrical contacts, the relationship between the mechanical reliability and electrical reliability of solder/copper and silver-epoxy/copper joints was addressed. Mechanical pull-out testing was conducted on solder/copper and silver-epoxy/copper joints, while the contact electrical resistivity was measured. Cleansing of the copper surface was more effective for the reliability of silver-epoxy/copper joint than that of solder/copper joint. For sealing contacts, this work evaluated flexible graphite as an electromagnetic shielding gasket material. Flexible graphite was found to be at least comparable to conductive filled silicone (the state of the art) in terms of the shielding effectiveness. The conformability of flexible graphite with its mating metal surface under repeated compression was characterized by monitoring the contact electrical resistance, as the conformability is important to both electromagnetic scaling and fluid waling using flexible graphite. For mechanical contacts, this work focused on the correlation of the interface structure (such as elastic

  12. Application of spatial frequency response as a criterion for evaluating thermal imaging camera performance

    Science.gov (United States)

    Lock, Andrew; Amon, Francine

    2008-04-01

    Police, firefighters, and emergency medical personnel are examples of first responders that are utilizing thermal imaging cameras in a very practical way every day. However, few performance metrics have been developed to assist first responders in evaluating the performance of thermal imaging technology. This paper describes one possible metric for evaluating spatial resolution using an application of Spatial Frequency Response (SFR) calculations for thermal imaging. According to ISO 12233, the SFR is defined as the integrated area below the Modulation Transfer Function (MTF) curve derived from the discrete Fourier transform of a camera image representing a knife-edge target. This concept is modified slightly for use as a quantitative analysis of the camera's performance by integrating the area between the MTF curve and the camera's characteristic nonuniformity, or noise floor, determined at room temperature. The resulting value, which is termed the Effective SFR, can then be compared with a spatial resolution value obtained from human perception testing of task specific situations to determine the acceptability of the performance of thermal imaging cameras. The testing procedures described herein are being developed as part of a suite of tests for possible inclusion into a performance standard on thermal imaging cameras for first responders.

  13. Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C., E-mail: amir@cdtn.br, E-mail: aclc@cdtn.br, E-mail: lcdl@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

  14. Evaluation of the thermal safety margin for the KMRR using the hot-spot-factor approach

    International Nuclear Information System (INIS)

    In the thermal-hydraulic (T/H) design of a reactor, several design bases are usually established at the outset and they serve throughout the design process as a set of fundamental criteria on which the adequacy of the thermal-hydraulic design can be judged. Some examples of these are the maximum fuel temperature design bases (DB), departure from nucleate boiling (DNB) BD, and maximum linear heat generation rate DB for pressurized water reactors. The same line of design approach was adopted to design the Korean Multipurpose Research Reactor to be built in the Korean Advanced Energy Research Institute by the end of 1990. The most stringent design basis for T/H design is found to be the onset of nucleate boiling (ONB) margin which is defined as the difference between the ONB wall temperature and the actual wall temperature at the hottest spot in the core. The main objective of this study is to identify what kinds of uncertainties are to be included in the core thermal-margin calculation, to set up a core thermal-margin calculation procedure into which these uncertainty effects are implemented, and to model and quantify their effects on the safety margin. The hot channel factor method for three-dimensional core thermal analysis proposed by Chelemer was used with COBRA-IV-I as the mathematical tool to evaluate the core thermal margin and its sensitivity on these uncertainties

  15. EVALUATION OF FLAT-PLATE PHOTOVOLTAIC THERMAL HYBRID SYSTEMS FOR SOLAR ENERGY UTILIZATION.

    Energy Technology Data Exchange (ETDEWEB)

    ANDREWS,J.W.

    1981-06-01

    The technical and economic attractiveness of combined photovoltaic/thermal (PV/T) solar energy collectors was evaluated. The study was limited to flat-plate collectors since concentrating photovoltaic collectors require active cooling and thus are inherently PV/T collectors, the only decision being whether to use the thermal energy or to dump it. it was also specified at the outset that reduction in required roof area was not to be used as an argument for combining the collection of thermal and electrical energy into one module. Three tests of economic viability were identified, all of which PV/T must pass if it is to be considered a promising alternative: PV/T must prove to be competitive with photovoltaic-only, thermal-only, and side-by-side photovoltaic-plus-thermal collectors and systems. These three tests were applied to systems using low-temperature (unglazed) collectors and to systems using medium-temperature (glazed) collectors in Los Angeles, New York, and Tampa. For photovoltaics, the 1986 DOE cost goals were assumed to have been realized, and for thermal energy collection two technologies were considered: a current technology based on metal and glass, and a future technology based on thin-film plastics. The study showed that for medium-temperature applications PV/T is not an attractive option in any of the locations studied. For low-temperature applications, PV/T appears to be marginally attractive.

  16. Evaluation of the thermal behaviour of different 'local climate zones' in Belgium

    Science.gov (United States)

    Verdonck, Marie-Leen; Demuzere, Matthias; Hooyberghs, Hans; Van Coillie, Frieke

    2016-04-01

    Urban areas are one of the most important human habitats; already 50% of the world's population is living there and this percentage is expected to rise to 70% by 2050. Global warming and the increasing world population will only put more pressure on the living conditions in these habitats. From a thermal comfort point of view it is clear that there is a need for sustainable urban planning that integrates the thermal behaviour of these new developments. To develop sustainable urban planning it is key to know what the influence of a new development will be on the thermal behaviour of the city. Classifying the city according to the local climate zone (LCZ) scheme can provide insights in the thermal behaviour of a city. The WUDAPT LCZ classification framework makes it possible to do so in a spatially explicit manner. This study presents an evaluation of the thermal behaviour of LCZ in three different Belgian cities (Brussels, Antwerp, Ghent) based on modelled air and surface temperature. First LCZ maps were delineated for the three cities. The maps were built based on Landsat and high resolution LiDAR images conform to the WUDAPT LCZ classification framework. Meter- resolution LiDAR images provide useful information on building height and were used to improve the LCZ maps. An accuracy assessment stage was added to confirm the validity of the maps. Secondly, the LCZ maps were used as input data for the URBCLIM model to model air and surface temperature. With the modelling results we characterized the thermal behaviour of every LCZ. In a next step the results for the different cities are compared and the generic character of the WUDAPT LCZ classification framework is evaluated. The main incentive for this study is to investigate whether LCZ maps can be used to foresee the influence of future urban growth scenario's on the thermal comfort in cities in Belgium.

  17. Non-destructive evaluation of thermal aging of cast duplex stainless steel using thermoelectric power measurement

    International Nuclear Information System (INIS)

    Cast duplex stainless steel is frequently used in main coolant pipes of PWR (Pressurized water reactor) type nuclear power plants because of its excellent material strength, toughness and superior corrosion resistance. However, it is known that material deterioration referred to as thermal aging occurs when this material is exposed to temperatures over 300degC. As a result, the material toughness decreases. Therefore, in managing the components made of cast duplex stainless steel, it is necessary to evaluate non-destructively such deterioration. In this study, measurement of thermoelectric power, which is sensitive to micro-structural change in materials, was used for the evaluation of thermal aging. First, we investigated change in mechanical properties (hardness, tensile stress and notch toughness) due to thermal aging in cast duplex stainless steel. Secondly we measured thermoelectric power (TEP) and investigated change in TEP de to thermal aging and the effect of temperature of a specimen on TEP.Then the TEP was compared with the mechanical properties. As a result, TEP increases with aging time and the tendency becomes more remarkable as ferrite content increases. The increase in TEP of a specimen with 21.3% ferrite due to thermal aging (400degC-10000 h) is 0.61 μV/degC. The TEP slightly decreases with temperature of the specimen at a rate of about -0.009 μV/degC2. Finally we found good correlation between the TEP and ductile-brittle transition temperature, the TEP and notch toughness at 325degC. The correlation coefficients are respectively 0.886∼0.957 and -0.890∼ -0.978. Therefore, by using TEP measurement, material deterioration of cast duplex stainless steel due to thermal aging can be evaluated. (author)

  18. Evaluation of antioxidants stability by thermal analysis and its protective effect in heated edible vegetable oil

    Directory of Open Access Journals (Sweden)

    Seme Youssef Reda

    2011-06-01

    Full Text Available In this work, through the use of thermal analysis techniques, the thermal stabilities of some antioxidants were investigated, in order to evaluate their resistance to thermal oxidation in oils, by heating canola vegetable oil, and to suggest that antioxidants would be more appropriate to increase the resistance of vegetable oils in the thermal degradation process in frying. The techniques used were: Thermal Gravimetric (TG and Differential Scanning Calorimetry (DSC analyses, as well as an allusion to a possible protective action of the vegetable oils, based on the thermal oxidation of canola vegetable oil in the laboratory under constant heating at 180 ºC/8 hours for 10 days. The studied antioxidants were: ascorbic acid, sorbic acid, citric acid, sodium erythorbate, BHT (3,5-di-tert-butyl-4-hydroxytoluene, BHA (2, 3-tert-butyl-4-methoxyphenol, TBHQ (tertiary butyl hydroquinone, PG (propyl gallate - described as antioxidants by ANVISA and the FDA; and also the phytic acid antioxidant and the SAIB (sucrose acetate isobutyrate additive, which is used in the food industry, in order to test its behavior as an antioxidant in vegetable oil. The following antioxidants: citric acid, sodium erythorbate, BHA, BHT, TBHQ and sorbic acid decompose at temperatures below 180 ºC, and therefore, have little protective action in vegetable oils undergoing frying processes. The antioxidants below: phytic acid, ascorbic acid and PG, are the most resistant and begin their decomposition processes at temperatures between 180 and 200 ºC. The thermal analytical techniques have also shown that the SAIB antioxidant is the most resistant to oxidative action, and it can be a useful choice in the thermal decomposition prevention of edible oils, improving stability regarding oxidative processes.

  19. Inversion method for defects in depth evaluation and thermal wave imaging

    Institute of Scientific and Technical Information of China (English)

    吕跃凯; 张淑仪; 周庆标

    2001-01-01

    A hybrid Newton-like iterative method and a regularization method are employed to perform the numerical simulations of the defects in depth evaluation and the thermal wave imaging for defects-included solid sample by analysis of the surface photo-thermal signals. A simple and effective data processing method is suggested to improve the reconstructed data. The results of the numerical calculation demonstrate that the algorithm presented in this paper is very effective, and can be used for qualitative and quantitative analyses of homogeneous materials with defects in depth included. It is also proved that the algorithm is stable even with noise disturbance.

  20. FIBWR2 evaluation of fuel thermal limits during density wave oscillaions in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Nik, N.; Rajan, S.R.; Karasulu, M. [New York Power Authority, White Plains, NY (United States)

    1995-09-01

    Analyses were performed to evaluate hydraulic and thermal margin responses of three different BWR fuel designs subjected to the same periodic power/flow oscillations, such as those that might be exhibited during an instability event. The power/flow versus time information from the oscillations was used as a forcing function to calculate the hydraulic response and the MCPR performance of the limiting fuel bundles during the regional oscillations using the analytical code FIBWR2. The results of the calculations were used to determine the thermal margin variation as a function of oscillation magnitude.

  1. The Evaluation of Feasibility of Thermal Energy Storage System at Riga TPP-2

    Directory of Open Access Journals (Sweden)

    Ivanova P.

    2015-12-01

    Full Text Available The installation of thermal energy storage system (TES provides the optimisation of energy source, energy security supply, power plant operation and energy production flexibility. The aim of the present research is to evaluate the feasibility of thermal energy system installation at Riga TPP–2. The six modes were investigated: four for non-heating periods and two for heating periods. Different research methods were used: data statistic processing, data analysis, analogy, forecasting, financial method and correlation and regression method. In the end, the best mode was chosen – the increase of cogeneration unit efficiency during the summer.

  2. The Evaluation of Feasibility of Thermal Energy Storage System at Riga TPP-2

    Science.gov (United States)

    Ivanova, P.; Linkevics, O.; Cers, A.

    2015-12-01

    The installation of thermal energy storage system (TES) provides the optimisation of energy source, energy security supply, power plant operation and energy production flexibility. The aim of the present research is to evaluate the feasibility of thermal energy system installation at Riga TPP-2. The six modes were investigated: four for non-heating periods and two for heating periods. Different research methods were used: data statistic processing, data analysis, analogy, forecasting, financial method and correlation and regression method. In the end, the best mode was chosen - the increase of cogeneration unit efficiency during the summer.

  3. Raman spectroscopic evaluation of meat batter structural changes induced by thermal treatment and salt addition.

    Science.gov (United States)

    Herrero, A M; Carmona, P; López-López, I; Jiménez-Colmenero, F

    2008-08-27

    Raman spectroscopy, texture, proximate composition, and water binding analysis were carried out to evaluate the effect of thermal treatment and/or salt addition to meat batter. For this purpose, different meat batters were elaborated: control meat batter (no salt) and meat batters with low (1.0%) and high (2.5%) NaCl content with and without thermal treatment (70 degrees C/30 min). Increase (P batter. Raman spectroscopy analysis revealed a significant (P batter. In this way, a significant correlation was found between beta-sheets, salt content, hardness, and chewiness in heated samples.

  4. Three-Dimensional Numerical Evaluation of Thermal Performance of Uninsulated Wall Assemblies: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Ridouane, E. H.; Bianchi, M.

    2011-11-01

    This study describes a detailed three-dimensional computational fluid dynamics modeling to evaluate the thermal performance of uninsulated wall assemblies accounting for conduction through framing, convection, and radiation. The model allows for material properties variations with temperature. Parameters that were varied in the study include ambient outdoor temperature and cavity surface emissivity. Understanding the thermal performance of uninsulated wall cavities is essential for accurate prediction of energy use in residential buildings. The results can serve as input for building energy simulation tools for modeling the temperature dependent energy performance of homes with uninsulated walls.

  5. Evaluating Thermal Comfort in a Naturally Conditioned Office in a Temperate Climate Zone

    Directory of Open Access Journals (Sweden)

    Andrés Gallardo

    2016-07-01

    Full Text Available This study aims to determine the optimal approach for evaluating thermal comfort in an office that uses natural ventilation as the main conditioning strategy; the office is located in Quito-Ecuador. The performance of the adaptive model included in CEN Standard EN15251 and the traditional PMV model are compared with reports of thermal environment satisfaction surveys presented simultaneously to all occupants of the office to determine which of the two comfort models is most suitable to evaluate the thermal environment. The results indicate that office occupants have developed some degree of adaptation to the climatic conditions of the city where the office is located (which only demands heating operation, and tend to accept and even prefer lower operative temperatures than those considered optimum by applying the PMV model. This is an indication that occupants of naturally conditioned buildings are usually able to match their comfort temperature to their normal environment. Therefore, the application of the adaptive model included in CEN Standard EN15251 seems like the optimal approach for evaluating thermal comfort in naturally conditioned buildings, because it takes into consideration the adaptive principle that indicates that if a change occurs such as to produce discomfort, people tend to react in ways which restore their comfort.

  6. Fuzzy Evaluation of Thermal Comfort in Naturally Ventilated Residential Buildings in China

    Institute of Scientific and Technical Information of China (English)

    WANG Yi; NIU Tian-cai; LIU Jia-ping; XIAO Yong-qiang

    2009-01-01

    In order to assess the differences between the human body thermal sensation in naturally ventilat-ed space and that in air-conditioned space,the fuzzy evaluation model was adopted in the research of thermal sensation in naturally ventilated space.Based on the questionnaires and field measurements,the membership functions were presented by the statistic of the covering frequency to the fuzzy subset.Dry-bulb temperature was taken as the only independent variable for membership functions.The maximum values of membership grades are all at 0.5 or so, which is a distinction character on thermal comfort of naturally ventihted space.By the cal-culating resultS of membership grades value to different fuzzy evaluation subsets,the Predicted Mean Vote (PMV)was obtained.Furthermore,energy coefficient(Ea) was introduced to calculate the energy consump-tion,and the prediction methods of residential building energy consumption were also discussed.Finally,the importance of evaluation model of thermal sense is shown through the energy consumption prediction in a specificresidential building.

  7. A dynamic tester to evaluate the thermal and moisture behaviour of the surface of textiles.

    Science.gov (United States)

    Li, Wenbin; Xu, Weilin; Wang, Hao; Wang, Xin

    2016-01-01

    The thermal and moisture behaviour of the microclimate of textiles is crucial in determining the physiological comfort of apparel, but it has not been investigated sufficiently due to the lack of particular evaluation techniques. Based on sensing, temperature controlling and wireless communicating technology, a specially designed tester has been developed in this study to evaluate the thermal and moisture behaviour of the surface of textiles in moving status. A temperature acquisition system and a temperature controllable hotplate have been established to test temperature and simulate the heat of human body, respectively. Relative humidity of the surface of fabric in the dynamic process has been successfully tested through sensing. Meanwhile, wireless communication technology was applied to transport the acquired data of temperature and humidity to computer for further processing. Continuous power supply was achieved by intensive contact between an elastic copper plate and copper ring on the rotating shaft. This tester provides the platform to evaluate the thermal and moisture behaviour of textiles. It enables users to conduct a dynamic analysis on the temperature and humidity together with the thermal and moisture transport behaviour of the surface of fabric in moving condition. Development of this tester opens the door of investigation on the micro-climate of textiles in real time service, and eventually benefits the understanding of the sensation comfort and wellbeing of apparel wearers. PMID:26724193

  8. Benchmark problems on thermal striping evaluation of FAENA and TIFFSS sodium experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kasahara, Naoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Lejeail, Yves [CEA-Cadarache DER/SERSI/LECC (France)

    2000-06-01

    Since thermal striping is a coupled thermohydraulic and thermomechanical phenomenon, sodium mock-up tests were usually required to confirm structural integrity. CEA and JNC have developed evaluation procedures of thermal striping to establish design-by-analysis methodology for this phenomenon. Attenuation of temperature and stress amplitude was one of the most important factors in the integrity assessment. Since this attenuation depends on frequencies of temperature fluctuation, benchmark problems based on frequency control tests were planned to confirm above procedures. One of benchmarks provided by CEA is temperature and fatigue evaluation of tubes and plates due to channel flows. Another problem from JNC is the same evaluation of plates subjected to vertical jets. This report explains details of both experiments and defines the benchmark problems. (author)

  9. New Method for Evaluating Thermal Wear of Rolls in Rolling Process

    Institute of Scientific and Technical Information of China (English)

    LI Chang-sheng; LIU Xiang-hua; WANG Guo-dong

    2008-01-01

    A new method was developed by a thermal wear machine to evaluate the thermal wear of rolls in steel rolling process.The steel strip and rolls were simulated by upper and lower heating disks.The upper heating disk could be kept at a temperature of over 900℃ by induction heating.The pressure between the disks as high as 323.2 MPa could be achieved and the slipping rate could be 12.7%.The thermal wear of high speed steel(HSS)roll material,the wear rate of the HSS roll,and the SEM morphology of a worn HSS roll surface were investigated.This method was useful and could be employed to simulate friction and wear between strip and roll during the strip rolling process.

  10. A method for evaluating pressure locking and thermal binding of gate valves

    Energy Technology Data Exchange (ETDEWEB)

    Dogan, T.

    1996-12-01

    A method is described to evaluate the susceptibility of gate valves to pressure locking and thermal binding. Binding of the valve disc in the closed position due to high pressure water trapped in the bonnet cavity (pressure locking) or differential thermal expansion of the disk in the seat (thermal binding) represents a potential mechanism that can prevent safety-related systems from functioning when called upon. The method described here provides a general equation that can be applied to a given gate valve design and set of operating conditions to determine the susceptibility of the valve to fail due to disc binding. The paper is organized into three parts. The first part discusses the physical mechanisms that cause disc binding. The second part describes the mathematical equations. The third part discusses the conclusions.

  11. Thermal Interface Evaluation of Heat Transfer from a Pumped Loop to Titanium-Water Thermosyphons

    Science.gov (United States)

    Jaworske, Donald A.; Sanzi, James L.; Gibson, Marc A.; Sechkar, Edward A.

    2009-01-01

    Titanium-water thermosyphons are being considered for use in the heat rejection system for lunar outpost fission surface power. Key to their use is heat transfer between a closed loop heat source and the heat pipe evaporators. This work describes laboratory testing of several interfaces that were evaluated for their thermal performance characteristics, in the temperature range of 350 to 400 K, utilizing a water closed loop heat source and multiple thermosyphon evaporator geometries. A gas gap calorimeter was used to measure heat flow at steady state. Thermocouples in the closed loop heat source and on the evaporator were used to measure thermal conductance. The interfaces were in two generic categories, those immersed in the water closed loop heat source and those clamped to the water closed loop heat source with differing thermal conductive agents. In general, immersed evaporators showed better overall performance than their clamped counterparts. Selected clamped evaporator geometries offered promise.

  12. High-resolution thermal imaging methodology for non-destructive evaluation of historic structures

    Science.gov (United States)

    Hess, Michael; Vanoni, David; Petrovic, Vid; Kuester, Falko

    2015-11-01

    This paper presents a methodology for automated, portable thermography, for the acquisition of high-resolution thermal image mosaics supporting the non-destructive evaluation of historic structures. The presented approach increases the spatial resolution of thermal surveys to a level of detail needed for building scale analysis. The integration of a robotic camera platform enables automated alignment of multiple images into a high-resolution thermal image mosaic giving a holistic view of the structure while maintaining a level of detail equaling or exceeding that of traditional spot surveys using existing cameras. Providing a digital workflow for automated data and metadata recording increases the consistency and accuracy of surveys regardless of the location or operator. An imaging workflow and instrumentation are shown for a case-study on buildings in Florence, Italy demonstrating the effectiveness of this methodology for structural diagnostics.

  13. Effects of temperature on concrete cask in a dry storage facility for spent nuclear fuels

    International Nuclear Information System (INIS)

    In the dry storage of spent nuclear fuels,concrete cask serves both as a shielding and a structural containment. The concrete in the storage facility is expected to endure the decay heat of the spent nuclear fuel during its service life. Thus, effects of the sustaining high temperature on concrete material need be evaluated for safety of the dry storage facility. In this paper, we report an experimental program aimed at investigating possible high temperature effects on properties of concrete, with emphasis on the mechanical stability, porosity,and crack-resisting ability of concrete mixes prepared using various amounts of Portland cement, fly ash, and blast furnace slag. The experimental results obtained from concrete specimens exposed to a temperature of 94 degree C for 90 days indicate that: (1) compressive strength of the concrete remains practically unchanged; (2) the ultrasonic pulse velocity, and dynamic modulus of elasticity of the concrete decrease in early stage of the high-temperature exposure,and gradually become stable with continuing exposure; (3) shrinkage of concrete mixes exhibits an increase in early stage of the exposure and does not decrease further with time; (4) concrete mixes containing pozzolanic materials,including fly ash and blast furnace slag, show better temperature-resisting characteristics than those using only Portland cement. (authors)

  14. Development of Innovative Aerogel Based Plasters: Preliminary Thermal and Acoustic Performance Evaluation

    Directory of Open Access Journals (Sweden)

    Cinzia Buratti

    2014-09-01

    Full Text Available The thermal and acoustic properties of innovative insulating systems used as building coatings were investigated: Granular silica aerogel was mixed with natural plaster in different percentages. This coating solution is transpiring and insulating, thanks to the use of a natural lime coat and aerogel, a highly porous light material with very low thermal conductivity. The thermal conductivity of the proposed solution was evaluated by means of a Heat Flow meter apparatus (EN ISO 12667, considering different percentages of aerogel. The natural plaster without aerogel has a thermal conductivity of about 0.50 W/m K; considering a percentage of granular aerogel of about 90% in volume, the thermal conductivity of the insulating natural coating falls to 0.050 W/m K. Increasing the percentage of granular aerogel, a value of about 0.018–0.020 W/m K can be reached. The acoustic properties were also evaluated in terms of the acoustic absorption coefficient, measured by means of a Kundt’s Tube (ISO 10534-2. Two samples composed by a plasterboard support, an insulation plaster with aerogel (thicknesses 10 mm and 30 mm respectively and a final coat were assembled. The results showed that the absorption coefficient strongly depends on the final coat, so the aerogel-based plaster layer moderately influences the final value. The application of this innovative solution can be a useful tool for new buildings, but also for the refurbishment of existing ones. This material is in development: until now, the best value of the thermal conductivity obtained from manufacturers is about 0.015 W/m K.

  15. Conceptual Design Report Cask Loadout Sys and Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform at 105 K West

    Energy Technology Data Exchange (ETDEWEB)

    LANGEVIN, A.S.

    1999-07-12

    This conceptual design report documents the redesign of the IPSS and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5/A.6, Canister Transfer Facility Modifications. Project A.5/A.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The function of the IPSS is to suspend, guide, and position the immersion pail. The immersion pail protects the cask-MCO from contamination by basin water and acts as a lifting device for the cask-MCO. The OIP provides operator access to the south loadout pit. Previous analyses studied the effects of a cask-MCO drop on the south loadout pit concrete structure and on the IPSS. The most recent analysis considered the resulting loads at the pit slab/wall joint (Kanjilal, 1999). This area had not been modeled previously, and the analysis results indicate that the demand capacity exceeds the allowable at the slab/wall joint. The energy induced on the south loadout pit must be limited such that the safety class function of the basin is maintained. The solution presented in this CDR redesigns the IPSS and the OIP to include impact-absorbing features that will reduce the induced energy. The impact absorbing features of the new design include: Impact-absorbing material at the IPSS base and at the upper portion of the IPSS legs. A sleeve which provides a hydraulic means of absorbing energy. Designing the OIP to act as an impact absorber. The existing IPSS structure in 105 KW will be removed. This conceptual design considers only loads resulting from drops directly over the IPSS and south loadout pit area. Drops in other areas of the basin are not considered, and will be covered as part of a future revision to this CDR.

  16. Conceptual Design Report - Cask Loadout System Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform, 105 K West Basin, Project A.5/A.6

    International Nuclear Information System (INIS)

    This conceptual design report documents the redesign of the immersion pail support structure (IPSS) and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5lA.6, Canister Transfer Facility Modifications. Project A.5lA.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The junction of the IPSS is to suspend, guide, and position the immersion pail. The immersion pail protects the cask-MCO from contamination by basin water and acts as a lifting device for the cask-MCO. The OIP provides operator access to the south loadout pit. Previous analyses studied the effects of a cask-MCO drop on the south loadout pit concrete structure and on the IPSS. The most recent analysis considered the resulting loads at the pit slab/wall joint (Kanjilal, 1999). This area had not been modeled previously, and the analysis results indicate that the demand capacity exceeds the allowable at the slablwall joint. The energy induced on the south loadout pit must be limited such that the safety class function of the basin is maintained. The solution presented in this CDR redesigns the IPSS and the OIP to include impact-absorbing features that will reduce the induced energy. The impact absorbing features of the new design include: Impact-absorbing material at the IPSS base and at the upper portion of the IPSS legs. A sleeve which provides a hydraulic means of absorbing energy. Designing the OIP to act as an impact absorber. The existing IPSS structure in 105 KW will be removed. This conceptual design considers only loads resulting from drops directly over the IPSS and south loadout pit area. Drops in other areas of the basin are not considered, and will be covered as part of a future revision to this CDR

  17. An Evaluation of Thermal Imaging Based Respiration Rate Monitoring in Children

    Directory of Open Access Journals (Sweden)

    Farah AL-Khalidi

    2011-01-01

    Full Text Available Problem statement: An important indicator of an individual’s health is respiration rate. It is the average number of times air is inhaled and exhaled per minute. Existing respiration monitoring methods require an instrument to be attached to the patient’s body during the recording. This is a discomfort to the patient and the instrument can be dislodged from its position. Approach: In this study a novel noncontact, thermal imaging based respiration rate measurement method is developed and evaluated. Facial thermal videos of 16 children (age: Median = 6.5 years, minimum = 6 months, maximum = 17 years were processed in the study. The recordings were carried out while the children rested comfortably on a bed. The children’s respiration rates were also simultaneously measured using a number of conventional contact based methods. Results: This allowed comparisons with the thermal imaging method to be carried out. The image capture rate was 50 frames per second and the duration of a thermal video recording was 2 min per child. The thermal images were filtered and segmented to identify the nasal region. An algorithm was developed to automatically track the identified nasal area. This region was partitioned into eight equal concentric segments. The pixel values within each segment were averaged to produce a single thermal feature for that segment of the image. A respiration signal was obtained by plotting each segment��€™s feature against time. Conclusion: Respiration rate values were automatically calculated by determining the number of oscillations in the respiration signals per minute. A close correlation (coefficient = 0.994 was observed between the respiration rates measured using the thermal imaging method and those obtained using the most effective conventional contact based respiration method.

  18. Vacuum Ultraviolet Radiation and Atomic Oxygen Durability Evaluation of HST Bi-Stem Thermal Shield Materials

    Science.gov (United States)

    Dever, Joyce; deGroh, Kim K.

    2002-01-01

    Bellows-type thermal shields were used on the bi-stems of replacement solar arrays installed on the Hubble Space Telescope (HST) during the first HST servicing mission (SMI) in December 1993. These thermal shields helped reduce the problem of thermal gradient- induced jitter observed with the original HST solar arrays during orbital thermal cycling and have been in use on HST for eight years. This paper describes ground testing of the candidate solar array bi-stem thermal shield materials including backside aluminized Teflon(R)FEP (fluorinated ethylene propylene) with and without atomic oxygen (AO) and ultraviolet radiation protective surface coatings for durability to AO and combined AO and vacuum ultraviolet (VOV) radiation. NASA Glenn Research Center (GRC) conducted VUV and AO exposures of samples of candidate thermal shield materials at HST operational temperatures and pre- and post-exposure analyses as part of an overall program coordinated by NASA Goddard Space Flight Center (GSFC) to determine the on-orbit durability of these materials. Coating adhesion problems were observed for samples having the AO- and combined AO/UV-protective coatings. Coating lamination occurred with rapid thermal cycling testing which simulated orbital thermal cycling. This lack of adhesion caused production of coating flakes from the material that would have posed a serious risk to HST optics if the coated materials were used for the bi-stem thermal shields. No serious degradation was observed for the uncoated aluminized Teflon(R) as evaluated by optical microscopy, although atomic force microscopy (AFM) microhardness testing revealed that an embrittled surface layer formed on the uncoated Teflon(R) surface due to vacuum ultraviolet radiation exposure. This embrittled layer was not completely removed by AO erosion, No cracks or particle flakes were produced for the embrittled uncoated material upon exposure to VUV and AO at operational temperatures to an equivalent exposure of

  19. Comparative evaluation of different thermally modified wood samples finishing with UV-curable and waterborne coatings

    Science.gov (United States)

    Herrera, René; Muszyńska, Monika; Krystofiak, Tomasz; Labidi, Jalel

    2015-12-01

    Thermally modified wood has been developed as an industrial method to improve durability and dimensional stability of wood and thus extends the range of uses and service life of wood-based products. Despite the improvements gained by treatment, surface finishing using coatings prevents esthetical changes such as color degradation or occasional growth of mold adding protection in outdoor use and extending the service life of products. The wood finishing process was carried out with commercially available waterborne and UV-curable coatings on industrially modified at 192, 200, 212 °C and unmodified European ash (Fraxinus excelsior L.) wood, using an industrial rollers system and a laboratory brushing system. Changes caused by thermal treatment which could affect the surface finish were measured and compared with control samples, such as water uptake, wettability and acidity. Following the wood finishing, surface properties and esthetic changes were evaluated; as well as the coatings performance. Thermally modified wood presented improved adherence compared with unmodified wood with a significant improvement in samples modified at 212 °C, which also present the highest hardness when UV-cured. Finishes with UV-curing maintain the hydrophobic effect of thermally modified wood, whereas waterborne finishes increase the surface wettability. Thermal modification did not negatively influence on the elastic properties of the coated substrate and thus allows this material to be finished with different coating systems in the same conditions as unmodified wood.

  20. Evaluation on numerical simulation accuracy of the commercial CFD program for FBR thermal-hydraulic conditions and applications. Single phase multi-dimensional thermal-hydraulic evaluation problems

    International Nuclear Information System (INIS)

    Commercial computational fluid dynamic program is taken up to be employed for nuclear thermal-hydraulic applications due to the advantages in high-speed solution and easy-to-use operation. The principal objective of this report is evaluating the numerical simulation accuracy of the Fluent, on single-phase multi-dimensional thermal hydraulic problems. The evaluation problems are: 1) Laminar flow over a backward-facing step, 2) Turbulent flow over a backward-facing step, 3) Temperature of a inner rectangular rotating flow, 4) Thermal-driven natural convection flow in a square cavity, and 5) Turbulent flow in a cubic cavity, those were selected in supposing nuclear reactor thermal-hydraulic conditions by the technical committee of the Japan atomic energy society. The features on numerical method and accuracy of the Fluent being identified are: 1) Spatial differential schemes for convection term: 1st upwind, power-law, 2nd upwind, and Quick, upgrade the numerical accuracy in this order. Each scheme has the same accuracy as of the existing referenced numerical results. Quick scheme employs numerical stability oriented filtering so that no over- or under-shoots are observed. Yet, 2nd central differential scheme -used in large eddy simulation (LES)- leads numerical instability (i.e. temporal oscillation in pressure, and spatial wavering in velocity) typically when we deal with in low-resolution domains. 2) Turbulent models: (Standard, RNG, Realizable) k-ε, (Standard, SST) k-ω, and, (Standard, Quadratic) RST, necessitate to involve non-equilibrium wall function to take numerical accuracy and stability. The Fluent evaluations on re-attaching points and velocity distributions show nearly the same as -and on several counts more accurate than- those of the existing reference results. The LES turbulent model can be used only for 3-D simulations. 3) The evaluations of thermal-driven natural convection flow, which is one of the heat transfer and fluidics coupling problem, show

  1. Integrated cask storage systems for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Since 1979 Tennessee Valley Authority TVA has participated in conceptual design studies of dry storage vaults, silos, casks, ad dry wells, and, with DOE and others, has undertaken limited demonstrations of rod consolidation and cask dry storage at TVA's Browns Ferry Nuclear Plant in Alabama. TVA believes the integrated storage cask concept is worthy of consideration as an alternative for spent fuel management. Placing spent fuel in a secure passive storage mode at an early date and avoiding unnecessary handling and repackaging reduces the potential for occupational and public radiological exposure. Therefore the notion of a universal cask used for storage, shipment, and disposal is appealing from a safety, environmental, and public perception standpoint. The universal cask can also serve as a dispersed monitored retrievable storage (MRS), thus eliminating the need for redundant facilities, and it does not foreclose future options. It also appears that this concept would simplify repository design, ease retrievability, and provide greater flexibility in repository siting. 2 figures, 2 tables

  2. An attempt to estimate gamma-ray dose rate from radioactive waste storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M.; Plecas, I.; Pavlovic, R.; Pavlovic, S. [Institute for Nuclear Sciences ' ' Vinca' ' , Belgrade (Yugoslavia); Sokcic-Kostic, M. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Hauptabteilung Zyklotron

    2001-07-01

    Radioactive waste - {sup 137}Cs and {sup 60}Co contaminated sludge from the irradiated fuel storage pool of the RA research reactor in the Vinca Institute of Nuclear Sciences, Belgrade, is conditioned and stored in specially designed casks in 1997. Main purpose of this paper is to describe an attempt to estimate a gamma-ray ambient dose equivalent rate from the cask with the conditioned sludge by reference Monte Carlo MCNP code and compare the result to the measuring data. The aim of the study is to master with a reliable computational tool that allows faithful estimation of the total ambient gamma-ray dose equivalent rates from the cask. Values of so obtained gamma-ray ambient dose equivalent rates are compared to the measured values at the same spots of the cask and acceptable agreement were found. These data could be used in a further study on minimisation of the total ambient dose equivalent rate of a regular array or random pile of casks, stored in the storage location. (orig.)

  3. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.

  4. Behavior of Full-Scale Model Cask Under 9 m Drop Test and Simulation

    International Nuclear Information System (INIS)

    The nuclear spent fuel transport and storage cask is used for transport of the spent fuel from a nuclear power station to an intermediate storage facility. Leak tightness and subcriticality on transportation required from IAEA TS-R1[1] have to be assured by a 9 m drop test and its numerical simulation. This paper describes the drop test using a full-scale prototype test cask The test was conducted by German Federal Institute for Materials Research and Testing (BAM) at their test facility in Horstwalde, Germany and comparison of the test result with the 'MH1 (Mitsubishi Heavy Industries, Ltd.)' numerical simulation using LS-DYNA code. The drop orientations of the tests were slap down and vertical. From the drop test the following is demonstrated: The leak rate of He gas after the drop tests satisfied the IAEA's criteria. The numerical simulation which modeled the cask body enabled dynamic response such as acceleration and strain of the cask body. This means the simulation method qualified the relation of dynamic response of the cask body and leakage behavior. (authors)

  5. Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gardner, Levi D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Huber, Tanja K. [Technische Universität München, Munich (Germany); Breitkreutz, Harald [Technische Universität München, Munich (Germany)

    2015-02-11

    The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

  6. Preliminary survey and evaluation of nonaquifer thermal energy storage concepts for seasonal storage

    Energy Technology Data Exchange (ETDEWEB)

    Blahnik, D.E.

    1980-11-01

    Thermal energy storage enables the capture and retention of heat energy (or cold) during one time period for use during another. Seasonal thermal energy storage (STES) involves a period of months between the input and recovery of energy. The purpose of this study was to make a preliminary investigation and evaluation of potential nonaquifer STES systems. Current literature was surveyed to determine the state of the art of thermal energy storage (TES) systems such as hot water pond storage, hot rock storage, cool ice storage, and other more sophisticated concepts which might have potential for future STES programs. The main energy sources for TES principally waste heat, and the main uses of the stored thermal energy, i.e., heating, cooling, and steam generation are described. This report reviews the development of sensible, latent, and thermochemical TES technologies, presents a preliminary evaluation of the TES methods most applicable to seasonal storage uses, outlines preliminary conclusions drawn from the review of current TES literature, and recommends further research based on these conclusions. A bibliography of the nonaquifer STES literature review, and examples of 53 different TES concepts drawn from the literature are provided. (LCL)

  7. Evaluation Framework and Analyses for Thermal Energy Storage Integrated with Packaged Air Conditioning

    Energy Technology Data Exchange (ETDEWEB)

    Kung, F.; Deru, M.; Bonnema, E.

    2013-10-01

    Few third-party guidance documents or tools are available for evaluating thermal energy storage (TES) integrated with packaged air conditioning (AC), as this type of TES is relatively new compared to TES integrated with chillers or hot water systems. To address this gap, researchers at the National Renewable Energy Laboratory conducted a project to improve the ability of potential technology adopters to evaluate TES technologies. Major project outcomes included: development of an evaluation framework to describe key metrics, methodologies, and issues to consider when assessing the performance of TES systems integrated with packaged AC; application of multiple concepts from the evaluation framework to analyze performance data from four demonstration sites; and production of a new simulation capability that enables modeling of TES integrated with packaged AC in EnergyPlus. This report includes the evaluation framework and analysis results from the project.

  8. An analysis of contingencies for making casks available for use during the early years of federal waste management system operations

    International Nuclear Information System (INIS)

    This paper reports on a study which has been performed to examine the contingencies that could be pursued by the Department of energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) for shipping spent fuel beginning in 1998. OCRWM's current plan is to initiate operations using early production units of Initiative I truck and rail/barge casks that are presently being designed. Contingencies to this plan were considered in case some unforeseen event occurs that precludes the Initiative I casks from entering into service early in 1998 in sufficient quantities (both numbers and types) to satisfy DOE's shipping needs. Specifically, the study addressed the potential availability of cask systems, selected several cask usage scenarios, determined the requirements for casks under these scenarios, generically assessed different strategies for acquiring casks or the use of casks, and generically assessed cask fabrication capabilities. Issues concerning both domestic and foreign resources were addressed with a focus on the first five years of Federal Waste Management System (FWMS) operation

  9. Nondestructive Evaluation of Thermal Spray Coating Interface Quality by Eddy Current Method

    Science.gov (United States)

    Mi, Bao; Zhao, Xiaoliang (George); Bayles, Robert

    2007-03-01

    Thermal spray coating is usually applied through directing molten or softened particles at very high velocities onto a substrate. An eddy current non-destructive inspection technique is presented here for thermal spray coating interface quality characterization. Several high-velocity-oxy-fuel (HVOF) coated steel plates were produced with various surface preparation conditions or spray process parameters. A quad-frequency eddy current probe was used to manually scan over the coating surface to evaluate the bonding quality. Experimental results show that different surface preparation conditions and varied process parameters can be successfully differentiated by the impedance value observed from the eddy current probe. The measurement is fairly robust and consistent. This non-contact, nondestructive, easy-to-use technique has the potential for evaluating the coating quality immediately after its application so that any defects can be corrected immediately.

  10. Preliminary study of small scale solar test cells for solar thermal evaluation of building components

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, G. [CIEMAT, Renewable Energy Dept., Madrid (Spain)]|[CENIDET, Mechanical Engineering Dept., Cuernavaca, Morelos (Mexico); Jimenez, M.J.; Heras, M.R. [CIEMAT, Renewable Energy Dept., Madrid (Spain)

    2004-07-01

    This paper presents a preliminary study to validate small scale solar test cells for thermal evaluation of building components such as windows and roofs. The description and performance of the scale test cells are described. The validation of the thermal performance was made with the real test cell Passys indirectly by applying the classical averaging method that was initially used for the Passys cells to determine overall heat loss coefficient, UA, and the solar heat gain, gA. The use of this methodology was selected, as a first approach, to evaluate the viability of the study of scale test cells. Our preliminary results indicate that some percentage differences were high for some tests. Therefore, it is necessary to increase the period of time of the measurements of the scale test cells, in order to use dynamic system methods to reduce the percentage differences of UA and gA. (orig.)

  11. An experimental and theoretical evaluation of increased thermal diffusivity phase change devices

    Science.gov (United States)

    White, S. P.; Golden, J. O.; Stermole, F. J.

    1972-01-01

    This study was to experimentally evaluate and mathematically model the performance of phase change thermal control devices containing high thermal conductivity metal matrices. Three aluminum honeycomb filters were evaluated at five different heat flux levels using n-oct-adecane as the test material. The system was mathematically modeled by approximating the partial differential equations with a three-dimensional implicit alternating direction technique. The mathematical model predicts the system quite well. All of the phase change times are predicted. The heating of solid phase is predicted exactly while there is some variation between theoretical and experimental results in the liquid phase. This variation in the liquid phase could be accounted for by the fact that there are some heat losses in the cell and there could be some convection in the experimental system.

  12. Evaluation of the effects of vegetation and green walls on building thermal performance and energy consumption

    Science.gov (United States)

    Susorova, Irina

    This research explored the use of vegetation in building facades as a potential solution to the problems of urban ecology and the excessive energy consumption in buildings. Vegetated facades substantially reduce building energy use, reduce the urban heat island effect, improve air quality, and increase the biodiversity of plants and animals in cities. The goal of this research was to evaluate the effects of plants on building thermal performance and energy consumption by developing a thermal model of a building facade covered with a layer of plants. The developed mathematical model accounts for thermal physical processes in a vegetated exterior wall including solar radiation, infrared radiative exchange between the facade and sky, the facade and ground, the facade and vegetation layer, convection to and from the facade, evapotranspiration from the plant layer, heat storage in the facade material, and heat conduction through the facade. The model calculates vegetated facade surface temperature and heat flux through the facade for multiple weather conditions, plant physiological characteristics, and facade parameters inputs. The model was validated with the results of a one-week long experiment measuring the thermal properties of bare and vegetated facades on the Illinois Institute of Technology campus. The experiment and subsequent sensitivity analysis demonstrated that a plant layer can effectively reduce the facade exterior surface temperature, daily temperature fluctuations, exterior wall temperature gradient, and, as a result, provide as much additional thermal insulation to the facade as a 2.5 cm layer of expanded polystyrene insulation. The vegetated facade model was also used to analyze the reduction in energy consumption in generic office and residential thermal zones for multiple parameters. The simulations showed that energy reduction could be as high as 6.2% of annual total energy use and 34.6% of cooling energy use in residential thermal zones. Overall

  13. Synthesis and evaluation of thermal, electrochemical and nonlinear optical properties of dicyano- and tricyanovinyl-bithienylpyrroles

    OpenAIRE

    Fonseca, A. Maurício C.; Castro, M. Cidália R.; Belsley, M.; Raposo, M. Manuela M.

    2012-01-01

    Materials exhibiting large optical nonlinearities continue to be the subject of intense studies because of their potential applications in modern communication technologies involving optical data processing, transmission, or storage, where they are gradually replacing classical electronic devices.[1] Earlier we reported on the synthesis and evaluation of thermal, redox and NLO properties of donor-acceptor substituted thienylpyrrole derivatives. Our studies have shown that replacing the ben...

  14. Evaluation of Atomic, Physical and Thermal Properties of Tellurium Powder: Impact of Biofield Energy Treatment

    OpenAIRE

    Trivedi, Mahendra Kumar

    2015-01-01

    Tellurium has gained significant attention due to its photoconductivity, piezoelectricity, and thermo conductivity properties. The aim of this study was to evaluate the effect of biofield energy treatment on thermal, physical and atomic properties of tellurium powder. The tellurium powder was equally divided in two parts: control and treated (T). The treated part was subjected to Mr. Trivedi’s biofield energy treatment, whereas the control part was remained untreated. Subsequently, the ...

  15. Evaluation of Biofield Treatment on Atomic and Thermal Properties of Ethanol

    OpenAIRE

    Trivedi, Mahendra Kumar

    2015-01-01

    Ethanol is a polar organic solvent, and frequently used as a fuel in automobile industries, principally as an additive with gasoline due to its higher octane rating. It is generally produced from biomass such as corn, sugar and some other agriculture products. In the present study, impact of biofield treatment on ethanol was evaluated with respect to its atomic and thermal properties. The ethanol sample was divided into two parts i.e., control and treatment. Control part was remained untreate...

  16. Evaluation of multielement catheter-cooled interstitial ultrasound applicators for high-temperature thermal therapy

    International Nuclear Information System (INIS)

    Catheter-cooled (CC) interstitial ultrasound applicators were evaluated for their use in high-temperature coagulative thermal therapy of tissue. Studies in ex vivo beef muscle were conducted to determine the influences of applied electrical power levels (5-20 W per element), catheter flow rate (20-60 ml min-1), circulating water temperature (7-40 deg. C), and frequency (7-9 MHz) on temperature distribution and thermal lesion geometry. The feasibility of using multiple interstitial applicators to thermally coagulate a predetermined volume of tissue was also investigated. Results of these studies revealed that the directional shape of the thermal lesions is maintained with increasing time and power. Radial depths of the thermal lesions ranged from 10.7±0.7 mm after heating for 4 min with an applied power level of 5 W, to 16.2±1.4 mm with 20 W. The axial length of the thermal lesions is controlled tightly by the number of active transducers. A catheter flow rate of 20 to 40 ml min-1 (52.2±5.5 kPa at 40 ml min-1) with 22 deg. C water was determined to provide sufficient cooling of the transducers for power levels used in this study. In vivo temperatures measured in the center of a 3-cm-diam peripheral implant of four applicators in pig thigh muscle reached 89.3 deg. C after 4 min of heating, with boundaries of coagulation clearly defined by applicator position and directivity. Conformability of heating in a clinically relevant model was demonstrated by inserting two directional CC applicators with a 2 cm separation within an in vivo canine prostate, and generating a thermal lesion measuring 3.8 cmx2.2 cm in cross section while directing energy away from, and protecting the rectum. Maximum measured temperatures at midgland exceeded 90 deg. C within 20 min of heating. The results of this study demonstrate the utility of single or multiple CC applicators for conformal thermal coagulation and high temperature thermal therapy, with potential for clinical applications in

  17. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    Energy Technology Data Exchange (ETDEWEB)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  18. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L. (Sandia National Labs., Albuquerque, NM (United States)); Jordan, H. (EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant); Pasupathi, V. (Battelle, Columbus, OH (United States)); Mings, W.J. (USDOE Idaho Field Office, Idaho Falls, ID (United States)); Reardon, P.C. (GRAM, Inc., Albuquerque, NM (United States))

    1991-09-01

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs.

  19. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    International Nuclear Information System (INIS)

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs

  20. Neutron measurements around storage casks containing spent fuel and vitrified high-level radioactive waste at ZWILAG.

    Science.gov (United States)

    Buchillier, T; Aroua, A; Bochud, F O

    2007-01-01

    Spectrometric and dosimetric measurements were made around a cask containing spent fuel and a cask containing high-level radioactive waste at the Swiss intermediate waste and spent fuel storage facility. A Bonner sphere spectrometer, an LB 6411 neutron monitor and an Automess Szintomat 6134A were used to characterise the n-gamma fields at several locations around the two casks. The results of these measurements show that the neutron fluence spectra around the cask containing radioactive waste are harder and higher in intensity than those measured in the vicinity of the spent fuel cask. The ambient dose equivalents measured with the LB 6411 neutron monitor are in good agreement with those obtained using the Bonner spheres, except for locations with soft neutron spectra where the monitor overestimates the neutron ambient dose equivalent by almost 50%. PMID:17494980

  1. Evaluation of Silicon Neutron Resonance Parameters in the Energy Range Thermal to 1800 keV

    Energy Technology Data Exchange (ETDEWEB)

    Derrien, H.

    2002-09-30

    The evaluation of the neutron cross sections of the three stable isotopes of silicon in the energy range thermal to 20 MeV was performed by Hetrick et al. for ENDF/B-VI (Evaluated Nuclear Data File). Resonance parameters were obtained in the energy range thermal to 1500 keV from a SAMMY analysis of the Oak Ridge National Laboratory experimental neutron transmission data. A new measurement of the capture cross section of natural silicon in the energy range 1 to 700 keV has recently been performed at the Oak Ridge Electron Linear Accelerator. Results of this measurement were used in a SAMMY reevaluation of the resonance parameters, allowing determination of the capture width of a large number of resonances. The experimental data base is described; properties of the resonance parameters are given. For the first time the direct neutron capture component has been taken into account from the calculation by Rauscher et al. in the energy range from thermal to 1 MeV. Results of benchmark calculations are also given. The new evaluation is available in the ENDF/B-VI format.

  2. Post Irradiation Evaluation of Thermal Control Coatings and Solid Lubricants to Support Fission Surface Power Systems

    Science.gov (United States)

    Bowman, Cheryl L.; Jaworske, Donald A.; Stanford, Malcolm K.; Persinger, Justin A.; Khorsandi, Behrooz; Blue, Thomas E.

    2007-01-01

    The development of a nuclear power system for space missions, such as the Jupiter Icy Moons Orbiter or a lunar outpost, requires substantially more compact reactor design than conventional terrestrial systems. In order to minimize shielding requirements and hence system weight, the radiation tolerance of component materials within the power conversion and heat rejection systems must be defined. Two classes of coatings, thermal control paints and solid lubricants, were identified as material systems for which limited radiation hardness information was available. Screening studies were designed to explore candidate coatings under a predominately fast neutron spectrum. The Ohio State Research Reactor Facility staff performed irradiation in a well characterized, mixed energy spectrum and performed post irradiation analysis of representative coatings for thermal control and solid lubricant applications. Thermal control paints were evaluated for 1 MeV equivalent fluences from 10(exp 13) to 10(exp 15) n per square centimeters. No optical degradation was noted although some adhesive degradation was found at higher fluence levels. Solid lubricant coatings were evaluated for 1 MeV equivalent fluences from 10(exp 15) to 10(exp 16) n per square centimeters with coating adhesion and flexibility used for post irradiation evaluation screening. The exposures studied did not lead to obvious property degradation indicating the coatings would have survived the radiation environment for the previously proposed Jupiter mission. The results are also applicable to space power development programs such as fission surface power for future lunar and Mars missions.

  3. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    Science.gov (United States)

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

  4. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  5. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  6. Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L [ed.

    1985-09-01

    An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures.

  7. Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures

  8. Nuclear waste management systems issues related to transportation cask design: At-reactor spent fuel storage, Monitored Retrievable Storage and modal mix

    International Nuclear Information System (INIS)

    This report provides background information on nuclear waste transportation issues for an upcoming review of waste shipping cask designs. The focus of this report is related issues pertaining to at-reactor storage, monitored retrievable storage, and the mix of spent fuel transportation modes (railroad, highway and waterways) that will determine impacts of spent fuel transportation to a geologic repository. Part 1 traces the evolution of the civilian radioactive waste management program from its inception through passage of the 1987 amendments to the Nuclear Waste Policy Act (NWPA) of 1982. It emphasizes the factors that will influence the configuration of the transportation system for high-level nuclear waste and related cask design. Part 2 deals with at-reactor storage of wastes. Options for at-reactor storage of waste include expanding pool storage, consolidation and compaction of wastes in pool storage, and various forms of dry storage. Storage needs at-reactor are estimated, and storage options are evaluated on the basis of their ability to meet those needs. Part 3 deals with the MRS facility. The status of the MRS is reviewed starting with the 1987 Nuclear Waste Policy Amendments Act. Studies of the MRS by the State of Tennessee and DOE are reviewed. Alternatives to the MRS, such as the Integrated No-MRS waste management system configuration are discussed. The activities of the MRS Review Commission are also reviewed. Part 4 deals with transportation of wastes from reactors to the MRS or final disposal facility. Road, rail and water transport are evaluated, as are mixtures of different modes. The implications of transportation mode on cask design are discussed, as is the potential for transportation system optimization. The last section applies the as-low-as-reasonably-achievable (ALARA) standard to the problem of radioactive waste transportation. 106 refs., 14 figs., 14 tabs

  9. Evaluation of Surface Modification as a Lunar Dust Mitigation Strategy for Thermal Control Surfaces

    Science.gov (United States)

    Gaier, James R.; Waters, Deborah L.; Misconin, Robert M.; Banks, Bruce A.; Crowder, Mark

    2011-01-01

    Three surface treatments were evaluated for their ability to lower the adhesion between lunar simulant dust and AZ93, AlFEP, and AgFEP thermal control surfaces under simulated lunar conditions. Samples were dusted in situ and exposed to a standardized puff of nitrogen gas. Thermal performance before dusting, after dusting, and after part of the dust was removed by the puff of gas, were compared to perform the assessment. None of the surface treatments was found to significantly affect the adhesion of lunar simulants to AZ93 thermal control paint. Oxygen ion beam texturing also did not lower the adhesion of lunar simulant dust to AlFEP or AgFEP. But a workfunction matching coating and a proprietary Ball Aerospace surface treatment were both found to significantly lower the adhesion of lunar simulants to AlFEP and AgFEP. Based on these results, it is recommended that all these two techniques be further explored as dust mitigation coatings for AlFEP and AgFEP thermal control surfaces.

  10. TMI-2 (Three-Mile Island-Unit 2) rail cask and railcar maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Tyacke, M.J.; Ayers, A.L., Jr.; Ball, L.J.; Anselmo, A.A.

    1988-02-01

    This paper describes the NuPac 125-B cask system (i.e., cask and railcar), and the maintenance and inspection requirements for that system. The paper discusses the operations being done to satisfy those requirements and how, in some cases, it has been efficient for the operations to be more rigorous than the requirements. Finally, this paper discusses the experiences gained from those operations and how specific hardware and procedural enhancements have resulted in a reliable and continuous shipping campaign. 2 refs., 2 figs.

  11. Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads due to Aircraft Engine Crash

    Science.gov (United States)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters(1) and seismic tests subjected to strong earthquake motions(2). Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001(3)-(6). This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are

  12. IEA SHC Task 42/ECES Annex 29 – A Simple Tool for the Economic Evaluation of Thermal Energy Storages

    DEFF Research Database (Denmark)

    Rathgeber, Christoph; Hiebler, Stefan; Lävemann, Eberhard;

    2016-01-01

    Within the framework of IEA SHC Task 42 / ECES Annex 29, a simple tool for the economic evaluation of thermal energy storages has been developed and tested on various existing storages. On that account, the storage capacity costs (costs per installed storage capacity) of thermal energy storages......, seasonal heat storage is only economical via large sensible hot water storages. Contrary, if the annual number of storage cycles is sufficiently high, all thermal energy storage technologies can become competitive....

  13. Evaluation of the Thermal Response of the 5-DHLW Waste Package-Hypothetical Fire Accident

    International Nuclear Information System (INIS)

    The purpose of this calculation is to determine the thermal response of the 5-defense high level waste (DHLW)/Department of Energy (DOE) codisposal waste package (WP) to the hypothetical fire accident. The objective is to calculate the temperature response of the DHLW glass to the hypothetical short-term fire defined in 10 CFR 71, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. The scope is intended to cover a DHLW WP. This WP is loaded with DHLW canisters containing glass from the Savannah River Site (SRS) and a DOE canister containing Training, Research, and Isotope General Atomics (TRIGA) spent nuclear fuel (SNF). The information provided by the sketches attached to this calculation is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the central DOE canister and DHLW thermal loads are determined. Also, the effects of varying values of the flame and WP outer surface emissivities are evaluated

  14. Evaluation of the litcit software for thermal simulation of superficial lasers such as hair removal lasers

    Directory of Open Access Journals (Sweden)

    Shirkavand A

    2007-01-01

    Full Text Available Background and Objectives : In this study, we evaluate LITCIT software for its application as a thermal simulation software for superficial hair removal laser systems. Materials and Methods: Two articles were used as our references. Complete information regarding the tissues, such as optical/thermal properties and geometrical modeling and also the laser systems such as wavelength, spot size, pulse duration and fluence were extracted from these texts. Then, this information regarding the tissues and systems was entered into the LITCIT simulation software. Further, we ran the program and saved the results. Finally, we compared our results with the results in references and evaluated the. Results : Output results of the LITCIT show that they are consistent with the results of references that were calculated with a different thermal modeling. Such a small average error shows the accuracy of the software for simulation and calculating the temperature. Conclusions : This simulating software has a good ability to be used as a treatment planning software for superficial lasers. Thus, it can be used for the optimization of treatment parameters and protocols.

  15. Thermal Manikin Evaluation of Liquid Cooling Garments Intended for Use in Hazardous Waste Management

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, J. P.; Semeniuk, K.; Makris, A.; Teal, W.; Laprise, B.

    2003-02-26

    Thermal manikins are valuable tools for quantitatively evaluating the performance of protective clothing ensembles and microclimate cooling systems. The goal of this investigation was to examine the performance of Coretech personal cooling systems, designed to reduce the effects of physiological and environmental heat stress, using a sweating thermal manikin. A sweating manikin takes into account the effective physiological evaporative heat transfer. Three tubesuits containing different densities of tubing were evaluated on the thermal manikin in conjunction with body armor and two Chemical-Biological suits (SPM and JSLIST). The experiments were carried out in an environmental chamber set at a temperature of 35 C with a relative humidity of 30%. For the tubesuits, two flow rates were tested and the heat removal rates were obtained by measuring the amount of power required to maintain the manikin's surface at a constant temperature of 35 C. The sweating rates were adjusted to maintain a fully wetted manikin surface at the above environmental conditions. For fluid flow rates ranging from approximately 250 to 750 ml/min, and inlet temperatures to the tubesuit ranging from 7 to 10 C, heat removal rates between 220 W to 284 W were measured, indicating the effectiveness of tubesuits at removing excessive body heat. This research was performed at the U.S. Army Soldier and Biological Chemical Command (SBCCOM) in Natick, Massachusetts.

  16. Evaluation of thermal water in patients with functional dyspepsia and irritable bowel syndrome accompanying constipation

    Institute of Scientific and Technical Information of China (English)

    Giovanni Gasbarrini; Marcello Candelli; RiccardoGiuseppe Graziosetto; Sergio Coccheri; Ferdinando Di Iorio; Giuseppe Nappi

    2006-01-01

    AIM: To evaluate the efficacy of water supplementation treatment in patients with functional dyspepsia or irritable bowe syndrome (IBS) accompanying predominant constipation.METHODS: A total of 3 872 patients with functional dyspepsia and 3609 patients with irritable bowel syndrome were enrolled in the study by 18 Italina thermal centres.Patients underwent a first cycle of thermal therapy for 21 d. A year later patients were re-evaluated at the same centre and received another cycle of thermal therapy.A questionnare to collect personal data on social and occupational status, family and pathological case history,life style, clinical records, utilisation of welfare and health structure and devices was administered to each patient at basal time and one year after each thermal treatment.Sixty patients with functional dyspepsia and 20 with IBS and 80 healthy controls received an evaluation of gastric output and oro-cecal transit time by breath test analysis.Breath test was performed at basal time and after water supplementaton therapies. Gastrointestinal symptoms were evaluated at the same time points. Breath samples were analyzed with a mass spectometer and a gascromatograph. Results were expressed as T1/2 and T-lag for octanoic acid breath test and as oro-cecal transit time for lactulose breath test.RESULTS: A significant reduction of prevalence of symptoms was observed at the end of the first and second cycles of thermal therapy in dyspeptic and IBS patients.The analysis of variance showed a real and persistant improvement of symptoms in all patients. After water supplementation for 3 wk a reduction of gastric output was observed in 49 (87.5%) of 56 dyspeptic patients.Both T1/2 and T-lag were significantly reduced after the therapy compared to basal values [91± 12 (T1/2) and 53 ± 11 (T-lag), Tables 1 and 2] with results of octanoic acid breath test similar to healthy subjects. After water supplementation for 3 wk oro-cecal transit time was shorter than that at the

  17. Evaluation of nondestructive testing techniques for the space shuttle nonmetallic thermal protection system

    Science.gov (United States)

    Tiede, D. A.

    1972-01-01

    A program was conducted to evaluate nondestructive analysis techniques for the detection of defects in rigidized surface insulation (a candidate material for the Space Shuttle thermal protection system). Uncoated, coated, and coated and bonded samples with internal defects (voids, cracks, delaminations, density variations, and moisture content), coating defects (holes, cracks, thickness variations, and loss of adhesion), and bondline defects (voids and unbonds) were inspected by X-ray radiography, acoustic, microwave, high-frequency ultrasonic, beta backscatter, thermal, holographic, and visual techniques. The detectability of each type of defect was determined for each technique (when applicable). A possible relationship between microwave reflection measurements (or X-ray-radiography density measurements) and the tensile strength was established. A possible approach for in-process inspection using a combination of X-ray radiography, acoustic, microwave, and holographic techniques was recommended.

  18. Integrity evaluation of lower thermal shield under exposure to HFBR environment

    Energy Technology Data Exchange (ETDEWEB)

    Kassir, M.; Weeks, J.; Bandyopadhyay, K. [Brookhaven National Lab., Upton, NY (United States); Shewmon, P. [Ohio State Univ., Columbus, OH (United States)

    1998-01-01

    The effects of exposure to the HFBR environment on the carbon steel in the HFBR lower thermal shield were evaluated. Corrosion was found to be a non-significant degradation process. Radiation embrittlement has occurred; portions of the plate closest to the reactor are currently operating in the lower-shelf region of the Charpy impact curve (i.e., below the fracture toughness transition temperature). In this region, the effects of radiation on the mechanical properties of carbon steel are believed to have been saturated, so that no further deterioration is anticipated. A fracture toughness analysis shows that a large factor of safety (> 1.5) exists against propagation of credible hypothetical flaws. Therefore, the existing lower thermal shield structure is suitable for continued operation of the HFBR.

  19. Developing a structural health monitoring system for nuclear dry cask storage canister

    Science.gov (United States)

    Sun, Xiaoyi; Lin, Bin; Bao, Jingjing; Giurgiutiu, Victor; Knight, Travis; Lam, Poh-Sang; Yu, Lingyu

    2015-03-01

    Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. In total, there are over 1482 dry cask storage system (DCSS) in use at US plants, storing 57,807 fuel assemblies. Nondestructive material condition monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health", and more importantly, to guarantee the safe operation of radioactive waste storage systems (RWSS) during their extended usage period. A state-of-the-art nuclear structural health monitoring (N-SHM) system based on in-situ sensing technologies that monitor material degradation and aging for nuclear spent fuel DCSS and similar structures is being developed. The N-SHM technology uses permanently installed low-profile piezoelectric wafer sensors to perform long-term health monitoring by strategically using a combined impedance (EMIS), acoustic emission (AE), and guided ultrasonic wave (GUW) approach, called "multimode sensing", which is conducted by the same network of installed sensors activated in a variety of ways. The system will detect AE events resulting from crack (case for study in this project) and evaluate the damage evolution; when significant AE is detected, the sensor network will switch to the GUW mode to perform damage localization, and quantification as well as probe "hot spots" that are prone to damage for material degradation evaluation using EMIS approach. The N-SHM is expected to eventually provide a systematic methodology for assessing and monitoring nuclear waste storage systems without incurring human radiation exposure.

  20. OECD benchmark on thermal fatigue problem - Second step: Evaluation based on FEM

    International Nuclear Information System (INIS)

    In the context of the OECD benchmark on thermal fatigue problem, this report corresponds to the second step of the procedure, i.e. integrity assessment in thermo-mechanical test interpretations. In this second step, participants were asked to estimate the number of cycles to crack initiation on the inner surface, without any initial notch. This report consists of evaluations based on FEM. Thermal and stress calculations are done by 3D-FEM, and based on these results, the number of cycles to crack initiation and propagation are estimated. The main conclusions are the following. Three dimensional Finite Element Analysis with JNC fatigue evaluation procedure estimated the most likely average crack initiation cycle as 2.8 x 104. The maximum Von Mises stress was 708 MPa, and the stress range was 653 MPa. The total strain was 0.48 pc. The estimated cracks are multiple longitudinal ones. Cracking area is from Z=213 to 198 mm and θ is equal to +/- 30 degrees. The cycle number for penetrating the thickness is estimated to be 4.6 x 103 and the crack will reach the outer surface during the planning test period. A detailed estimation of heat transfer coefficients Hwater and Houter is inevitable for a precise stress prediction. Precise estimations of thermal and mechanical properties, especially thermal conductivity, are important for a precise stress prediction. The difference among 3D, 2D, and 1D FEM is large; only 3D FEM can consider global circumstantial temperature difference, which leads to generate large membrane stress. With this membrane stress, the cycle number for crack penetration becomes 9 times shorter than the one without membrane stress. This effect is quite large and the precise estimation of the membrane stress is important