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Sample records for cask thermal evaluation

  1. THERMAL EVALUATION OF ALTERNATE SHIPPING CASK FOR GTRI EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-06-01

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

  2. Thermal model of spent fuel transport cask

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

    1996-01-01

    The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

  3. SCOPE, Shipping Cask Optimization and Parametric Evaluation

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: Given the neutron and gamma-ray shielding requirements as input, SCOPE may be used as a conceptual design tool for the evaluation of various casks designed to carry square fuel assemblies, circular canisters of nuclear waste material, or circular canisters containing 'intact' spent-fuel assemblies. It may be used to evaluate a specific design or to search for the maximum number of full assemblies (or canisters) that might be shipped in a given type of cask. In the 'search' mode, SCOPE will use built-in packing arrangements and the tabulated shielding requirements input by the user to 'design' a cask carrying one fuel assembly (or canister); it will then continue to increment the number of assemblies (or canisters) until one or more of the design limits can no longer be met. In each case (N = 1,2,3...), SCOPE will calculate the steady-state temperature distribution throughout the cask and perform a complete 1-D space/time transient thermal analysis following a postulated half-hour fire; then it will edit the characteristic dimensions of the cask (including fins, if required), the total weight of the loaded case, the steady-state temperature distribution at selected points, and the maximum transient temperature in key components. With SCOPE, the effects of various design changes may be evaluated quickly and inexpensively. 2 - Method of solution: SCOPE assumes that the user has already made an independent determination of the neutron and gamma-ray shielding requirements for the particular type of cask(s) under study. The amount of shielding required obviously depends on the type of spent fuel or nuclear waste material, its burnup and/or exposure, the decay time, and the number of assemblies or canisters in the cask. Source terms (and spectra) for spent PWR and BWR fuel assemblies are provided at each of 17 decay times, along with recommended neutron and gamma-ray shield thicknesses for Pb, Fe, and U-metal casks containing a

  4. Thermal tests of a transport / Storage cask in buried conditions

    International Nuclear Information System (INIS)

    Yamakawa, H.; Gomi, Y.; Saegusa, T.; Ito, C.

    1998-01-01

    Thermal tests for a hypothetical accident which simulated accidents caused by building collapse in case of an earthquake were conducted using a full-scale dry type transport and storage cask (total heat load: 23 kW). The objectives of these tests were to clarify the heat transfer features of the buried cask under such accidents and the time limit for maintaining the thermal integrity of the cask. Moreover, thermal analyses of the test cask under the buried conditions were carried out on basis of experimental results to establish methodology for the thermal analysis. The characteristics of the test cask are described as well as the test method used. The heat transfer features of the buried cask under such accidents and a time for maintaining the thermal integrity of the cask have been obtained. (O.M.)

  5. Thermal analysis for spent fuel casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1986-01-01

    Thermal analyses for spent fuel storage or transportation must demonstrate that the heat produced in the fuel will be removed without causing excessive fuel cladding temperatures or cask surface temperatures. The time required to ship fuel from a reactor to another site is a matter of days so temperature control during shipment is a short term concern and temperatures of 1100-1200 degrees F are acceptable in a hypothetical fire scenario, but storage is envisioned for twenty years or more and temperatures are limited to 380 degrees C (716 F). Elevated temperatures do more that weaken fuel cladding: the materials of the fuel basket also weaken with higher temperature, thermal stresses may cause cracks to form in welds, and impact limiter materials may soften or degrade. The high temperatures of a fire cause loss of liquid neutron shields and may cause solid materials to char and insulate a cask after the fire

  6. Nondestructive Evaluation of the VSC-17 Cask

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Nondestructive Evaluation of the VSC-17 Cask

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-01

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

  8. STACE: source term analyses for containment evaluations of transport casks

    International Nuclear Information System (INIS)

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

    1993-01-01

    STACE evaluates the calculated fuel rod response against failure criteria based on the cladding residual ductility and fracture properties as functions of irradiation and thermal environments. The fuel rod gap inventory contains three forms of releasable RAM: (1) gaseous, e.g., 85 Kr, (2) volatiles, e.g., 134 Cs and 137 Cs, and (3) actinides associated with fuel fines. The quantities of these products are limited to that contained within the fuel-cladding gap region and associated interconnected voids. Cladding pinhole failure will also result in the ejection of about 0.003 percent of the fuel, in the form of fines, into the cask cavity. Significant attenuation of the aerosol concentration in the transport cask can occur, depending upon the residence time of the aerosol in the cask compared with its rate of escape from the cask into the environment. (J.P.N.)

  9. Thermal analyses of the IF-300 shipping cask

    International Nuclear Information System (INIS)

    Meier, J.K.

    1978-07-01

    In order to supply temperature data for structural testing and analysis of shipping casks, a series of thermal analyses using the TRUMP thermal analyzer program were performed on the GE IF-300 spent fuel shipping cask. Major conclusions of the analyses are: (1) Under normal cooling conditions and a cask heat load of 262,000 BTU/h, the seal area of the cask will be roughly 100 0 C (180 0 F) above the ambient surroundings. (2) Under these same conditions the uranium shield at the midpoint of the cask will be between 69 0 C (125 0 F) and 92 0 C (166 0 F) above the ambient surroundings. (3) Significant thermal gradients are not likely to develop between the head studs and the surrounding metal. (4) A representative time constant for the cask as a whole is on the order of one day

  10. Prediction of thermal capability for a proposed LMFBR spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Reed, A.W.

    1979-02-01

    Prediction of the thermal capability of a fast breeder spent fuel shipping cask requires consideration of a large number of variables, such as geometry, canister packing fluid, gaseous coolant pressure, surface emissivities, and fin densities. In order to evaluate the individual influences of the large number of controlling parameters, a parametric analysis of a conceptual cask has been performed. Results of the analysis are applied to four options, each having a different configuration. Comparison of the options indicates that substantial increases in cask thermal capability may be obtained by the use of liquid sodium as a primary coolant and aluminum as a basket material.

  11. 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)

  12. Analysis technology on the temperature and thermal stress of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    Shin, L.Y.; Jin, C.Y.; Soo, K.H.; Hwan, C.S.

    2005-01-01

    The cask is used to transport the radioactive materials. It is required to withstand for the thirty minute under the hypothetical fire accident condition of the 800□. According to development of the computer simulation, finite element analysis is applied to the calculation widely. But finite element method for the hypothetical accident conditions is not established in domestic regulations. In this study, the temperature and thermal stress analysis of KSC-4 cask under 800□ fire condition is conducted using by ANSYS 7.0 code. In order to analyze finite elements, two-dimensional model of KSC-4 cask is used. Symmetric boundary, convection, and radiation condition are applied in the analysis. As the results, maximum temperature and thermal stress of the KSC-4 cask is evaluated. (orig.)

  13. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1997-01-01

    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

  14. 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.

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

    International Nuclear Information System (INIS)

    Yamakawa, H.; Gomi, Y.; Ozaki, S.; Kosaki, A.

    1993-01-01

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-20

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

  17. Thermal Analysis of Concrete Storage Cask with Bird Screen Meshes

    International Nuclear Information System (INIS)

    Lee, Ju-Chan; Bang, K.S.; Yu, S.H.; Cho, S.S.; Choi, W.S.

    2016-01-01

    In this study, a thermal analysis of the cask with bird screen meshes has been performed using a porous media model. The overpack consists of a structural material, a concrete shielding, and a ventilation system. Heat is removed from the cask to the environment by a passive means only. Air inlet and outlet ducts are installed at the bottom and top of the cask for a ventilation system. Bird screen meshes are installed at the air inlet and outlet ducts to inhibit intrusion of debris from the external environment. The presence of this screens introduce an additional resistance to air flow through the ducts. Five types of meshes for bird screen were considered in this study. The bird screen meshes at the inlet and outlet vents reduce the open area for flow by about 44 - 79 %. Flow resistance coefficients for porous media model were deduced from the fluid flow analysis of bird screen meshes. Thermal analyses for the concrete cask have been carried out using a porous media model. The analysis results agreed well with the test results. Therefore, it was shown that the porous media model for the screen mesh was established to estimate the cask temperatures

  18. Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

    2013-12-15

    Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

  19. 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.

  20. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    International Nuclear Information System (INIS)

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

    2007-01-01

    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

  1. Spent fuel shipping cask accident evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Fields, S.R.

    1975-12-01

    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.

  2. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  3. Thermal-Hydraulic Results 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)

    2017-09-01

    The thermal performance of commercial nuclear spent fuel dry storage casks is 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 canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground 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 aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric

  4. 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.

  5. Cooling performance evaluation of the concrete cask

    International Nuclear Information System (INIS)

    Maruoka, Kunio; Matsunaga, Kenichi; Abe, Ganji; Irino, Mitsuhiro; Arikawa, Hiroshi; Tamaki, Mitsuo

    2002-01-01

    The concrete cask storage system stores spent fuel by first sealing it within canisters and then containing such canisters inside a concrete cask. This report describes the results of a full-size model test performed to examine the heat dissipation characteristics of the concrete cask and to ascertain its ability to deal with elevated temperature. The specification to which a full-size concrete cask model was fabricated assumed an interim storage of 17x17UO 2 fuel that was burned in PWR, estimating the heating value of spent fuel containing canister to be approximately 20 kW apiece. The test, which actually covered canister heating values ranging from 10 kW to 30 kW per unit to allow for temperature variations likely to be experienced in actual operation, verified that the concrete cask member did not exceed temperature limits. Test condition anticipated highest air temperature inside the spent fuel storage facility to be 30degC and, with reference to existing standard, set temperature limits of 65degC or less for the main body of concrete and 90degC or less for the local part as criteria. Preliminary 3-D thermo hydrodynamic analysis done prior to the test indicated that the temperature of the local part of the concrete cask member would be below 90degC. It also confirmed that steel material used as the structural member of the canisters or concrete cask would remain around 200degC even in an area where it was highest, validating that the integrity of such material would pose no problem from the analytical point of view. Heat dissipation performance test conducted in steady state verified that the concrete cask was able to have a sufficient cooling capacity against per-canister heating values in the 10 kW to 30 kW range which had been chosen in anticipation of temperature variation thought to be encountered in actual service. Also, to examine the consequence of the concrete cask having lost its cooling ability, another heat dissipation test was carried out under

  6. Evaluation of the KN-12 spent fuel shipping cask

    International Nuclear Information System (INIS)

    Chung, Sung Hwan; Yoon, Jeong Hyoun; Yang, Ke Hyung; Kim, Jung Mook; Lee, Heung Young; Ha, Jong Hyun; Song, Myung Jae; Diersch, Rudolf; Laug, Reiner

    2001-01-01

    The KN-12 shipping cask is a new design of a transport package intended for dry and wet transportation of up to 12 spent nuclear fuel assemblies from pressure water reactors. The cask has been designed basing on KEPCO-NETEC's requirements and evaluated as a transport package that complies with the requirements of IAEA Safety Standards Series No.ST-1, US 10 CFR Part 71 and Korea Atomic Energy Act for Type B(U)F package. The cask will be licensed in accordance with Korea Atomic Energy Act. The cask provides containment, radiation shielding, structural integrity, criticality control and passive heat removal for normal transport conditions and hypothetical accident conditions. The W.H. 14x14, 16x16 and 17x17 fuel assemblies will be loaded and subsequently transported in the cask. The maximum allowable initial enrichment of the fuel is 5.0wt%, the fuel assembly burnup is limited to a maximum average of 50,000MWD/MTU, and the fuel must have a minimum cooling time of 7 years. And, the KN-12 cask will be fabricated in accordance with the requirements of ASME B and PV Code Section III, Division 3

  7. Internal pressure changes of liquid filled shipping casks due to thermal environment

    International Nuclear Information System (INIS)

    Jackson, J.E.

    1978-01-01

    A discussion of the significance of internal pressure calculations in liquid filled shipping casks subjected to a high temperature thermal environment is presented. Some basic thermodynamic relationships are introduced and discussed as they apply to the two-phase mixture problem encountered with liquid filled casks. A model of the liquid filled cask is developed and the assumptions and limitations of the mathematical model are discussed. A relationship is derived which can be used to determine internal cask pressures as a function of initial thermodynamic loading conditions, initial fluid volume ratio and final mixture temperature. The results for water/air filled casks are presented graphically in a parametric form. The curves presented are particularly useful for preliminary design verification purposes. A qualitative discussion of the use of the results from an error analysis aspect is presented. Some pressure calculation problems frequently seen by NRC for liquid filled cask designs are discussed

  8. Numerical Simulation of the Thermal Performance of a Dry Storage Cask for Spent Nuclear Fuel

    Directory of Open Access Journals (Sweden)

    Heui-Yung Chang

    2018-01-01

    Full Text Available In this study, the heat flow characteristics and thermal performance of a dry storage cask were investigated via thermal flow experiments and a computational fluid dynamics (CFD simulation. The results indicate that there are many inner circulations in the flow channel of the cask (the channel width is 10 cm. These circulations affect the channel airflow efficiency, which in turn affects the heat dissipation of the dry storage cask. The daily operating temperatures at the top concrete lid and the upper locations of the concrete cask are higher than those permitted by the design specification. The installation of the salt particle collection device has a limited negative effect on the thermal dissipation performance of the dry storage cask.

  9. Transport package response to severe thermal events, part 2: legal weight truck cask

    International Nuclear Information System (INIS)

    Greiner, M.; Faulkner, R.J.; Jin, Y.Y.

    1998-01-01

    The response of intact and damaged versions of the GA-4 Legal Weight Truck Cask to a range of severe thermal events is simulated using finite element computer analysis. The minimum fire durations that cause the containment seals and fuel cladding to reach their respective temperature limits are evaluated for a range of hydrocarbon fire temperatures. Containment seals reach their temperature limit in shorter duration fires as compared to the cladding, for both an undamaged package and a cask whose impact limiter is destroyed moments before the fire begins. However, if the neutron shield is destroyed, the cladding reaches its limit first in high temperature fires. A margin of safety exists between the conditions of the IAEA regulatory fire test and all of the performance envelopes calculated in this work. (author)

  10. Evaluation of canister weld flaw depth for concrete storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Tae Chul; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Jung, Sung Hun; Lee, Young Oh; Jung, In Su [Korea Nuclear Engineering and Service Corp, Daejeon (Korea, Republic of)

    2017-03-15

    Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radioactive materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B and PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

  11. Basic Considerations for Dry Storage of Spent Nuclear Fuels and Revisited CFD Thermal Analysis on the Concrete Cask

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Jae Soo [ACT Co. Ltd., Daejeon (Korea, Republic of); Park, Younwon; Song, Sub Lee [BEES Inc., Daejeon (Korea, Republic of); Kim, Hyeun Min [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    The integrity of storage facility and also of the spent nuclear fuel itself is considered very important. Storage casks can be located in a designated area on a site or in a designated storage building. A number of different designs for dry storage have been developed and used in different countries. Dry storage system was classified into two categories by IAEA. One is container including cask and silo, the other one is vault. However, there is various way of categorization for dry storage system. Dry silo and cask are usually classified separately, so the dry storage system can be classified into three different types. Furthermore, dry cask storage can be categorized into two types based on the type of the materials, concrete cask and metal cask. In this paper, the design characteristics of dry storage cask are introduced and computational fluid dynamics (CFD) based thermal analysis for concrete cask is revisited. Basic principles for dry storage cask design were described. Based on that, thermal analysis of concrete dry cask was introduced from the study of H. M. Kim et al. From the CFD calculation, the temperature of concrete wall was maintained under the safety criteria. From this fundamental analysis, further investigations are expected. For example, thermal analysis on the metal cask, thermal analysis on horizontally laid spent nuclear fuel assemblies for transportation concerns, and investigations on better performance of natural air circulation in dry cask can be promising candidates.

  12. Structural analysis of deformations in the seal region of a radioactive material transport cask under thermal loadings

    International Nuclear Information System (INIS)

    Turula, P.; Wang, Z.

    1991-01-01

    In the design of casks for the transport of highly radioactive materials, a major concern is preventing leakage under both the normal conditions of transport and the hypothetical accident conditions specified in the federal regulations. The designs must accommodate various loadings, including thermal loadings. The latter are particularly difficult to withstand because thermal considerations require exceptional structural flexibility, while other design considerations require structural stiffness. To guarantee a design that will be leaktight, the designer needs to quantify and limit the range of possible movement between the cask body and the lid under all conditions. This paper first shows the deformations obtained from a simplified model of the basic configuration of a sample design under selected normal and hypothetical thermal loading conditions. The design is then modified and re-analyzed to evaluate the effectiveness of two configuration modifications intended to reduce relative movement at the closure interface between cask body and the lid. Comparisons of required bolting forces demonstrate that accommodating the thermal loadings must be a primary consideration in designing the cask. 5 refs., 6 figs., 1 tab

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

    International Nuclear Information System (INIS)

    Mimura, Masahiro; Tsuda, Kazuaki; Yamada, Nobuyuki; O-iwa, Akio.

    1993-01-01

    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.)

  14. Structural evaluation of spent fuel dry storage cask

    International Nuclear Information System (INIS)

    Su, K. S.; Lee, J. H.; Kang, K. H.; Pak, S. W.; Jung, S. H.

    2003-01-01

    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

  15. CASKETSS-2: a computer code system for thermal and structural analysis of nuclear fuel shipping casks (version 2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1991-08-01

    A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  16. Experimental assessment on the thermal effects of the neutron shielding and heat-transfer fin of dual purpose casks on open pool fire

    International Nuclear Information System (INIS)

    Bang, Kyoung-Sik; Yu, Seung-Hwan; Lee, Ju-Chan; Seo, Ki-Seog; Choi, Woo-Seok

    2016-01-01

    Highlights: • An open pool fire test was performed to estimate not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin of the dual purpose cask. • The heat transfer to the inside of the dual purpose cask was reduced, when the neutron shielding burns. • The surface temperatures are lower in the present of the heat transfer fins. • If inflammable material is used as the components of the cask, evaluating thermal integrity using the thermal test would be desirable. - Abstract: Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 °C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 °C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the

  17. T-3 cask users' manual. Revision 1

    International Nuclear Information System (INIS)

    1986-06-01

    This user's manual for the T-C spent fuel cask provides information on: operating procedures; inspection and maintenance procedures; criticality evaluation; shielding evaluation; thermal evaluation; structural evaluation; and limitations

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

    Energy Technology Data Exchange (ETDEWEB)

    Frost, R.L.

    1999-02-26

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

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

    International Nuclear Information System (INIS)

    Kim, Chang Hyun

    1997-02-01

    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

  20. Numerical study of thermal test of a cask of transportation for radioactive material

    International Nuclear Information System (INIS)

    Vieira, Tiago A.S.; Santos, André A.C. dos; Vidal, Guilherme A.M.; Silva Junior, Geraldo E.

    2017-01-01

    In this study numerical simulations of a transport cask for radioactive material were made and the numerical results were compared with experimental results of tests carried out in two different opportunities. A mesh study was also made regarding the previously designed geometry of the same cask, in order to evaluate its impact in relation to the stability of numerical results for this type of problem. The comparison of the numerical and experimental results allowed to evaluate the need to plan and carry out a new test in order to validate the CFD codes used in the numerical simulations

  1. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    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 have little or no impact on the criticality results and/or conclusions

  2. An evaluation of the use of depleted uranium as a structural component for transport casks

    International Nuclear Information System (INIS)

    McConnell, P.; Salzbrenner, R.; Wellman, G.W.; Sorenson, K.B.

    1993-01-01

    Depleted uranium (DU) alloys are currently used for gamma-ray shielding in casks and as shielding blocks. For the transport cask application, a significant weight and dimensional penalty exists when using the DU solely for shielding. If credit could be taken for the structural use of the DU for containment in a transport cask, greater payloads may be realized. Mechanical property measurements of several uranium alloys and finite element analyses of prototype transport casks assumed to be constructed, in part, from selected uranium materials were performed to evaluate the potential for the use of DU alloys for cask containment. These data and analyses support the concept of the use of DU alloys for the containment function even under hypothetical accident conditions. A conclusion is that the properties of certain DU alloys are therefore sufficient to warrant further consideration of the material for this purpose

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

    International Nuclear Information System (INIS)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D.; Choi, B.I.; Lee, H.Y.; Song, M.J.

    2004-01-01

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

  4. 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.

  5. 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)

    Lee, Dew Hey; Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Choi, Kyung Joo; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun; Kim, Seong Jong

    2002-03-01

    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

  6. Evaluation of Cask Drop Criticality Issues at K Basin

    Energy Technology Data Exchange (ETDEWEB)

    GOLDMANN, L.H.

    2000-01-24

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

  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. Safety analysis report for Hot-Cell irradiated specimen cask

    Energy Technology Data Exchange (ETDEWEB)

    Ku, J. H.; Lee, J. C.; Seo, K. S.; Lee, D. W. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    For the examination of spent fuels and radioactive materials by using scanning electron microscope, a irradiated specimen cask is needed to transport the specimen from the hot-cell to the shielded glove box in which the scanning electron microscope is installed. This cask should be easy to handle and transport, has safe to maintain the shielding safety of operators as well as the thermal and structural integrities under prescribed load conditions by the regulations as requirements. Also the cask should be assured that docked perfectly maintaining shielding integrity with the interfaces of hot-cell and shield glove box. Accordingly, the main features of cask were analyzed with functional capabilities, and the integrities of cask under required load conditions were evaluated. Therefore, it was verified that the cask is suitable to use at the outside transport as well as Post Irradiated Examination Facility in KAERI. 9 refs., 50 figs., 14 tabs. (Author)

  9. TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

    International Nuclear Information System (INIS)

    Shah, M.J.; Klymyshyn, Nicholas A.; Adkins, Harold E.; Koeppel, Brian J.

    2007-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 (1). Both the International Atomic Energy Agency regulations for transporting radioactive materials (2, paragraph 727), and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 (3), for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask (3) for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS (4) and LS DYNA (5) finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event

  10. Source-term Determination and Dose-rates Evaluation For RTP Fuel Transfer Cask Design

    International Nuclear Information System (INIS)

    Mohamad Hairie Rabir

    2016-01-01

    This paper describes MCNPX calculations of the gamma-ray doses rates arising from two TRIGA spent-fuel transfer casks candidate for PUSPATI TRIGA Reactor (RTP). Calculations of radiation fields at several points along the axial length of the casks are reported. These external dose rates are reported for both primary gamma rays (arising from fission and activation products in the fuel). Photon shielding calculations and the source term were evaluated by the MCNPX. From the calculation, it was found that the maximum surface dose rates of the fuel transfer cask with the two candidates (41 cm and 55 cm in diameter) were estimated within the limit (2 mSv/ hr). (author)

  11. 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)

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

    1995-02-01

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

  12. Seismic and cask drop excitation evaluation of the tower shielding reactor

    International Nuclear Information System (INIS)

    Harris, S.P.; Stover, R.L.; Johnson, J.J.; Sumodobila, B.N.

    1989-01-01

    During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations. 6 figs

  13. Seismic and cask drop excitation evaluation of the Tower Shielding Reactor

    International Nuclear Information System (INIS)

    Stover, R.L.; Harris, S.P.; Johnson, J.J.; Sumodobila, B.N.

    1989-01-01

    During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations

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

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1988-12-01

    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)

  15. 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)

  16. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    International Nuclear Information System (INIS)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

    2016-01-01

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Andress, D. (Andress (David) and Associates, Inc., Kensington, MD (USA)); Joy, D.S. (Oak Ridge National Lab., TN (USA)); McLeod, N.B. (Johnson (E.R.) Associates, Inc., Oakton, VA (USA)); Peterson, R.W. (Bentz (E.J.) and Associates, Inc., Alexandria, VA (USA)); Rahimi, M. (Jacobs Engineering Group, Inc., Washington, DC (USA))

    1991-01-01

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

  19. Safety analysis report for radwaste foam transport cask

    International Nuclear Information System (INIS)

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

    1999-08-01

    For the tests and examinations of radwaste foam which generated in domestic nuclear power plants a radioactive material transport cask is needed to transport the radwaste foam from the power plants to KAERI. This cask should be easy to handle in the facilities and safe to maintain the shielding safety of operators. According to the regulations, it should be verified that this cask maintains the thermal and structural integrities under prescribed load conditions by the regulations. The basic structural functions and the integrities of the cask under required load conditions were evaluated. Therefore, it was verified that the cask is suitable to transport radwaste foam from nuclear power plants to KAERI. (author). 11 refs., 10 tabs., 25 figs

  20. 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.

  1. Evaluation of the impact behavior of the contents of reprocessing radioactive waste shipping cask subjected to drop impact

    International Nuclear Information System (INIS)

    Shirai, K.; Ito, C.; Funahashi, M.

    1993-01-01

    In this study, to investigate the impact response characteristics of the contents in the cask precisely, we performed the laboratory-scale drop tests, and on the basis of the test results, we proposed the construction of the spring-mass model and confirmed the accuracy of the proposed drop analysis method by comparison with drop test using a full-scale cask for high level wastes. Following the results of the drop tests and analysis, the outline of the contents and results is summarized below. 1) The drop tests onto the unyielding surface using a scale model containing several contents were performed and the effect of the interaction between the contents and the cask body on the impact response experimentally. 2) The above interaction can be characterized by the gap between the contents and the cask body caused by the release of the gravitational force at the moment the drop started. So, we proposed the analysis method for considering gap using spring-mass model by comparing the laboratory-scale drop test results. 3) We applied the proposed analysis method to a drop test using a full-scale cask for high level wastes, and it was found that this method seems to be good and convenient enough to evaluate the impact behavior of the contents in a transport cask subjected to drop impact. (J.P.N.)

  2. Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory Shipping Cask D-38. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Box, W.D.; Shappert, L.B.; Seagren, R.D.; Watson, C.D.; Hammond, C.R.; Klima, B.B.

    1979-09-01

    An analytical evaluation of the Oak Ridge National Laboratory Shipping Cask D-38 (solids shipments) was made to demonstrate its compliance with the regulations governing off-site radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the cask complies with the applicable regulations.

  3. Fire protection device for LMFBR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Moss, M.; Heckman, R.C.

    1977-08-01

    This report describes the analysis and experimental evaluation of a device to provide fire protection for a shipping cask used to transport liquid metal fast breeder reactor (LMFBR) fuel rods. Thermal analyses of various fire protection schemes were conducted by means of the finite difference code, CINDA. The choice of materials for the cask body was determined to be less important than the dimensions and number of cooling fins attached to the cask. Of several protection methods considered, radiation/convection shields between the cooling fins looked most attractive; these were tested on a quarter-scale cask model. The shields consisted of bimetallic bands which expanded when heated, blocking the fire, and retracted when cooled. This automatic reversibility would allow the fins to dissipate internally generated heat of a full-size cask once the fire was out. Tests showed that, even in an asymmetric fire exposure, the bands expanded to provide protection to the cask. The directly exposed surface of the cask model, protected this way, reached a temperature of only 180 0 C compared with 295 0 C in the unprotected state, when subjected to a butane flame for 0.5 h at a distance of 15 cm. Greater relative effectiveness could be expected in an engulfing fire in which the bands would expand more symmetrically

  4. Criticality Evaluation - Cask Unloading Using the Shielded Transfer System

    International Nuclear Information System (INIS)

    Blanchard, A.; Nadeau, M.L.

    1998-01-01

    This evaluation reviewed the criticality aspects of Shielded Transfer System (STS) operation. The existing Nuclear Safety Control (NSC) elements were examined and new NSC elements were developed, as needed

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

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, R.J.

    1998-11-17

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

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

    International Nuclear Information System (INIS)

    SMITH, R.J.

    1998-01-01

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

  7. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

    The thermal performance of commercial nuclear spent fuel dry storage casks is 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 also by increasing the internal convection through greater canister helium pressure. These same 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 current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an above-ground configuration.

  8. Structural evaluation of spent nuclear fuel storage facilities under aircraft crash impact. Numerical study on evaluation of sealing performance of metal cask subjected to impact force

    International Nuclear Information System (INIS)

    Namba, Kosuke; Shirai, Koji; Saegusa, Toshiari

    2008-01-01

    A lot of safety evaluations on the important nuclear facilities against the aircraft crash have been reported in other countries. But the condition and the evaluation method to define impact force of aircraft crash have not been described clearly in the reports. In Japan, public concern with the safety evaluation against aircraft crash is increasing. It is important to make clear the behavior of the storage facilities installing the metal casks on impact loading due to aircraft crash. In this study, concerning crash between commercial aircraft and storage facility, impact analysis using dynamic analysis code LS-DYNA has been executed. The results showed that the storage facility was not completely destroyed. But the rigid aircraft engine may penetrate into the storage facility with local failure. Thus, we assumed the engine hit a metal cask in the storage facility and evaluated sealing performance of the metal cask under the impact loading. If the engine with 90m/s crashed the storage facility having concrete wall of 85cm in thickness, the remaining velocity became 60m/s after penetration. We calculated impact force of the engine with 60m/s crashing into the metal cask. Concerning the metal cask loaded the impact force, impact analysis was executed. We assumed two directions of impact force. One is vertical load and another is horizontal load against the cask. The result showed that plastic strain was not generated on flanges of the 1st lid and the sealing performance of the cask was maintained in each impact case. (author)

  9. Description of from-reactor transportation cask designs

    International Nuclear Information System (INIS)

    Lake, W.H.

    1990-01-01

    This paper describes two from-reactor cask development program contracts. They are a contract for legal weight truck cask designs, and a contract for a rail/barge cask design. The paper also presents several general considerations affecting the cask development program. Two of these which are covered in some detail are the technical topics of burnup credit and source term evaluation

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

    International Nuclear Information System (INIS)

    Sanders, C.; Kimball, D.

    2005-01-01

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

  11. GA-4/GA-9 legal weight truck from reactor spent fuel shipping casks

    International Nuclear Information System (INIS)

    1990-04-01

    The preliminary design report presents the results of General Atomics (GA) preliminary design effort to develop weight truck from reactor spent fuel shipping casks. The thermal evaluation of the Office of Civilian Radioactive Waste Management (OCRWM) cask considered normal and hypothetical accident conditions of transport. We employed analytical modeling as well as fire testing of the neutron shielding material to perform the evaluation. This document addresses the thermal design features of the cask, discusses thermal criteria, and summarizes the results of the thermal evaluation, as well as results of structural containment and nuclear evaluations that support the design. Also included are the results of trade-off studies. 69 refs., 103 figs., 76 tabs

  12. Modeling of fuel elements for the thermal design of CASTOR casks; Modellierung der Brennelemente bei der thermischen Auslegung von CASTOR-Behaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Greza, H.; Leber, A.; Kueck, T. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Funke, T. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2012-11-01

    For the thermal design of transport and storage casks for radioactive wastes numerical calculation procedures are used to demonstrate the safe after-heat removal. In spite of the advanced software the modeling size and the level detail are still restricted. Therefore parts of the calculation models that exhibit a high level of detail are replaced by homogenized zones with effective material characteristics. The material data have to be determined by appropriate methods in order to provide calculated results of sufficient safety margins. The authors demonstrate the procedure using homogenized zones for the modeling of fuel elements in the CASTOR cask V/19. Comparative calculations using detailed modeling of the fuel elements show agreement of the maximum temperatures for the fuel elements within the calculation accuracy.

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

    International Nuclear Information System (INIS)

    Lundeen, J.E.

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

    Cha, Jeonghun; Kim, S.N.; Choi, K.W.

    2009-01-01

    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)

  15. Full scale torch tests on spent fuel cask shipping system

    International Nuclear Information System (INIS)

    Vigil, M.G.; Trujillo, A.A.; Yoshimura, H.R.; Joseph, B.J.; Eggers, P.E.; Crawford, H.L.

    1982-01-01

    Full scale experimental measurements, including the instrumentation designed to obtain the data, are presented on the thermal effects of torch fires on a large, spent nuclear fuel shipping cask. The measured temperature data in the various materials of the multilayered cask are unique, since no torch tests have been previously performed on a cask: These data were obtained during a series of four torch tests which simulate a situation in which the relief valve of a liquefied gas tank railcar has been opened and and the contents are vented and ignited so that the resultant torch impinges on the cask. The modified cask instrumentation geometry and materials are discussed. Temperature data throughout the cask are compared for two cask on the corrugated outer jacket surface, within the neutron shield, on the carbon steel shell, on the inner stainless steel shell and near the cask head closure seals are presented for the four torch tests

  16. 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.

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

    International Nuclear Information System (INIS)

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

    1992-11-01

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

  18. Seal performance of thermal aged metal gasket of dual purpose metal cask for interim spent fuel storage after external impact load

    International Nuclear Information System (INIS)

    Takeshi Yokoyama; Masami Kato; Satoshi Itooka

    2005-01-01

    As for interim storage for spent nuclear fuels using dual purpose dry metal cask in Japan, we recognize one of the important technical issues that there is a possibility for the cask with degraded metal gasket during storage to apply to transportation. In our study until 2003 focused on degradation of important components for the cask safety performance during storage and application to transportation, for metal gasket, we conducted property tests for degradation and influence of lid movement on seal performance, and furthermore verification tests. From the results, we developed the method to evaluate leak rate from lid with degraded metal gasket at accidents during transportation and in addition, we found as follows: Lid would hardly move and leak rate would not increase seriously during fire event. Leak rate from lid with degraded metal gasket could be evaluated by using results of leak rate trend depending on horizontal displacement of lid by external impact load. So, with regard to influence of lid movement on seal performance, we conducted additional test for extending horizontal displacement in lid moving in 2004. In addition, seal performance was discussed from the results, both previous and latest test. (authors)

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

    International Nuclear Information System (INIS)

    Pope, R.B.; Diggs, J.M.

    1982-04-01

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

  20. Fire resistant nuclear fuel cask

    Science.gov (United States)

    Heckman, Richard C.; Moss, Marvin

    1979-01-01

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

  1. B cell remote-handled waste shipment cask alternatives study

    International Nuclear Information System (INIS)

    RIDDELLE, J.G.

    1999-01-01

    The decommissioning of the 324 Facility B Cell includes the onsite transport of grouted remote-handled radioactive waste from the 324 Facility to the 200 Areas for disposal. The grouted waste has been transported in the leased ATG Nuclear Services 3-82B Radioactive Waste Shipping Cask (3-82B cask). Because the 3-82B cask is a U.S. Nuclear Regulatory Commission (NRC)-certified Type B shipping cask, the lease cost is high, and the cask operations in the onsite environment may not be optimal. An alternatives study has been performed to develop cost and schedule information on alternative waste transportation systems to assist in determining which system should be used in the future. Five alternatives were identified for evaluation. These included continued lease of the 3-82B cask, fabrication of a new 3-82B cask, development and fabrication of an onsite cask, modification of the existing U.S. Department of Energy-owned cask (OH-142), and the lease of a different commercially available cask. Each alternative was compared to acceptance criteria for use in the B Cell as an initial screening. Only continued leasing of the 3-82B cask, fabrication of a new 3-82B cask, and the development and fabrication of an onsite cask were found to meet all of the B Cell acceptance criteria

  2. TITAN Legal Weight Truck cask preliminary design report

    International Nuclear Information System (INIS)

    1990-04-01

    The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of this document includes the LWT cask with fuel baskets, impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. The results of the tradeoff studies and evaluations that were performed during the preliminary design are presented in Appendix A to this report. 51 figs., 17 tabs

  3. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Advanced Nuclear Fuel Cycle Technologies; Lindgren, Eric Richard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Advanced Nuclear Fuel Cycle Technologies

    2016-09-01

    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 also by increasing the internal convection through greater canister helium pressure. These same 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 belowground 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 aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask

  4. Development of cask and transportation system

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Kang, Hee Dong; Lee, Heung Young; Seo, Ki Suk; Koo, Jung Hoe; Jung, Sung Hwan; Yoon, Jung Hyun; Lee, Ju Chan; Bang, Kyung Sik; Baek, Chang Yeol

    1992-03-01

    The major goal of this project is to establish the safe transport system and obtain the necessary data for cask development by during research work for the design and safety test of shipping cask. The analysis technique using computer code for design has been studied in the field of structure, thermal and shielding analysis in this study. And also the test and measurement technology was developed for the measuring system of drop and fire test. It is expected that research activity ensured in this job will enable us to ultilize the basic data for the cask development. (Author)

  5. A validated methodology for evaluating burn-up credit in spent fuel casks

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  6. Evaluation of three partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

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

    1981-02-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Richard, R.F.

    1995-01-01

    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

  10. Experimental study of heat removal performance of a concrete storage cask

    International Nuclear Information System (INIS)

    Bang, K.S.; Yu, S.H.; Lee, S.H.; Lee, J.C.; Seo, K.S.

    2014-01-01

    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. 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, a concrete storage cask must be designed including heat removal capabilities with an appropriate reliability. However, the thermal conductivity of concrete is not good and the allowable temperature of concrete is lower than that of steel. Therefore, thermal test was performed to evaluate the heat removal performance of the concrete storage cask under 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 air through a passive heat removal system under normal conditions reached 93.5 %. Accordingly, the passive heat removal system was well designed and worked adequately. The maximum temperature of the over-pack inner surface under accident condition was measured at 103 deg. C. Therefore, the thermal integrity of the concrete is safe enough under accident conditions. (authors)

  11. 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)

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

    1998-03-01

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

  12. Spent fuel transportation cask response to a tunnel fire scenario

    International Nuclear Information System (INIS)

    Bajwa, C.S.; Adkins, H.E.; Cuta, J.M.

    2004-01-01

    On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. 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 an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation cask designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the ANSYS registered and COBRA-SFS computer codes to evaluate the thermal performance of different cask designs. The staff concluded that the transportation casks analyzed would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event. No release of radioactive materials would result from exposure of the casks analyzed to such an event. This paper describes the methods and approach used for this assessment

  13. Numerical study of thermal test of a cask of transportation for radioactive material; Estudo numérico do ensaio térmico de um casco de transporte para material radioativo

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Tiago A.S.; Santos, André A.C. dos; Vidal, Guilherme A.M.; Silva Junior, Geraldo E., E-mail: tiago.vieira.eng@gmail.com, E-mail: gvidal.ufmg@gmail.com, E-mail: aacs@cdtn.br, E-mail: geraldo.esilva@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    In this study numerical simulations of a transport cask for radioactive material were made and the numerical results were compared with experimental results of tests carried out in two different opportunities. A mesh study was also made regarding the previously designed geometry of the same cask, in order to evaluate its impact in relation to the stability of numerical results for this type of problem. The comparison of the numerical and experimental results allowed to evaluate the need to plan and carry out a new test in order to validate the CFD codes used in the numerical simulations.

  14. Development of metal cask for nuclear spent fuel

    International Nuclear Information System (INIS)

    Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    It is one of the realistic solutions against increasing demand on interim storage of spent fuel assemblies arising from nuclear power plants in Japan to apply dual purpose (transport and storage) metal casks. Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities, etc. in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage casks use various new design concepts and materials to improve thermal performance of the cask, structural integrity of the basket, durability of the neutron shielding material and so on. This paper summarizes an outline of the cask design that can accommodate BWR spent fuel assemblies as well as the new technologies applied to the design and fabrication. (author)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  16. Evaluation of microwave cavity gas sensor for in-vessel monitoring of dry cask storage systems

    Science.gov (United States)

    Bakhtiari, S.; Gonnot, T.; Elmer, T.; Chien, H.-T.; Engel, D.; Koehl, E.; Heifetz, A.

    2018-04-01

    Results are reported of research activities conducted at Argonne to assess the viability of microwave resonant cavities for extended in-vessel monitoring of dry cask storage system (DCSS) environment. One of the gases of concern to long-term storage in canisters is water vapor, which appears due to evaporation of residual moisture from incompletely dried fuel assembly. Excess moisture could contribute to corrosion and deterioration of components inside the canister, which would in turn compromise maintenance and safe transportation of such systems. Selection of the sensor type in this work was based on a number of factors, including good sensitivity, fast response time, small form factor and ruggedness of the probing element. A critical design constraint was the capability to mount and operate the sensor using the existing canister penetrations-use of existing ports for thermocouple lances. Microwave resonant cavities operating at select resonant frequency matched to the rotational absorption line of the molecule of interest offer the possibility of highly sensitive detection. In this study, two prototype K-band microwave cylindrical cavities operating at TE01n resonant modes around the 22 GHz water absorption line were developed and tested. The sensors employ a single port for excitation and detection and a novel dual-loop inductive coupling for optimized excitation of the resonant modes. Measurement of the loaded and unloaded cavity quality factor was obtained from the S11 parameter. The acquisition and real-time analysis of data was implemented using software based tools developed for this purpose. The results indicate that the microwave humidity sensors developed in this work could be adapted to in-vessel monitoring applications that require few parts-per-million level of sensitivity. The microwave sensing method for detection of water vapor can potentially be extended to detection of radioactive fission gases leaking into the interior of the canister through

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

    International Nuclear Information System (INIS)

    Otton, Camille; Sicard, Damien

    2007-01-01

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

  18. Concrete storage cask for interim storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Nabemoto, Toyonobu; Fujiwara, Hiroaki; Kobayashi, Shunji; Shionaga, Ryosuke

    2004-01-01

    Experiments and analytical evaluation of the fabrication, non-destructive inspection and structural integrity of reinforced concrete body for storage casks were carried out to demonstrate the concrete storage cask for spent fuel generated from nuclear power plants. Analytical survey on the type of concrete material and fabrication method of the storage cask was performed and the most suitable fabrication method for the concrete body was identified to reduce concrete cracking. The structural integrity of the concrete body of the storage cask under load conditions during storage was confirmed and the long term integrity of concrete body against degradation dependent on environmental factors was evaluated. (author)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. 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.

  1. A cask fleet operations study

    International Nuclear Information System (INIS)

    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

  2. Remote cask handling and implications for cask system design

    International Nuclear Information System (INIS)

    Griesmeyer, J.M.; Thunborg, S.

    1988-08-01

    Robotic handling of nuclear waste shipping casks has the potential to significantly reduce occupational radiation exposure and computer monitoring of operator interactions with the system can improve safety. Furthermore, robot programmability can provide an automated audit trail for quality assurance. This report discussed the impact of cask design on the potential application of robotic systems to repository based nuclear waste shipping cask handling operations. The main conclusions are: (1) incorporation of cask system design features which facilitate robotic cask handling can be achieved with minimal impact on cask functional features, (2) proper cask design allows robotic cask handling operations from unbolting cask tie-down straps to radiation surveys to be performed quickly and reliably without direct human intervention, and (3) design for remote handling also facilitates manual handling operations. 12 refs., 9 figs., 4 tabs

  3. Analysis of radiation measurement data of the BUSS cask

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tang, J.S.

    1995-01-01

    The Beneficial Uses Shipping System (BUSS) is a Type-B packaging developed for shipping nonfissile, special-form radioactive materials to facilities such as sewage, food, and medical-product irradiators. The primary purpose of the BUSS cask is to provide shielding and confinement, as well as impact, puncture, and thermal protection for its certified special-form contents under both normal transport and hypothetical accident conditions. A BUSS cask that contained 16 CsCl capsules (2.723 x 10 4 TBq total activity) was recently subjected to radiation survey measurements at a Westinghouse Hanford facility, which provided data that could be used to validate computer codes. Two shielding analysis codes, MICROSHIELD (User's Manual 1988) and SAS4 (Tan 1993), that are used at Argonne National Laboratory to evaluate the safety of packaging of radioactive materials during transportation, have been selected for analysis of radiation data obtained from the BUSS cask. MICROSHIELD, which performs only gamma radiation shielding calculation, is based on a point-kernel model with idealized geometry, whereas SAS4 is a control module in the SCALE code system (1995) that can perform three-dimensional Monte Carlo shielding calculation for photons and neutrons, with built-in procedures for cross-section data processing and automated variance reduction. The two codes differ in how they model the details of the physics of gamma photon attenuation in materials, and this difference is reflected in the associated engineering cost of the analysis. One purpose of the analysis presented in this paper, therefore, is to examine the effects of the major modeling assumptions in the two codes on calculated dose rates, and to use the measured dose rates for comparison. The focus in this paper is on analysis of radiation dose rates measured on the general body of the cask and away from penetrations

  4. Evaluation method of gas leakage rate from transportation casks of radioactive materials (gas leakage rates from scratches on O-ring surface)

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Li Ninghua; Asano, Ryoji; Kawa, Tsunemichi

    2004-01-01

    A sealing function is essential for transportation and/or storage casks of radioactive materials under both normal and accidental operating conditions in order to prevent radioactive materials from being released into the environment. In the safety analysis report, the release rate of radioactive materials into the environment is evaluated using the correlations specified in the ANSI N14.5, 1987. The purposes of the work are to reveal the underlying problems on the correlations specified in the ANSI N14.5 related to gas leakage rates from a scratch on O-ring surface and from multi-leak paths, to offer a data base to study the evaluation method of the leakage rate and to propose the evaluation method. In this paper, the following insights were obtained and clarified: 1. If a characteristic value of a leak path is defined as D 4 /a ('D' is the diameter and 'a' is the length), a scratch on the O-ring surface can be evaluated as a circular tube. 2. It is proper to use the width of O-ring groove on the flange as the leak path length for elastomer O-rings. 3. Gas leakage rates from multi leak paths of the transportation cask can be evaluated in the same manner as a single leak path if an effective D4/a is introduced. (author)

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

    International Nuclear Information System (INIS)

    Wieser, G.; Qiao, L.; Eberle, A.; Voelzke, H.

    2004-01-01

    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 (FE) to calculate the effects of mechanical impacts onto the bolted lid structures as well as effects caused by severe fire scenarios. In the case of fire it might not be enough to perform only a thermal heat transfer analysis. A complex cask design together 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 their counterparts that can be analysed by a so-called thermomechanical calculation. Although it is currently not possible to directly correlate leakage rates with results from deformation analyses, 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 for the lid bolts, the geometry of the cask and the mechanical loading is axisymmetric 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 modellings on the relative displacements at the seating of the seals. The influence of bolt modelling, thermal properties and the detailed geometry of the two-dimensional finite-element models on the

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

    International Nuclear Information System (INIS)

    Wieser, G.; Qiao Linan; Eberle, A.; Voelzke, H.

    2004-01-01

    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

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

    International Nuclear Information System (INIS)

    Uldrich, E.D.; Hawkes, B.D.

    1998-04-01

    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

  8. Cask crush pad analysis using detailed and simplified analysis methods

    International Nuclear Information System (INIS)

    Uldrich, E.D.; Hawkes, B.D.

    1997-01-01

    A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped spent nuclear fuel shipping cask into a 44-ft. deep cask unloading pool at the Fluorinel and Storage Facility (FAST). This facility, located at the Idaho Chemical Processing Plant (ICPP) at the Idaho national Engineering and Environmental Laboratory (INEEL), is a US Department of Energy site. The basis for this study is an analysis by Uldrich and Hawkes. The purpose of this analysis was to evaluate various hypothetical cask drop orientations to ensure that the crush pad design was adequate and the cask deceleration at impact was less than 100 g. It is demonstrated herein that a large spent fuel shipping cask, when dropped onto a foam crush pad, can be analyzed by either hand methods or by sophisticated dynamic finite element analysis using computer codes such as ABAQUS. Results from the two methods are compared to evaluate accuracy of the simplified hand analysis approach

  9. Evaluation of the salt deposition on the canister surface of concrete cask. Measurement test of the salt deposition in the laboratory and the field

    International Nuclear Information System (INIS)

    Wataru, Masumi; Saegusa, Toshiari

    2010-01-01

    Concerning the interim storage of spent nuclear fuel using the concrete cask, there is a problem of stress corrosion cracking. To evaluate the occurrence of SCC or emerging rust, it is important to verify the estimation of the sea salt accumulation transported by salty cooling air on the metal canister surface. The cooling air goes up along the canister surface in the concrete cask. To measure the deposition velocity, two kinds of test were performed. One is a laboratory test, another is a field test. The amount of sea salt deposition and acceleration factor of environment were obtained. In the laboratory test, it is found that the sea salt deposition density is higher when the air velocity is high and the temperature of the specimen is low. In the field test, it is found that the sea salt deposition density is very low because the density of the salt particle in the air was very low compared with the laboratory test. Using relation between laboratory data and filed data, it is possible to evaluate the salt deposition velocity on the canister surface. (author)

  10. Cask handling method and apparatus

    International Nuclear Information System (INIS)

    Yoli, A.H.; Husain, I.

    1977-01-01

    The method of transferring radioactive material into and out of the cask comprises positioning a tank with an open end in a well. Then a cask having a passage for moving radioactive material into and out of the cask is placed in the tank through the opening in the tank. The tank opening is then sealed to the cask relative to the well without sealing the passage relative to the well to prevent water filled into the well from leaking into the tank. Then the well is filled with water above the seal, and radioactive material is then moved through the water in the well through the passage into the cask. The tank may be filled with demineralized water from a separate source to pressurize the space in the tank on the other side of the seal from the well to prevent water in the well from entering the tank. The water level in the well and in the tank is then lowered, the tank opening to the cask seal is removed, and a cover is attached to the cask passage to maintain the radioactive material and contaminated water in the cask. The apparatus which accomplishes the above method comprises a tank in a well for receiving a cask therein. A seal between the tank and the cask prevents water in the well from flowing into the tank about the cask and permits water in the well to flow through the cask opening into the cask. A first water supply means raises and lowers the water level in the well, and a second water supply means supplies clean demineralized water to the tank under pressure to prevent water in the well from leaking into the tank. The seal is annularly shaped and is attached to the top of the tank. The central portion of the annular seal is aligned with the cask opening and it has means to seal the annular seal to the cask

  11. Handbook for structural analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1991-04-01

    This paper described structural analysis method of radioactive material transport casks for use of a handbook of safety analysis and evaluation. Safety analysis conditions, computer codes for analyses and stress evaluation method are also involved in the handbook. (author)

  12. Designing environment friendly casks

    Energy Technology Data Exchange (ETDEWEB)

    Lallemant, T. [TN International (France)

    2009-06-15

    Environmental protection has become an integral part of the entire cask life cycle: - Design and Fabrication; - Transportation; - Loading, Unloading and Operation; - Maintenance, End of Life Cycle and Dismantling. Three new environmentally-friendly cask concepts were developed: the MX6, the TN112 and the multipurpose Shell. Eco design approach was used to reduce their environmental impacts and address environmental issues such as waste generation and management, greenhouse gases linked to transportation, and scarcity of raw materials, all while improving operating performance. The result is greater responsiveness to the customer and a more comprehensive approach to potential regulatory changes. The cask design optimizes all of the impacting factors, such as raw materials quantity, radiological protection capacity, and the use of more environmentally-friendly materials and fabrication processes. To illustrate, the new MX6 concept may be compared to the standard FS65 concept. In the area of Design and Fabrication, the MX6 requires fewer raw materials and can lower the cask weight/transported assembly weight ratio. An adjustable basket concept was developed. - The new cask is also more effective in terms of radiation protection as a result of improved materials performance. - In the area of fabrication, all waste is transferred to suitable processing systems that promote recycling. In the area of Transportation, shipments are reduced by 33%, thus reducing diesel fuel consumption and greenhouse gas emissions. Also, in the area of Operations, the amount of waste generated per assembly shipped is minimized, as are the number of operations and the time spent on them by the operators. In the area of Maintenance and Dismantling, extensive, time-consuming maintenance operations are minimized, as is the waste generated by them. The TN112 cask reduces exposure by 15%, uses fewer raw materials, and has a faster turnaround time due to optimized tooling. For waste subject to

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-01

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

  14. Transportation cask contamination weeping

    International Nuclear Information System (INIS)

    Bennett, P.C.; Doughty, D.H.; Chambers, W.B.

    1993-01-01

    This paper describes the problem of cask contamination weeping, and efforts to understand the phenomenon and to eliminate its occurrence during spent nuclear fuel transport. The paper summarizes analyses of field experience and scoping experiments, and concentrates on current modelling and experimental validation efforts. (J.P.N.)

  15. Key technology studies of GY-20 and GY-40 High-capacity cobalt-60 transport casks

    International Nuclear Information System (INIS)

    Liu Huifang; Zhang Xin

    2012-01-01

    GY-20 and GY-40 high-capacity cobalt-60 transport casks are used to transport cobalt-60 industrial irradiators and cobalt-60 bundles. The radioactive contents have special features of high-activity and high residual heat, so only a few countries such as Canada, England and Russia have design capacity. The key technologies and corresponding solutions were studied for the design and manufacture of the cask taking into account the structural, thermal, mechanics and shield requests. A series of tests prove that the cask structure design, design criteria for lead coating structure and quality control measurements are reasonable and effective, and the cask shield integrity can be ensured for all conditions. The casks have ability to transport high-activity sealed sources safely, and the design of cask satisfies the requirement of design code and standard. It can provide reference for other B type package. (authors)

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

    International Nuclear Information System (INIS)

    BRAVERMAN, J.I.; MORANTE, R.J.; XU, J.; HOFMAYER, C.H.; SHAUKAT, S.K.

    2003-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1987-06-01

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

  18. 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.

  19. 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.

  20. Cask development, testing, and licensing

    International Nuclear Information System (INIS)

    Quinn, G.J.; Haelsig, R.T.; Warrant, M.M.

    1986-01-01

    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

  1. Safety Assessment of a Metal Cask under Aircraft Engine Crash

    OpenAIRE

    Sanghoon Lee; Woo-Seok Choi; Ki-Seog Seo

    2016-01-01

    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 freestanding on a...

  2. Issues related to the transport of a transportable storage cask after storage

    International Nuclear Information System (INIS)

    McConnell, P.; Brimhall, J.L.; Creer, J.M.; Gilbert, E.R.; Sanders, T.L.; Jones, R.H.

    1991-01-01

    An evaluation was performed to assess whether the reliability of a transportable storage cask system and the risks associated with its use are comparable to those associated with existing transport cask systems and, if they are not, determine how the transportable storage cask system can be made as reliable as existing systems. Reliability and failure mode analyses of both transport-only casks and transportable storage casks and implementation options are compared. Current knowledge regarding the potential effects of a long-term dry storage environment on spent fuel and cask materials is reviewed. A summary assessment of the consideration for deploying a transportable storage cask (TSC) system with emphasis on preliminary design, validation and operational recommendations for TSC implementations is presented. The analyses conclude that a transportable storage cask can likely be shipped upopened by applying a combination of design considerations and operational constraints, including environmental monitoring and pretransport assessments of functional reliability of the cask. A proper mix of these constraints should yield risk parity with any existing transport cask

  3. Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

    International Nuclear Information System (INIS)

    Tadano, Tomoaki

    2016-01-01

    The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

  4. Materials issues in cask development

    International Nuclear Information System (INIS)

    Chapman, R.L.; Sorensen, K.B.

    1987-01-01

    This paper identifies potential new materials as a function of their use in the cask. To the extent that identified materials are not yet qualified for their intended application, this paper identifies probable technical issues and development efforts that may be required to qualify the materials for use in transportation casks. 1 tab

  5. 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.

  6. Development of neutron shielding material for cask

    International Nuclear Information System (INIS)

    Najima, K.; Ohta, H.; Ishihara, N.; Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    Since 1980's Mitsubishi Heavy Industries, Ltd (MHI) has established transport and storage cask design 'MSF series' which makes higher payload and reliability for long term storage. MSF series transport and storage cask uses new-developed neutron shielding material. This neutron shielding material has been developed for improving durability under high condition for long term. Since epoxy resin contains a lot of hydrogen and is comparatively resistant to heat, many casks employ epoxy base neutron shielding material. However, if the epoxy base neutron shielding material is used under high temperature condition for a long time, the material deteriorates and the moisture contained in it is released. The loss of moisture is in the range of several percents under more than 150 C. For this reason, our purpose was to develop a high durability epoxy base neutron shielding material which has the same self-fire-extinction property, high hydrogen content and so on as conventional. According to the long-time heating test, the weight loss of this new neutron shielding material after 5000 hours heating has been lower than 0.04% at 150 C and 0.35% at 170 C. A thermal test was also performed: a specimen of neutron shielding material covered with stainless steel was inserted in a furnace under condition of 800 C temperature for 30 minutes then was left to cool down in ambient conditions. The external view of the test piece shows that only a thin layer was carbonized

  7. 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)

  8. 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.

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

    International Nuclear Information System (INIS)

    Halstead, R. J.; Dilger, F.

    2003-01-01

    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

  12. Management of spent fuel from power and research reactors using CASTOR and CONSTOR casks and licensing experience worldwide

    International Nuclear Information System (INIS)

    Becher, D.

    2003-01-01

    An overview of the spent fuel storage in CASTOR and CONSTOR casks during the last 30 years is made. Design characteristics of the both types of casks are presented. CASTOR casks fulfill both the requirements for type B packages according to the IAEA requirements covering different accident situations in storage sites. Analyses of nuclear and thermal behavior and strength are carried out for CONSTOR concept. Special experimental program for verification of mechanical and thermomechanical properties is implemented. Licensing experience of the casks in German storage facilities is presented. Special modifications of CASTOR casks for WWER-440 and RBMK fuel assemblies have been designed for implementation in Eastern Europe. Contracts for GNB spent fuel casks delivery are concluded with Czech Republic, Slovakia, Hungary and Lithuania

  13. 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.

  14. Cask fleet operations study

    International Nuclear Information System (INIS)

    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

  15. FUEL CASK IMPACT LIMITER VULNERABILITIES

    International Nuclear Information System (INIS)

    Leduc, D.; England, J.; Rothermel, R.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Ingham, J.G.

    1980-01-01

    The IDENT 1578 container, which is a 110-in. long 5.5-in. OD tube, is designed for shipping FFTF fuel elements in T-3 casks between HEDL, HFEF, and other laboratories. The thermal analysis was conducted to evaluate whether or not the container satisfies its thermal design criteria

  17. 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)

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

    1990-12-01

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

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

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. Impact analysis of shipping casks

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.

    1989-01-01

    Shipping casks are being used in the United States Department of Energy to transport irradiated experiments, reactor fuel, radioactive waste, etc. One of the critical requirements in shipping cask analysis is the necessity to withstand severe impact environments. It is still conventional to develop the design and to verify the design requirements by hand calculations. Full three dimensional computations of impact scenarios have been performed but they are too expensive and time consuming for design purposes. Typically, on the order of more than an hour of CRAY time is required for a detailed, three dimensional analysis. The paper describes how simpler two- and three-dimensional models can be used to provide an intermediate level of detail between full three dimensional finite element calculations and hand calculations. The regulation that is examined here is: 10 CFR-71.73 hypothetical accident conditions, free drop. Free drop for an accident condition of a Class I package (approximate weight of 22,000 lb) is defined as a 30 foot drop onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. Three free drop scenarios are analyzed to assess the integrity of the cask when subjected to large bending and axial stresses. These three drop scenarios are: (1) a thirty foot axial drop on either end, (2) a thirty foot oblique angle drop with the cask having several different orientations from the vertical with impact on the top end cask corner, and (3) a thirty foot side drop with simultaneous impact on the strength of the various components that comprise the cask. The predicted levels of deformation and stresses in the cask will be used to assess the potential damage level. 5 refs., 5 figs., 1 tab

  1. Safety Assessment of a Metal Cask under Aircraft Engine Crash

    Directory of Open Access Journals (Sweden)

    Sanghoon Lee

    2016-04-01

    Full Text Available 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 freestanding 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.

  2. 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.

  3. Tip-over stability of spent fuel storage cask subjected to seismic load

    International Nuclear Information System (INIS)

    Shirai, Koji; Ito, Chihiro; Saegusa, Toshiari; Ryu, Hiroshi.

    1994-01-01

    Cask storage method is the most preferable to store spent fuel at a reactor site. Since the cask is vertically oriented within the storage facility, it is necessary to investigate the tipping-over stability of the cask subjected to seismic load. In this paper, an evaluation method of the seismic response of the cask subjected to earthquake motions is proposed. Firstly, shaking table tests were conducted using a scaled cask model on a reinforced concrete base, considering a similarity law. According to these test results, the characteristics of the tipping-over response of the cask under sinusoidal waves and natural earthquake waves were investigated, and the tipping-over stability of the cask subjected to seismic load was confirmed. Secondly, based on the assumption that the tipping-over behavior could be treated as a two-dimensional problem, the numerical simulations were carried out by means of distinct element method and compared with the experimental results. As a result, the applicability of this code for seismic response of the cask was made clear. (author)

  4. Spent fuel cask handling at an operating nuclear power plant

    International Nuclear Information System (INIS)

    Pal, A.C.

    1988-01-01

    The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices at all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant

  5. CASTORR 1000/19: Development and Design of a New Transport and Storage Cask

    International Nuclear Information System (INIS)

    Funke, Th.; Henig, Ch.

    2008-01-01

    The design of the new transport and storage cask type CASTOR R 1000/19 is presented in this paper. This cask was developed for the dry interim storage of spent VVER1000 fuel assemblies concerning the requirements of the Temelin NPP, Czech Republic. While the cask body is based on well-known ductile cast iron cask types with in-wall moderator, the basket follows a new concept. The basket is able to carry 19 fuel assemblies with a total decay heat power up to approximately 17 kW. The cask fulfils all requirements for a type B(U)F package. The main nuclear, mechanical and thermal properties of the cask are illustrated for normal conditions and for hypothetical accident scenarios during transport and storage. The main steps of the handling procedure such as loading the cask, drying the cavity and mounting the double lid system for tightness during interim storage are shown in principle. For this handling, boundary conditions at the NPP site such as dimensions, weight and the loading machine interface are considered. (authors)

  6. Status of spent fuel shipping cask development

    International Nuclear Information System (INIS)

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

    1989-01-01

    This paper discusses how several new-generation shopping cask systems are being developed for safe and economical transport of commercial spent nuclear fuel and other radioactive wastes for the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. Primary objectives of the from-reactor spent fuel cask development work are: to increase cask payloads by taking advantage of the increased at-reactor storage time under the current spent fuel management scenario, to facilitate more efficient cask handling operations with reduced occupational radiation exposure, and to promote standardization of the physical interfaces between casks and the shipping and receiving facilities. Increased cask payloads will significantly reduce the numbers of shipments, with corresponding reductions in transportation costs and risks to transportation workers, cask handling personnel, and the general public

  7. Two microcephaly-associated novel missense mutations in CASK specifically disrupt the CASK-neurexin interaction.

    Science.gov (United States)

    LaConte, Leslie E W; Chavan, Vrushali; Elias, Abdallah F; Hudson, Cynthia; Schwanke, Corbin; Styren, Katie; Shoof, Jonathan; Kok, Fernando; Srivastava, Sarika; Mukherjee, Konark

    2018-03-01

    Deletion and truncation mutations in the X-linked gene CASK are associated with severe intellectual disability (ID), microcephaly and pontine and cerebellar hypoplasia in girls (MICPCH). The molecular origin of CASK-linked MICPCH is presumed to be due to disruption of the CASK-Tbr-1 interaction. This hypothesis, however, has not been directly tested. Missense variants in CASK are typically asymptomatic in girls. We report three severely affected girls with heterozygous CASK missense mutations (M519T (2), G659D (1)) who exhibit ID, microcephaly, and hindbrain hypoplasia. The mutation M519T results in the replacement of an evolutionarily invariant methionine located in the PDZ signaling domain known to be critical for the CASK-neurexin interaction. CASK M519T is incapable of binding to neurexin, suggesting a critically important role for the CASK-neurexin interaction. The mutation G659D is in the SH3 (Src homology 3) domain of CASK, replacing a semi-conserved glycine with aspartate. We demonstrate that the CASK G659D mutation affects the CASK protein in two independent ways: (1) it increases the protein's propensity to aggregate; and (2) it disrupts the interface between CASK's PDZ (PSD95, Dlg, ZO-1) and SH3 domains, inhibiting the CASK-neurexin interaction despite residing outside of the domain deemed critical for neurexin interaction. Since heterozygosity of other aggregation-inducing mutations (e.g., CASK W919R ) does not produce MICPCH, we suggest that the G659D mutation produces microcephaly by disrupting the CASK-neurexin interaction. Our results suggest that disruption of the CASK-neurexin interaction, not the CASK-Tbr-1 interaction, produces microcephaly and cerebellar hypoplasia. These findings underscore the importance of functional validation for variant classification.

  8. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

    Griesmeyer, J.M.; Drotning, W.D.; Morimoto, A.K.; Bennett, P.C.

    1990-10-01

    Remote automated cask handling has the potential to reduce both the occupational exposure and the time required to process a nuclear waste transport cask at a handling facility. The ongoing Advanced Handling Technologies Project (AHTP) at Sandia National Laboratories is described. AHTP was initiated to explore the use of advanced robotic systems to perform cask handling operations at handling facilities for radioactive waste, and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof-of-concept robotic systems developed in AHTP are intended to extrapolate from currently available commercial systems to the systems that will be available by the time that a repository would be open for operation. The project investigates those cask handling operations that would be performed at a nuclear waste repository facility during cask receiving and handling. The ongoing AHTP indicates that design guidance, rather than design specification, is appropriate, since the requirements for robotic handling do not place severe restrictions on cask design but rather focus on attention to detail and design for limited dexterity. The cask system design features that facilitate robotic handling operations are discussed, and results obtained from AHTP design and operation experience are summarized. The application of these design considerations is illustrated by discussion of the robot systems and their operation on cask feature mock-ups used in the AHTP project. 11 refs., 11 figs

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  10. Studies of natural convection heat transfer in dry spent fuel storage casks using the cobra-SFS thermal-hydraulic computer code

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

    The purpose of this study is to identify the importance of natural convection cooling within a nuclear dry spent fuel storage system. In the past, applicants submitting requests to the United States Nuclear Regulatory Commission (USNRC) for a license for a dry spent fuel storage system design did not rigorously treat natural convection within the fuel package of the dry storage system. Typically, the applicant applies heat transfer correlations that raise the thermal conductivity of the materials (gas and solid structures) to account for the impact of convection on the thermal performance of the system. (author)

  11. Conceptual design and cost estimation of dry cask storage facility for spent fuel

    International Nuclear Information System (INIS)

    Maki, Yasuro; Hironaga, Michihiko; Kitano, Koichi; Shidahara, Isao; Shiomi, Satoshi; Ohnuma, Hiroshi; Saegusa, Toshiari

    1985-01-01

    In order to propose an optimum storage method of spent fuel, studies on the technical and economical evaluation of various storage methods have been carried out. This report is one of the results of the study and deals with storage facility of dry cask storage. The basic condition of this work conforms to ''Basic Condition for Spent Fuel Storage'' prepared by Project Group of Spent Fuel Dry Storage at July 1984. Concerning the structural system of cask storage facilities, trench structure system and concrete silo system are selected for storage at reactor (AR), and a reinforced concrete structure of simple design and a structure with membrance roof are selected for away from reactor (AFR) storage. The basic thinking of this selection are (1) cask is put charge of safety against to radioactivity and (2) storage facility is simplified. Conceptual designs are made for the selected storage facilities according to the basic condition. Attached facilities of storage yard structure (these are cask handling facility, cask supervising facility, cask maintenance facility, radioactivity control facility, damaged fuel inspection and repack facility, waste management facility) are also designed. Cost estimation of cask storage facility are made on the basis of the conceptual design. (author)

  12. ANSI N14.5 source term licensing of spent-fuel transport cask containment

    International Nuclear Information System (INIS)

    Seager, K.D.; Reardon, P.C.; James, R.J.; Foadian, H.; Rashid, Y.R.

    1993-01-01

    American National Standards Institute (ANSI) standard N14.5 states that ''compliance with package containment requirements shall be demonstrated either by determination of the radioactive contents release rate or by measurement of a tracer material leakage rate.'' The maximum permissible leakage rate from the transport cask is equal to the maximum permissible release rate divided by the time-averaged volumetric concentration of suspended radioactivity within the cask. The development of source term methodologies at Sandia National Laboratories (SNL) provides a means to determine the releasable radionuclide concentrations within spent-fuel transport casks by estimating the probability of cladding breach, quantifying the amount of radioactive material released into the cask interior from the breached fuel rods, and quantifying the amount of radioactive material within the cask due to other sources. These methodologies are implemented in the Source Term Analyses for Containment Evaluations (STACE) software. In this paper, the maximum permissible leakage rates for the normal and hypothetical accident transport conditions defined by 10 CFR 71 are estimated using STACE for a given cask design, fuel assembly, and initial conditions. These calculations are based on defensible analysis techniques that credit multiple release barriers, including the cladding and the internal cask walls

  13. Seismic Performance of Dry Casks Storage for Long- Term Exposure

    Energy Technology Data Exchange (ETDEWEB)

    Ibarra, Luis [Univ. of Utah, Salt Lake City, UT (United States); Sanders, David [Univ. of Nevada, Reno, NV (United States); Yang, Haori [Oregon State Univ., Corvallis, OR (United States); Pantelides, Chris [Univ. of Utah, Salt Lake City, UT (United States)

    2016-12-30

    The main goal of this study is to evaluate the long-term seismic performance of freestanding and anchored Dry Storage Casks (DSCs) using experimental tests on a shaking table, as well as comprehensive numerical evaluations that include the cask-pad-soil system. The study focuses on the dynamic performance of vertical DSCs, which can be designed as free-standing structures resting on a reinforced concrete foundation pad, or casks anchored to a foundation pad. The spent nuclear fuel (SNF) at nuclear power plants (NPPs) is initially stored in fuel-storage pools to control the fuel temperature. After several years, the fuel assemblies are transferred to DSCs at sites contiguous to the plant, known as Interim Spent Fuel Storage Installations (ISFSIs). The regulations for these storage systems (10 CFR 72) ensure adequate passive heat removal and radiation shielding during normal operations, off-normal events, and accident scenarios. The integrity of the DSCs is important, even if the overpack does not breach, because eventually the spent fuel-rods need to be shipped either to a reprocessing plant or a repository. DSCs have been considered as a temporary storage solution, and usually are licensed for 20 years, although they can be relicensed for operating periods of up to 60 years. In recent years, DSCs have been reevaluated as a potential mid-term solution, in which the operating period may be extended for up to 300 years. At the same time, recent seismic events have underlined the significant risks DSCs are exposed. The consideration of DCSs for storing spent fuel for hundreds of years has created new challenges. In the case of seismic hazard, longer-term operating periods not only lead to larger horizontal accelerations, but also increase the relative effect of vertical accelerations that usually are disregarded for smaller seismic events. These larger seismic demands could lead to casks sliding and tipping over, impacting the concrete pad or adjacent casks. The casks

  14. Development of cask and transportation system

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Do, Jae Bum; Lee, Heung Young

    1990-03-01

    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)

  15. 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.

  16. Radioactive fuel cask railcar humping study

    International Nuclear Information System (INIS)

    Wilson, L.T.

    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 1 / 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

  17. Transportation capabilities of the existing cask fleet

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  18. Transportation capabilities of the existing cask fleet

    International Nuclear Information System (INIS)

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

    1991-01-01

    This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 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. Life cycle cost report of VHLW cask

    International Nuclear Information System (INIS)

    1995-06-01

    This document, the Life Cycle Cost Report (LCCR) for the VHLW Cask, presents the life cycle costs for acquiring, using, and disposing of the VHLW casks. The VHLW cask consists of a ductile iron cask body, called the shielding insert, which is used for storage and transportation, and ultimately for disposal of Defense High Level Waste which has been vitrified and placed into VHLW canisters. Each ductile iron VHLW shielding insert holds one VHLW canister. For transportation, the shielding insert is placed into a containment overpack. The VHLW cask as configured for transportation is a legal weight truck cask which will be licensed by NRC. The purpose of this LCCR is to present the development of the life cycle costs for using the VHLW cask to transport VHLW canisters from the generating sites to a disposal site. Life cycle costs include the cost of acquiring, operating, maintaining, and ultimately dispositioning the VHLW cask and its associated hardware. This report summarizes costs associated with transportation of the VHLW casks. Costs are developed on the basis of expected usage, anticipated source and destination locations, and expected quantities of VHLW which must be transported. DOE overhead costs, such as the costs associated with source and destination facility handling of the VHLW, are not included. Also not included are costs exclusive to storage or disposal of the VHLW waste

  20. Used Fuel Cask Identification through Neutron Profile

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  1. Concrete spent fuel storage casks dose rates

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Trontl, K.

    1998-01-01

    Our intention was to model a series of concrete storage casks based on TranStor system storage cask VSC-24, and calculate the dose rates at the surface of the casks as a function of extended burnup and a prolonged cooling time. All of the modeled casks have been filled with the original multi-assembly sealed basket. The thickness of the concrete shield has been varied. A series of dose rate calculations for different burnup and cooling time values have been performed. The results of the calculations show rather conservative original design of the VSC-24 system, considering only the dose rate values, and appropriate design considering heat rejection.(author)

  2. Burnup credit in the design of spent-fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Wilmot, E.L.

    1987-01-01

    The spent-fuel carrying capacities of previous generation shipping casks have been primarily thermal and/or shielding limited. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus, a considerable weight margin is available to the designer for increasing the payload capacity. One method of achieving an increase in capacity is by reducing fuel assembly spacing. The amount of reduction in assembly spacing available is limited by criticality and fuel support structural concerns. The optimum fuel assembly achievable is then limited by requirements to control neutron multiplication and to ensure the structural integrity of fuel support components. An investigation of the feasibility of accounting for fuel burnup in the design of spent-fuel shipping casks was recently completed for the US Dept. of Energy's Office of Civilian Radioactive Waste Management. Criticality analyses have been performed, and potential impacts in terms of increased cask capacities with associated costs and safety benefits have been determined. The scope of the analysis is limited to typical burnups of full-cycle, discharged pressurized water reactor (PWR) fuel and generic shipping cask designs. A sensitivity analysis was performed to estimate the impact of cask capacity on total transportation system life cycle costs

  3. Evaluation of the salt deposition on the canister surface of concrete cask. Part 3. Long-term measurement of salt concentration in air and evaluation of the salt deposition

    International Nuclear Information System (INIS)

    Wataru, Masumi; Takeda, Hirofumi

    2015-01-01

    To realize the dry storage using concrete cask in Japan, it is important to develop the evaluation method of the SCC of the canister. One of the key issues is sea salt deposition on the canister surface during the storage period. If the amount of salt deposition exceeds the critical value, the SCC may occur. The amount of salt deposition depends on the ambient air condition. We developed the measurement device of salt in air to make clear the ambient condition. The device sucks the air including sea salt and the sea salt dissolves in water. We analyze the water including sea salt. This device works automatically for one or two months. In this study, the performance of this device was verified comparing the data obtained by the air sampler using filter pack. In Yokosuka area of CRIEPI, we measured the ambient air condition using this device for three years. Furthermore, we performed the salt deposition test using the small ducts in the same area. The ambient air including sea salt flows in the duct and the sea salt deposits on the test specimen put on the duct inner surface. We took out the specimen after certain time and measured the salt amount on the test specimen. Using these data, we obtained the relation between the salt deposition and the time on this ambient condition. The results of this study are useful to evaluate the SCC of the canister. (author)

  4. Study on heat removal capability concrete cask system with horizontal orientation

    International Nuclear Information System (INIS)

    Nabemoto, Toyonobu; Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatsugu

    2002-01-01

    In Japan, nuclear fuel cycle, has been promoted, so the recycle fuels formed at nuclear power stations are planned to be processed at reprocessing facilities in future. However, as forming quantities of the recycle fuels are more than reprocessing quantities of the facilities, it is needed to practice a facility (interim storage facility (ISF)) to temporarily store them among the recycle fuels will be reprocessed. The Ishikawajima-Harima Heavy Industries, Co., Ltd. has investigated on vault system and concrete cask system for dry storage system with excellent economical efficiency among various systems on ISFs. As the latter method has a number of actual results in U.S.A., its practice is progressed after some improvements suitable for Japan. When progressing this practice on the latter method on fiscal year 1999, at first, a concrete cask with actual size was experimentally produced, to confirm its productivity. On fiscal year 2000, aiming to establish heat removal evaluation at storage, a thermal load test simulated at the storage was carried out by using this trial product. Here was reported results obtained at a test simulated at repacking carried out on fiscal year 2001. (G.K.)

  5. Underground transportation and handling system for Pollux-casks

    International Nuclear Information System (INIS)

    Schrimpf, C.

    1988-01-01

    The concept for the underground transportation and handling system for Pollux-casks was optimized in a first phase by dividing the process in the repository up into the several transportation and manipulation steps. For each step, the possibilities were described and evaluated by means of a list of criteria (technical, safety and economical criteria). The following concept for the transportation and handling was developed: The casks are transported to the unloading area of the surface facilities by railway or truck. After removal of the transport protection, the entry control is performed. The cask is lifted from the vehicle and placed on a railbound transportation vehicle. This transport unit is transferred to the shaft and placed there ready for shaft hoisting. With the hoisting cage protruding, the transport unit is placed on the hoisting cage by means of a pushing-on device, locked, and then conveyed underground. After arrival on the emplacement level, the transport unit is pulled-off from the hoisting cage and taken over by a mine locomotive and transferred through the transportation and access drifts as far as to the emplacement site. There the locomotive pushed the rail transport vehicle into the emplacement drift, as far as to the designated emplacement position. At the emplacement position, the cask is again lifted by means of hoisting equipment. The rail transport vehicle is pulled out of the emplacement drift and returned to the surface for reloading. After deposition of the cask on the drift floor, the emplacement equipment is pulled back in order to give the operation space free for the slinger backfill truck. Within preceding tests two different backfilling techniques were investigated under realistic conditions: pneumatic backfilling and slinger backfilling. The slinger truck was found to be the most suitable for the designated purpose

  6. Annex IV. The technical and logistic benefits of non-uniform, zoned cask loading

    International Nuclear Information System (INIS)

    2006-01-01

    The purpose of this appendix is to describe the benefits of licensing and using non-uniform, zoned loading of casks, including the physical nature of the phenomena that underlie those benefits. Based on the systematics of the zoned loading analysis sequence, this appendix also outlines a regulatory approach for licensing and specifying the range of couplings of the outer and inner zone fuel characteristics that result in total external dose rates being at the regulatory limit. This Appendix has outlined an approximate method for evaluating the capability of zone loaded casks, and has used that method to evaluate zoned loading in a typical long term shipping situation. The results of the evaluation indicate that there are two types of benefit arising from the use of zoned cask loading when coupled with an optimised long term plan for fuel selection to accomplish the loadings. A technical benefit in which the radioactivity content of a cask is increased without an increase in the external dose rate, and a logistic benefit, realised through the use of an appropriate long term fuel selection and cask-loading plan, that significantly extends the usability of a cask design, delivers shipments with characteristics that are fairly stable over time, and is consistently loaded close to its license limit

  7. Force reconstruction for the slapdown test of a nuclear transportation cask

    International Nuclear Information System (INIS)

    Bateman, V.I.; Carne, T.G.; Gregory, D.L.; Attaway, S.W.; Yoshimura, H.R.

    1989-01-01

    Two force reconstruction techniques were used to evaluate the slapdown response of a 1/3 scale model solid steel, spent fuel cask dropped 30 ft onto an unyielding target. The two techniques are: the sum of weighted acceleration technique (SWAT) and the deconvolution technique (DECON). A brief description and the calibration of the techniques as applied to the cask are presented. For the slapdown test, both techniques yielded very similar resultant forces and provided more accurate definition of the force-time history for the cask than is available from conventional data reduction methods. An applied moment, a measurement previously unobtainable from conventional cask accelerometer data reduction techniques, was determined with SWAT. The angular velocity calculated with SWAT was verified with photometric measurements. 9 refs., 22 figs

  8. Design review report FFTF interim storage cask

    International Nuclear Information System (INIS)

    Scott, P.L.

    1995-01-01

    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

  9. Cask ownership: Options and strategic factors

    International Nuclear Information System (INIS)

    Smith, C.W.

    1986-01-01

    Because of the potential number of casks available through utility modular storage programs, it is imperative that the planning for the provision and operation of casks under the NWPA program include consideration of the utility owned casks. As to the remainder of the cask requirements for implementation of the NWPA, the author believes that the cost factor is an artificial one for determining the benefits to the taxpayers and ratepayers for cask ownership and that the decision should be made on the basis of capability of the industry to perform on a competitive bid basis and assurance that the shipments will be made on a timely, safe and cost effective basis. If the procurement process is structured to rally permit competitive bidding on spent fuel shipping services, the competition in the market place will assure that DOE and the ratepayers, receive safe, high quality, and cost effective transportation proposals from very capable companies

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

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. TITAN Legal Weight Truck cask preliminary design report

    International Nuclear Information System (INIS)

    1990-04-01

    The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of the document includes the LWT cask with fuel baskets; impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. 75 figs., 48 tabs

  13. Small gas leakage rate measuring and monitoring system for spent fuel transport/storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Asano, Ryoji [Hitachi Zosen Corp., Tokyo (Japan); Aritomi, Masanori; Matsuzaki, Mitsuo; Kawa, Tsunemichi

    1997-12-01

    A containment function of transport and/or storage casks of radioactive materials is essential to prevent the materials from being released excessively into the environment. It is not practical for containment tests to measure directly the radioactivity release so that gas volumetric leakage rates are usually assessed and gas pressure decrease or increase method is usually applied. As gas flow model for evaluation, the ISO standards has deleted the concept of choked flow which is adopted by ANSI N14.5. Provided that the choked flow is not adopted to the leakage rate evaluation, the criteria of the test should be severe, and a new leakage rate measuring system with high accuracy and reasonable measuring time is required. Transport casks are often inspected in a temporary cask-storage facility where simultaneous measurement of many casks is required. In a storage cask system, multiple casks are monitored on their containment function during a storage period, and the method for simultaneous monitoring at many points for long term is required. In this study, two kinds of small gas leakage rate measuring systems are developed. One is to measure gas leakage rates directly and is called `flow measuring system`, which can measure gas leakage rate of 10{sup -4} to 10{sup -2} cm{sup 3}/s with high accuracy and short measuring time. The other is to measure the pressure decreasing rate and is called `pressure decreasing rate measuring system`, which can monitor the pressure change at many points simultaneously. (author)

  14. Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

    International Nuclear Information System (INIS)

    Weiss, M.

    2001-01-01

    As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

  15. FFTF disposable solid waste cask

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-01-01

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

  16. FFTF disposable solid waste cask

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Sensitivity analyses of seismic behavior of spent fuel dry cask storage systems

    International Nuclear Information System (INIS)

    Luk, V.K.; Spencer, B.W.; Shaukat, S.K.; Lam, I.P.; Dameron, R.A.

    2003-01-01

    Sandia National Laboratories is conducting a research project to develop a comprehensive methodology for evaluating the seismic behavior of spent fuel dry cask storage systems (DCSS) for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission (NRC). A typical Independent Spent Fuel Storage Installation (ISFSI) consists of arrays of free-standing storage casks resting on concrete pads. In the safety review process of these cask systems, their seismically induced horizontal displacements and angular rotations must be quantified to determine whether casks will overturn or neighboring casks will collide during a seismic event. The ABAQUS/Explicit code is used to analyze three-dimensional coupled finite element models consisting of three submodels, which are a cylindrical cask or a rectangular module, a flexible concrete pad, and an underlying soil foundation. The coupled model includes two sets of contact surfaces between the submodels with prescribed coefficients of friction. The seismic event is described by one vertical and two horizontal components of statistically independent seismic acceleration time histories. A deconvolution procedure is used to adjust the amplitudes and frequency contents of these three-component reference surface motions before applying them simultaneously at the soil foundation base. The research project focused on examining the dynamic and nonlinear seismic behavior of the coupled model of free-standing DCSS including soil-structure interaction effects. This paper presents a subset of analysis results for a series of parametric analyses. Input variables in the parametric analyses include: designs of the cask/module, time histories of the seismic accelerations, coefficients of friction at the cask/pad interface, and material properties of the soil foundation. In subsequent research, the analysis results will be compiled and presented in nomograms to highlight the sensitivity of seismic response of DCSS to

  18. Performance of CASTORR HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    International Nuclear Information System (INIS)

    Wilmsmeier, Marco; Vossnacke, Andre

    2008-01-01

    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 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 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 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 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 as well. GNS long-term CASTOR R

  19. A cask maintenance facility feasibility study

    International Nuclear Information System (INIS)

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

    1989-01-01

    The Oak Ridge National Laboratory (ORNL) is supporting the USDOE Office of Civilian Radioactive Waste Management (OCRWM) in developing a transportation system for spent nuclear fuel (SNF) and defense high level waste (HLW) as a part of the Federal Waste Management System (FWMS). In early 1988, a feasibility study was undertaken to design a stand-alone, green field facility for maintaining the FWMS casks. The feasibility study provided an initial layout facility design, an estimate of the construction cost, and an acquisition schedule for a Cask Maintenance Facility (CMF). The study also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrieveable Storage (MRS) facility. The data, design, and estimated cost resulting from the study have been organized for use in the total transportation system decision-making process. Most importantly, the feasibility study also provides a foundation for continuing design and planning efforts. The feasibility study was based on an assumed stand-alone green field configuration because of the flexibility this design approach provides. A stand-alone facility requires the inclusion with support functions as well as the primary process facilities thus yielding a comprehensive design evaluation and cost estimate. For example, items such as roads, security and waste processing which might be shared with an integrated or collocated facility have been fully costed in the feasibility study. Thus, while the details of the facility design might change, the overall concept used in the study can be applied to other facility configurations as planning for the total FWMS develops

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

    International Nuclear Information System (INIS)

    Quercetti, T.; Ballheimer, V.; Zeisler, P.; Mueller, K.

    2003-01-01

    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

  1. Utility oversight of Cask System Development Program

    International Nuclear Information System (INIS)

    Vincent, J.A.; Jordan, J.M.; Schwartz, M.H.

    1993-01-01

    This paper will present the electric utility industry's perspective on the status and scope of the DOE's Office of Civilian Radioactive Waste Management's (DOE/OCRWM) transportation cask systems development activities, including the Cask Systems Development Program (CSDP) Initiative I transportation cask projects. This presentation is particularly timely because the CSDP Independent Management Review Group (IMRG), os which one of the authors is a member, completed an objective assessment of OCRWM's transportation cask system development activities and issued its first report in late August 1992. The perspective on these cask systems development activities that will be presented reflects conclusions based on (1) the industry's review of CSDP Preliminary and Draft Final Design Reports for the Initiative I cask projects, (2) the activities of one of the authors as a member of the IMRG, and (3) the positions that the industry has consistently taken on what it believes to be the appropriate scope and pace of the CSDP and its integration with other OCRWM activities. Background information on the OCRWM transportation cask systems development activities and the relevant industry activities will also be provided

  2. Burnup credit effect on proposed cask payloads

    International Nuclear Information System (INIS)

    Hall, I.K.

    1989-01-01

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

  3. Ageing characteristics of the VSC-17 concrete cask storing spent nuclear fuel for 15 years

    International Nuclear Information System (INIS)

    Shirai, Koji; Saegusa, Toshiari; Matsumura, Takuro; Sasahara, Akihiro; Hattori, Takatoshi; Morton, Sheryl L.; Winston, P.L.

    2009-01-01

    For long-term use of a concrete cask, radiation and environmental effects may cause chemical and physical alteration of the concrete that could result in excessive cracking, spalling, and loss of strength. To determine the durability performance of a concrete cask, the Ventilated Storage Cask (VSC-17) was identified as a candidate, because it has been used to store fuel as part of the dry cask storage demonstration project for more than 15 years in the Idaho National Laboratory (INL). Properties of the concrete material, surface temperature, dose rates and the concrete strength were surveyed by NDE (Non-Destructive Examination) methods. Moreover, the surface condition of the canister installed inside the cask was surveyed by optical fiber-scopes and its was found that the canister had no visual degradation. However, as a localized hot spot beneath an upper air vent was detected, an ultrasonic testing was conducted to investigate the distribution of the wave propagation velocity. As a result, it appears that the concrete beneath the upper air vent has a low wave propagation velocity, indicating that there may be reduction in its density due to the water dissipation under the severe thermal loading. This fast indicated that well-draining of the bleeding water during placing concrete process of the fresh concrete in the manufacture stage is essential. (author)

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

    International Nuclear Information System (INIS)

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

    1988-10-01

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

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

    International Nuclear Information System (INIS)

    Saueressig, P.T.

    1994-01-01

    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

  6. 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.

  7. Full-scale Simulation of a Dry Storage Cask by Computational Fluid Dynamics

    International Nuclear Information System (INIS)

    Yoo, Seung Hun; Kim, Hyeun Min; No, Hee Cheon

    2010-01-01

    The prediction of flow and heat transfer in the dry storage system has been carried out to meet the need for thermal-hydraulic analysis for cask designs or to investigate local and global phenomena which could not be measured by the experiments. Computational fluid dynamics (CFD) codes have been used for solving heat transfer inside the complex geometries such as spent fuel assemblies. And more accurate calculations were performed by them compared with those of one dimensional system codes. In order to provide for the reliability and the guidance of the use of CFD codes, diverse researches were performed with different dry systems and codes. In the present study, TN24P cask was selected for the reference experiment. FLUENT code with a set of CFD models was used for the fullscale simulation of TN24P cask and its results were compared with the experimental data and COBRA-SFS results

  8. CSER 94-014: Storage of metal-fuel loaded EBR-II casks in concrete vault on PFP grounds

    International Nuclear Information System (INIS)

    Hess, A.L.

    1994-01-01

    A criticality safety evaluation is presented to permit EBR-2 spent fuel casks loaded with metallic fuel rods to be stored in an 8-ft diameter, cylindrical concrete vault inside the PFP security perimeter. The specific transfer of three casks with Pu alloy fuel from the Los Alamos Molten Plutonium Reactor Experiment from the burial grounds to the vault is thus covered. Up to seven casks may be emplaced in the casing with 30 inches center to center spacing. Criticality safety is assured by definitive packaging rules which keep the fissile medium dry and at a low effective volumetric density

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-01

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

  10. Dynamic impact characteristics of KN-18 SNF transport cask - Part 1: An advanced numerical simulation and validation technique

    International Nuclear Information System (INIS)

    Kim, Kap-Sun; Kim, Jong-Soo; Choi, Kyu-Sup; Shin, Tae-Myung; Yun, Hyun-Do

    2010-01-01

    Domestic and international regulations for the transportation of radioactive materials strictly prescribe the design requirements for spent nuclear fuel (SNF) transport casks. According to the applicable codes, a transport cask must withstand a free-drop impact of 9 m onto an unyielding surface and a free-drop impact of 1 m onto a mild steel bar. However, the structural performance of a transport cask is not easy to evaluate precisely because the dynamic impact characteristics of the cask, which includes impact limiters to absorb the impact energy, are so complex. In this study, a more advanced and applicable numerical simulation method using the finite element (FE) method via the commercial FE code LS-DYNA is proposed and verified against the experimental results for a 1/3-scale model of the KN-18 SNF transport cask, recently developed in Korea. In addition, the detailed dynamic impact characteristics of the transport cask under free-drop conditions are investigated via the proposed numerical simulation method and actual drop tests to improve the accuracy and optimization of the SNF transport cask design.

  11. Criticality safety assessment of WWER-1000 spent fuel cask

    International Nuclear Information System (INIS)

    Apostolov, T.; Manolova, M.; Prodanova, R.

    2001-01-01

    A methodology that allows taking credit for burnup in the criticality safety analysis of WWER spent fuel casks is presented. It is based on the two world well known and used code systems:NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The results of criticality calculations of WWER-1000 spent fuel storage and transportation cask, applying burnup credit is shown. The depletion calculations have been carried out for three types of WWER-1000 fuel assemblies (with enrichment of 3.0%, 3.3% and 3.3% profiled) by modelling the real operational history of the first three fuel cycles at unit 6, Kozloduy NPP. The criticality calculational model has been developed on the basis of real fuel cask, designed by the Izorskie zavody. The results obtained show that the criticality safety criterion K eff < 0.95 is satisfied for both fresh and spent fuel. Besides, the implementation of burnup credit accounts for the reduced reactivity of spent fuel and allows evaluating the conservatism of the fresh fuel assumption. (author)

  12. Dynamic fracture toughness data for CASTOR {sup registered} casks

    Energy Technology Data Exchange (ETDEWEB)

    Winkler, H.P. [GNS Gesellschaft fuer Nuklear-Service mbH/GNB, Essen (Germany); Trubitz, P.; Pusch, G. [Technische Univ. Bergakademie Freiberg, Freiberg (Germany); Warnke, E.P. [Siempelkamp GmbH and Co. KG, Krefeld (Germany); Beute, K. [Gontermann-Peipers GmbH, Siegen (Germany); Novotny, V. [SKODA, HUTE, Plzen (Czech Republic)

    2004-07-01

    For the use of cast iron spherical graphite for heavy-sectioned casks for transportation and storage of radiactive materials a complete failure assessment including fracture mechanical analysis is necessary. The casks require an elaborate fracture mechanics design based on fracture mechanics evaluation. The extension of the existing code with respect to dynamic loading takes account new developments to extend the field of applications. It also includes new criteria to design these casks against operating and accident loadings. A fundamental requirement for the realisation of this standard and the calculation of admissible crack lengths of stresses under dynamic loads is the availability of fracture mechanical data. The paper presents-as a part of a large test-program-first results of dynamic fracture-toughness-investigations depending on structure and temperature. The test-program will incorporate investigations on more then 2500 specimens. The investigations that will be done include static and dynamic fracture mechanics tests, dynamic tensile and pressure-tests on different formed specimens. The temperatures and other test conditions follows the IAEA-regulations and the real service conditions. The test-program will be realised in partnership with different institutes.

  13. A cask maintenance facility feasibility study

    International Nuclear Information System (INIS)

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

    1989-01-01

    The Oak Ridge National Laboratory (ORNL) is developing a transportation system for spent nuclear fuel (SNF) and defense high level waste (HLW) as a part of the Federal Waste Management System (FWMS). In early 1988, a feasibility study was undertaken to design a stand-alone, ''green field'' facility for maintaining the FWMS casks. The feasibility study provided an initial layout facility design, an estimate of the construction cost, and an acquisition schedule for a Cask Maintenance Facility (CMF). The study also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrievable Storage (MRS) facility. The data, design, and estimated costs resulting from the study have been organized for use in the total transportation system decision-making process. Most importantly, the feasibility study also provides a foundation for continuing design and planning efforts. Fleet servicing facility studies, operational studies from current cask system operators, a definition of the CMF system requirements, and the experience of others in the radioactive waste transportation field were used as a basis for the feasibility study. In addition, several cask handling facilities were visited to observe and discuss cask operations to establish the functions and methods of cask maintenance expected to be used in the facility. Finally, a peer review meeting was held at Oak Ridge, Tennessee in August, 1988, in which the assumptions, design, layout, and functions of the CMF were significantly refined. Attendees included representatives from industry, the repository and transportation operations

  14. Force and moment reconstruction for a nuclear transportation cask using sum of weighted accelerations and deconvolution theory

    International Nuclear Information System (INIS)

    Yoshimura, H.R.; Bateman, V.; Carne, T.G.; Gregory, D.L.; Attaway, S.W.; Bronowski, D.R.

    1989-01-01

    A 9-m drop test was conducted of a 1/3-scale-model spent fuel cask onto an unyielding target. The structural response of the impact limiters and attachments was evaluated. A mass model of the cask body, with steel-sheathed redwood and balsa impact limiters, was tested in a 10-degree slapdown orientation. One end of the cask impact the target before the other end, with higher deceleration forces resulting from the second impact. The information desired from this test is the deformation of the two impact limiters on either end of the cask as a function of the applied force. The content in this paper will only discuss a summary of the applied force calculations. Additional details about the force and moment reconstruction methods and analysis results and test and hardware are provided elsewhere. Two new force reconstruction techniques were applied to the slapdown test data: the sum of weighted accelerations technique (SWAT) and deconvolution (DECON). The rigid-body acceleration is then multiplied by the cask mass to obtain an estimate of the applied force. The frequency content of this force is restricted to the cut-off frequency of the digital filter, typically about one-half of the lowest elastic mode of the cask. The new force reconstruction techniques demonstrate the potential for a better estimate of forces acting on the cask during the impact than the conventional method. The new force reconstruction techniques use the cask structure as a generalized force transducer. With these techniques, the elastic vibration response of the cask is eliminated from the acceleration data. The main advantages of the force reconstruction techniques are the extension of the frequency bandwidth (due to the elimination of the elastic modal response in that bandwidth) and the preservation of the force rise time

  15. Experimental and analytical study for demonstration program on shielding of casks for high-level wastes

    International Nuclear Information System (INIS)

    Ueki, K.; Nakazawa, M.; Hattorl, S.; Ozaki, S.; Tamaki, H.; Kadotani, H.; Ishizuka, T.; Ishikawa, S.

    1993-01-01

    The following remarks were obtained from the experiment and the DOT 3.5 and the MCNP analyses on the gamma ray and the neutron dose equivalent rates in the cask of interest. 1. The cask has thinner neutron shielding parts around the trunnions. Significant neutrons streaming around the trunnion parts was observed which was also cleared by the MCNP analysis for the 252 Cf source experiment. Accordingly, detailed neutron streaming calculations are required to evaluate the dose levels around the trunnions when loading the vitrified high-level wastes. 2. The room-scattered obstructive neutrons, mainly originating from the neutrons penetrating around the trunnions, at the top and the bottom of the cask are reduced significantly by preparing the water tank at the top and the water layer at the bottom. Therefore, a more accurate experiment is to be carried out in the future shielding experiment especially for neutrons. However, because the water tank and the layer do not exist in the actual high-level wastes transport cask, the experiment without the water tank and layer are not dispensable to demonstrate the transport conditions of the actual cask, too. 3. The gamma-ray and the neutron dose equivalent rate distributions obtained from the DOT 3.5 and the MCNP calculations, respectively, agreed closely with the measured values in the cask areas of interest. Accordingly, the DOT 3.5 code and the MCNP code with the NESX estimator can be employed not only for the shielding analysis of the future experiments, but also for making a safety analysis report of high-level wastes transport casks. (J.P.N.)

  16. Advanced handling technology project and implications for cask design

    International Nuclear Information System (INIS)

    Griesmeyer, J.M.; Bennett, P.C.; Sanders, T.L.

    1989-01-01

    This paper describes the results of the ongoing Advanced Handling Technologies Project (AHTP) at Sandia. AHTP was initiated in 1986 to explore the use of advanced robotic systems to perform cask handling operations at radioactive waste handling facilities and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof of concept systems developed in AHTP are intended to extrapolate from currently available commercial systems to those that would be available by the time than an actual repository would be open for operation. These systems provide test facilities for the investigation of the robotic handling of alternate cask design features. The following sections describe (1) the approach used in AHTP to select operations for proof of concept robotic systems and to identify the cask design implications, (2) the separate proof of concept robotic systems developed in AHTP, and (3) preliminary insights into the impact of cask system design features on the feasibility of robotic performance of cask handling operations. The main conclusions from AHTP to date regarding design for remote handling are: (1) incorporation of cask system design features which facilitate robotic cask handling can be achieved with minimal impact on cask functional features, (2) proper cask design allows robotic cask handling operations from manipulation of cask tie-down mechanisms to radiation surveys to be performed quickly and reliably without direct human intervention, and (3) design for remote handling also facilitates manual handling operations

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-12-01

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

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

    International Nuclear Information System (INIS)

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

    1986-12-01

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

  20. Shielding test of a model for FBR irradiation fuel transport cask

    International Nuclear Information System (INIS)

    Ohashi, A.; Ueki, K.; Iyori, I.; Uruwashi, S.; Iwanaga, S.; Takahashi, S.

    1993-01-01

    Shielding experiments and their Monte Carlo analyses of the half scale model of the PIE cask were carried out to 1. obtain the dose rate distributions around the model and deduce the characteristics of the actual cask dose distribution from the results, 2. examine the propriety of calculational techniques and the accuracy of Monte Carlo codes to be used in the shielding design or the analysis of the actual cask. The following remarks were obtained through the present study. 1. The C/E values were good for almost all the detector locations except for a few particular points. 2. The calculation geometry with scatterable materials around the cask is necessary to derive the details of dose distribution. 3. The solution in the thermal expansion room must be taken into account and added in the calculation geometry. 4. The magnitude of the dose rates of secondary gamma rays is approximately one-fifth of those of the neutron of the half model. Two peaks must be paid attention at 2.2 MeV and 7.6 MeV due to the (n,γ) reaction of hydrogen and iron, respectively. Hence the calculational techniques that were employed in these analyses can be applied to the design or to the safety analysis of the actual cask, which included the computer codes and the nuclear data. In future for the full scale model of the PIE cask, its shielding effect will be calculated by means of the replacement of the source spectrum from 252 Cf to a FBR fuel assembly of post-irradiation-experiment. (J.P.N.)

  1. Multiple-Angle Muon Radiography of a Dry Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Durham, J. Matthew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Guardincerri, Elena [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Morris, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poulson, Daniel Cris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bacon, Jeffrey Darnell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Morley, Deborah Jean [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Plaud-Ramos, Kenie Omar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-23

    A partially loaded dry storage cask was imaged using cosmic ray muons. Since the cask is large relative to the size of the muon tracking detectors, the instruments were placed at nine different positions around the cask to record data covering the entire fuel basket. We show that this technique can detect the removal of a single fuel assembly from the center of the cask.

  2. Thermal processes evaluation for RWMC wastes

    International Nuclear Information System (INIS)

    1991-01-01

    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

  3. Accelerating evaluation of converged lattice thermal conductivity

    Science.gov (United States)

    Qin, Guangzhao; Hu, Ming

    2018-01-01

    High-throughput computational materials design is an emerging area in materials science, which is based on the fast evaluation of physical-related properties. The lattice thermal conductivity (κ) is a key property of materials for enormous implications. However, the high-throughput evaluation of κ remains a challenge due to the large resources costs and time-consuming procedures. In this paper, we propose a concise strategy to efficiently accelerate the evaluation process of obtaining accurate and converged κ. The strategy is in the framework of phonon Boltzmann transport equation (BTE) coupled with first-principles calculations. Based on the analysis of harmonic interatomic force constants (IFCs), the large enough cutoff radius (rcutoff), a critical parameter involved in calculating the anharmonic IFCs, can be directly determined to get satisfactory results. Moreover, we find a simple way to largely ( 10 times) accelerate the computations by fast reconstructing the anharmonic IFCs in the convergence test of κ with respect to the rcutof, which finally confirms the chosen rcutoff is appropriate. Two-dimensional graphene and phosphorene along with bulk SnSe are presented to validate our approach, and the long-debate divergence problem of thermal conductivity in low-dimensional systems is studied. The quantitative strategy proposed herein can be a good candidate for fast evaluating the reliable κ and thus provides useful tool for high-throughput materials screening and design with targeted thermal transport properties.

  4. Welding issues associated with design, fabrication and loading of spent fuel storage casks

    International Nuclear Information System (INIS)

    Battige, C.K. Jr.; Howe, A.G.; Sturz, F.C.

    1999-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has observed a number of welding issues associated with design, fabrication, and loading of spent fuel storage casks. These emerging welding-related issues involving a certain dry cask storage system have challenged the safety basis for which NRC approved the casks for storage of spent nuclear fuel. During closure welding, problems have been encountered with cracking. Although the cracks have been attributed to several causes including material suitability, joint restraint and residual stresses, NRC believes hydrogen-induced cracking is the most likely explanation. In light of these cracking events and the potential for flaws in any welding process, NRC sought verification of the corrective actions and the integrity of the lid closure welds before allowing additional casks to be loaded. As a result, the affected utility companies modified the closure welding procedures and developed an acceptable ultrasonic inspection (UT) method. In addition, the casks already loaded at three power reactor sites will require additional non-destructive examinations (NDE) to determine their suitability for continued use. NRC plans to evaluate the generic implications of this issue for current designs and for those in the licensing process. (author)

  5. Releasable activity and maximum permissible leakage rate within a transport cask of Tehran Research Reactor fuel samples

    Directory of Open Access Journals (Sweden)

    Rezaeian Mahdi

    2015-01-01

    Full Text Available Containment of a transport cask during both normal and accident conditions is important to the health and safety of the public and of the operators. Based on IAEA regulations, releasable activity and maximum permissible volumetric leakage rate within the cask containing fuel samples of Tehran Research Reactor enclosed in an irradiated capsule are calculated. The contributions to the total activity from the four sources of gas, volatile, fines, and corrosion products are treated separately. These calculations are necessary to identify an appropriate leak test that must be performed on the cask and the results can be utilized as the source term for dose evaluation in the safety assessment of the cask.

  6. Safety analysis report: packages. Pu oxide and Am oxide shipping cask (Packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1980-05-01

    The PuO 2 cask or SP 5320-2 and 3 cask is designed for surface shipment of americium or plutonium. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, and Chapter I, Interstate Commerce Commission. The package has been assessed for transport of up to 357 grams of plutonium (403 grams PuO 2 powder) and up to 176 grams of americium (200 grams AmO 2 powder), having a maximum decay heat of 203 watts. Criticality evaluation alone would allow the shipment as Fissile Class II but the radiation level of the cask, measured at the time of shipment, may exceed 50 mrem/h at the surface and require shipment as Fissile Class III. Sample calculations address only the more restrictive of the two materials, which in most cases is 238 PuO 2

  7. Conceptual cask design with burnup credit

    International Nuclear Information System (INIS)

    Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong

    2003-01-01

    Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)

  8. Final version dry cask storage study

    International Nuclear Information System (INIS)

    1989-02-01

    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

  9. Initial version, dry cask storage study

    International Nuclear Information System (INIS)

    1988-08-01

    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

  10. Validation of a fracture mechanics approach to nuclear transportation cask design through a drop test program

    International Nuclear Information System (INIS)

    Sorenson, K.B.

    1986-01-01

    Sandia National Laboratories (SNL), under contract to the Department of Energy, is conducting a research program to develop and validate a fracture mechanics approach to cask design. A series of drop tests of a transportation cask is planned for the summer of 1986 as the method for benchmarking and, thereby, validating the fracture mechanics approach. This paper presents the drop test plan and background leading to the development of the test plan including structural analyses, material characterization, and non-destructive evaluation (NDE) techniques necessary for defining the test plan properly

  11. Development of integrated cask body and base plate

    International Nuclear Information System (INIS)

    Sasaki, T.; Koyama, Y.; Yoshida, T.; Wada, T.

    2015-01-01

    The average of occupancy of stored spent-fuel in the nuclear power plants have reached 70 percent and it is anticipated that the demand of metal casks for the storage and transportation of spent-fuel rise after resuming the operations. The main part of metal cask consists of main body, neutron shield and external cylinder. We have developed the manufacturing technology of Integrated Cask Body and Base Plate by integrating Cask Body and Base Plate as monolithic forging with the goal of cost reduction, manufacturing period shortening and further reliability improvement. Here, we report the manufacturing technology, code compliance and obtained properties of Integrated Cask body and Base Plate. (author)

  12. Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

    Energy Technology Data Exchange (ETDEWEB)

    Ibarra, Luis; Medina, Ricardo; Yang, Haori

    2018-03-23

    This study assessed the risk of loss of structural integrity of transportation casks and fuel cladding after extended storage. Although it is known that fuel rods discharged from NPPs have a small percentage of rod cladding defects, the behavior of fuel cladding and the structural elements of assemblies during transportation after long-term storage is not well understood. If the fuel degrades during extended storage, it could be susceptible to damage from vibration and impact loads during transport operations, releasing fission-product gases into the canister or the cask interior (NWTRB 2010). Degradation of cladding may occur due to mechanisms associated with hydrogen embrittlement, delayed hydride cracking, low temperature creep, and stress corrosion cracking (SCC) that may affect fuel cladding and canister components after extended storage of hundreds of years. Over extended periods at low temperatures, these mechanisms affect the ductility, strength, and fracture toughness of the fuel cladding, which becomes brittle. For transportation purposes, the fuel may be transferred from storage to shipping casks, or dual-purpose casks may be used for storage and transportation. Currently, most of the transportation casks will be the former case. A risk assessment evaluation is conducted based on results from experimental tests and simulations with advanced numerical models. A novel contribution of this study is the evaluation of the combined effect of component aging and vibration/impact loads in transportation scenarios. The expected levels of deterioration will be obtained from previous and current studies on the effect of aging on fuel and cask components. The emphasis of the study is placed on the structural integrity of fuel cladding and canisters.

  13. Thermal Environment evaluation in Commercial kitchens

    DEFF Research Database (Denmark)

    Simone, Angela; Olesen, Bjarne W.

    2012-01-01

    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......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...

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

    International Nuclear Information System (INIS)

    Clements, E.P.

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1997-01-01

    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

  16. 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.

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

    International Nuclear Information System (INIS)

    Kenneally, R.; Kessler, J.

    2001-01-01

    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)

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

    International Nuclear Information System (INIS)

    Kim, J. S.; Lee, K. S.; Choi, J. W.; Kwon, S.

    2010-01-01

    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

  19. 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

  20. A program to qualify ductile cast iron for use as a containment material for type B transport cask

    International Nuclear Information System (INIS)

    Golliher, K.G.; Sorenson, K.B.; Witt, C.R.

    1990-01-01

    This paper reports on the Department of Energy (DOE) investigations for the use of ductile cast iron (DCI) as a candidate material for radioactive material transportation cask construction. The investigation will include materials testing and full-scale cask testing. The major effort will focus on materials qualification and cask evaluation of the 9 meter and puncture drop test events. Interaction by contract with the private industry, the American Society for Testing and Materials (ASTM) Committee A4.04, and the Electric Power Research Institute (EPRI) will be actively pursued to establish material specification acceptance criteria for ductile iron use as a cask material in the United States of America (USA). All test results will be documented in the safety analysis report for packaging for submission to the U.S. Nuclear Regulatory Commission (NRC). The goal of this program is a certificate of compliance for DCI from the NRC to transport high-level radioactive materials. The acceptance of DCI within the USA cask design community will offer an alternative to present-day materials for cask construction, and its entry has the potential of providing significant cost-savings

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

    International Nuclear Information System (INIS)

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

    1995-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-01

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

  3. Developing cask designs in the USSR

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    To tackle the problem of transporting spent fuel from its VVER-1000s, the Soviet Union has developed two casks - the TK-10 and the TK-13 which are described here. Future developments of these designs may use a silicon-organics based material for the solid neutron shielding and no neutron absorbers in the fuel assembly basket. (author)

  4. Structural dimensioning of dual purpose cask prototype

    International Nuclear Information System (INIS)

    Silva, Luiz Leite da; Mourao, Rogerio Pimenta; Lopes, Claudio Cunha

    2005-01-01

    The structural dimensioning of a Type B(U) dual purpose cask prototype is part of the scope of work of the Brazilian institute CDTN in the IAEA regional project involving Latin American countries which operate research reactors (Argentina, Brazil, Chile, Mexico and Peru). In order to meet the dimensional and operational characteristics of the reactor facilities in these countries, a maximum weight of 10.000 kgf and a maximum dimension of 1 m in at least one direction were set for the cask. With these design restrictions, the cask's payload is either 21 MTR or 78 TRIGA fuel elements. The cask's most important components are main body, primary and secondary lids, basket and impact limiters. The main body has a sandwich-like wall with internal and external layers made of AISI 304 stainless steel with lead in-between. The lead provides biological shielding. The primary lid is similarly layered, but in the axial direction. It is provided with a double system of metallic rings and has ports for pressurization, sampling and containment verification. The secondary lid has the main function of protecting the primary lid against mechanical impacts. The basket structure is basically a tube array reinforced by bottom plate, feet and spacers. Square tubes are used for MTR elements and circular tubes for TRIGA elements. Finally, the impact limiters are structures made of an external stainless steel thin covering and a filling made of the wood composite OSB - Oriented Strand Board. The prototype is provided with bottom and top impact limiters, which are attached to each other by means of four threaded rods. The limiters are not rigidly attached to the cask body. A half scale cask model was designed to be submitted to a testing program. As its volume scales down to 1:8, the model weight is 1,250 kgf. This paper presents the methodology for the preliminary structural dimensioning of the critical parameters of the cask prototype. Both normal conditions of operation and hypothetical

  5. 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 conversion...

  6. Method to mount defect fuel elements i transport casks

    International Nuclear Information System (INIS)

    Borgers, H.; Deleryd, R.

    1996-01-01

    Leaching or otherwise failed fuel elements are mounted in special containers that fit into specially designed chambers in a transportation cask for transport to reprocessing or long-time storage. The fuel elements are entered into the container under water in a pool. The interior of the container is dried before transfer to the cask. Before closing the cask, its interior, and the exterior of the container are dried. 2 figs

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hartman, D.J.; Miller, D.D. [Bechtel National, Inc., San Francisco, CA (United States); Hill, R.R. [Sandia National Labs., Albuquerque, NM (United States)

    1992-04-01

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

  10. Transportation cask decontamination and maintenance at the potential Yucca Mountain repository

    International Nuclear Information System (INIS)

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

    1992-04-01

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

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

    International Nuclear Information System (INIS)

    Koploy, M.; Schlafer, W.; Zimmer, A.

    1987-01-01

    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

  12. Interfacing the existing cask fleet with the MRS

    International Nuclear Information System (INIS)

    Doman, J.W.; Hahn, R.E.

    1992-01-01

    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

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

    International Nuclear Information System (INIS)

    McCreery, P.N.

    1980-10-01

    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

  14. 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)

  15. 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 by...

  16. Development of heat resistant concrete and its application to concrete casks. Improvement of neutron shielding performance of concrete in high temperature environment

    International Nuclear Information System (INIS)

    Owaki, Eiji; Hata, Akihito; Sugihara, Yutaka; Shimojo, Jun; Taniuchi, Hiroaki; Mantani, Kenichi

    2003-01-01

    Heat resistant concrete with hydrogen, which is able to shield neutron at more than 100degC, was developed. Using this new type concrete, a safety concrete cask having the same concept of metal casks was designed and produced. The new type cask omitted the inhalation and exhaust vent of the conventional type concrete casks. The new concrete consists of Portland cement added calcium hydroxide, iron powder and iron fiber. It showed 2.17 g/cm 3 density, 10.8 mass% water content, 1.4 W/(m·K) thermal conductivity at 150degC. Increasing of heat resistance made possible to produce the perfect sealing type structure, which had high shielding performance of radiation no consideration for streaming of radiation. Moreover, a monitor of sealing can be set. General view of concrete casks, outer view of 1/3 scaled model, cask storage system in the world, properties of new developed heat resistant concrete, results of shielding calculation are contained. (S.Y.)

  17. MCO loading and cask loadout technical manual

    International Nuclear Information System (INIS)

    PRAGA, A.N.

    1998-01-01

    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

  18. Trunnions for spent fuel element shipping casks

    International Nuclear Information System (INIS)

    Cooke, B.

    1989-01-01

    Trunnions are used on spent fuel element shipping casks for one or more of a combination of lifting, tilting or securing to a transport vehicle. Within the nuclear transportation industry there are many different philosophies on trunnions, concerning the shape, manufacture, attachment, inspection, maintenance and repair. With the volume of international transport of spent fuel now taking place, it is recognized that problems are occurring with casks in international traffic due to the variance of the philosophies, national standards, and the lack of an international standard. It was agreed through the ISO that an international standard was required to harmonize. It was not possible to evolve an international standard. It was only possible to evolve an international guide. To evolve a standard would mean superseding any existing national standards which already cover particular aspects of trunnions i.e. deceleration forces imposed on trunnions used as tie down features. Therefore the document is a guide only and allows existing national standards to take precedence where they exist. The guide covers design, manufacture, maintenance, repair and quality assurance. The guide covers trunnions used on spent fuel casks transported by road, rail and sea. The guide details the considerations which should be taken account of by cask designers, i.e. stress intensity, design features, inspection and test methods etc. Manufacture, attachment and pre-service testing is also covered. The guide details user requirements which should also be taken account of, i.e. servicing frequency, content, maintenance and repair. The application of quality assurance is described separately although the principles are used throughout the guide

  19. 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.

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

    International Nuclear Information System (INIS)

    Mat sumura, T.; Sasahara, A.; Takei, M.; Takekawa, T.; Kagehira, K.; Nicolaou, G.; Betti, M.

    1998-01-01

    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)

  1. Long-term sealability of spent fuel casks

    International Nuclear Information System (INIS)

    Kato, O.; Kosaki, A.; Ito, C.; Ozaki, S.; Iitsuka, S.

    1993-01-01

    This report presents the results of long-term sealability test of full-scale model casks which have been in progress since 1990. (O.Kato and C.Ito 1992) The purpose is to confirm the long-term sealability of the casks. (J.P.N.)

  2. Shipment and Storage Containers for Tritium Production Transportation Casks

    International Nuclear Information System (INIS)

    Massey, W.M.

    1998-04-01

    The need for a shipping and storage container for the Tritium production transportation casks is addressed in this report. It is concluded that a shipping and storage container is not required. A recommendation is made to eliminate the requirement for this container because structural support and inerting requirements can be satisfied completely by the cask with a removable basket

  3. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    .... Background Section 218(a) of the Nuclear Waste Policy Act (NWPA) of 1982, as amended, requires that ``the... condition, the design of the cask would prevent loss of containment, shielding, and criticality control. If there is no loss of containment, shielding, or criticality control, the environmental impacts would be...

  4. Operation and maintenance of the T-3 cask system

    International Nuclear Information System (INIS)

    Hussey, M.W.; Berger, J.D.; Peterson, J.M.

    1983-01-01

    The T-3 cask system consists of three lead-shielded casks and the associated payload containers, internal fixturing, tiedowns, transportation trailers and handling devices. The three casks were designed to meet the requirements of Title 10 of the Code of Federal Regulations, Part 71. The Nuclear Regulatory Commission cask licensing activities for original design and for licensing revisions have required significant analytical support. Commercial transportation contractors can provide needed services including provisions of suitable equipment, compliances with security requirements, and safe movement of the shipment at a potential savings over DOE-owned transportation systems. Proper periodic inspection/maintenance activities supported by adequate decontamination facilities are a must in keeping the T-3 casks available for service

  5. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Proposed... spent fuel storage cask regulations by revising the Holtec International HI-STORM 100 dry cask storage... Amendment No. 8 to CoC No. 1014 and does not include other aspects of the HI-STORM 100 dry storage cask...

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

    International Nuclear Information System (INIS)

    Rathbun, R.

    1994-01-01

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

  7. Development of design and safety analysis supporting system for casks

    International Nuclear Information System (INIS)

    Ohsono, Katsunari; Higashino, Akira; Endoh, Shuji

    1993-01-01

    Mitsubishi heavy Industries has developed a design and safety analysis supporting system 'CADDIE' (Cask Computer Aided Design, Drawing and Integrated Evaluation System), with the following objectives: (1) Enhancement of efficiency of the design and safety analysis (2) Further advancement of design quality (3) Response to the diversification of design requirements. The features of this system are as follows: (1) The analysis model data common to analyses is established, and it is prepared automatically from the model made by CAD. (2) The input data for the analysis code is available by simple operation of conversation type from the analysis model data. (3) The analysis results are drawn out in diagrams by output generator, so as to facilitate easy observation. (4) The data of material properties, fuel assembly data, etc. required for the analyses are made available as a data base. (J.P.N.)

  8. Mechanical properties used for the qualification of transport casks

    International Nuclear Information System (INIS)

    Salzbrenner, R.; Crenshaw, T.B.; Sorenson, K.B.

    1993-01-01

    The qualification process that should be sufficient for qualification of a specific cask (material/geometry combination) has been examined. The prototype cask should be tested to determine its overall variation in microstructure, chemistry, and mechanical properties. This prototype may also be subjected to 'proof testing' to demonstrate the validity of the design analysis (including the mechanical properties used in the analysis). The complete mechanical property mapping does not necessarily have to precede the proof testing (i.e., portions of the cask which experience only low (elastic) loads during the drop test are suitable for mechanical test specimens). The behavior of the prototype cask and the production casks are linked by assuring that each cask possesses at least the minimum level of one or more critical mechanical properties. This may be done by measuring the properties of interest directly, or by relying on a secondary measurement (such as subsize mechanical test results or microstructure/compositional measurements) which has been statistically correlated to the critical properties. The database required to show the correlation between the secondary measurement and the valid design property may be established by tests on the material from the prototype cask. The production controls must be demonstrated as being adequate to assure that a uniform product is produced. The testing of coring (or test block or prolongation) samples can only be viewed as providing a valid link to the benchmark results provided by the prototype cask if the process used to create follow-on casks remains essentially similar. The MOSAIK Test Program has demonstrated the qualification method through the benchmarking stage. The program did not establish for qualifying serial production casks through, for example, a correlation between small specimen parameters and valid design fracture toughness properties. Such a correlation would require additional experimental work. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Howard, Rob; Van den Akker, Bret

    2014-01-01

    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)

  10. Use of inelastic analysis in cask design

    International Nuclear Information System (INIS)

    Ammerman, Douglas J.; Breivik, Nicole L.

    2000-01-01

    In this paper, the advantages and disadvantages of inelastic analysis are discussed. Example calculations and designs showing the implications and significance of factors affecting inelastic analysis are given. From the results described in this paper it can be seen that inelastic analysis provides an improved method for the design of casks. It can also be seen that additional code and standards work is needed to give designers guidance in the use of inelastic analysis. Development of these codes and standards is an area where there is a definite need for additional work. The authors hope that this paper will help to define the areas where that need is most acute

  11. 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)

    Wille, Frank; Droste, Bernhard; Zencker, Uwe; Komann, Steffen

    2010-01-01

    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.

  12. Test facilities for evaluating nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  13. 76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ... Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear Regulatory Commission. ACTION: Direct final... regulations to add the HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks... cask designs. Discussion This rule will add the Holtec HI-STORM Flood/Wind (FW) cask system to the list...

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

    International Nuclear Information System (INIS)

    Thunborg, S.

    1987-01-01

    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

  15. 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 cask's 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.

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

    International Nuclear Information System (INIS)

    1990-06-01

    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

  17. Thermal morphing anisogrid smart space structures: thermal isolation design and linearity evaluation

    Science.gov (United States)

    Phoenix, Austin A.

    2017-04-01

    To meet the requirements for the next generation of space missions, a paradigm shift is required from current structures that are static, heavy and stiff, toward innovative structures that are adaptive, lightweight, versatile, and intelligent. A novel morphing structure, the thermally actuated anisogrid morphing boom, can be used to meet the design requirements by making the primary structure actively adapt to the on-orbit environment. The anisogrid structure is able to achieve high precision morphing control through the intelligent application of thermal gradients. This active primary structure improves structural and thermal stability performance, reduces mass, and enables new mission architectures. This effort attempts to address limits to the author's previous work by incorporating the impact of thermal coupling that was initially neglected. This paper introduces a thermally isolated version of the thermal morphing anisogrid structure in order to address the thermal losses between active members. To evaluate the isolation design the stiffness and thermal conductivity of these isolating interfaces need to be addressed. This paper investigates the performance of the thermal morphing system under a variety of structural and thermal isolation interface properties.

  18. Maximizing allowable cask payloads using zone-loading and cooling table specifications

    International Nuclear Information System (INIS)

    Hopf, J.E.; Lloyd, T.

    2004-01-01

    The newer dual-purpose canister designs generally have a higher fuel assembly capacity than earlier designs. Due to the resulting increases in thermal and radiological source terms from the assembly payload, this will generally result in higher cask system temperatures and cask external dose rates, making it more difficult to meet 10CFR71 and 10CFR72 thermal and radiological requirements. One approach to addressing this issue would be to employ advanced, and potentially expensive, engineering features to enhance cask shielding and heat removal capabilities. Another approach involves the strategic loading of fuel assemblies in specific locations within the dual-purpose canister, along with a more rigorous analysis of the specific assembly payload configuration inside the canister. This second approach, which does not involve difficult engineering design and fabrication, and which does not add to the cost of the canister or cask, is the subject of this paper. Traditional cask licensing analyses simply model a uniform assembly payload over the entire canister interior. One, or perhaps a few ''design-basis'' combinations of burnup, enrichment, and cooling time are analyzed and qualified. All loaded assemblies must be completely bounded by one or more of the analyzed sets of design basis assembly parameters. Effectively, the ''hottest'' possible assembly is modeled in all loading slots. This paper discusses two techniques that could greatly increase the number of spent fuel pool assemblies that qualify for storage or transportation, especially when taken together. The first technique, referred to as ''zone loading'' involves loading relatively ''cold'' assemblies in the locations around the edge of the canister. The outer assemblies will almost entirely shield the neutron and gamma fluxes from the interior assemblies, reducing their contribution to cask external dose rate to very low levels. This allows much ''hotter'' possible assembly is modeled in all loading slots

  19. 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.

  20. 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.

  1. Nuclear cask testing films misleading and misused

    International Nuclear Information System (INIS)

    Audin, L.

    1991-10-01

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

  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. What drives Greek consumer preferences for cask wine?

    DEFF Research Database (Denmark)

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

    2012-01-01

    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...... 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....

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    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

  5. 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

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

    International Nuclear Information System (INIS)

    Shirai, Koji; Ito, Chihiro; Ryu, Hiroshi

    1993-01-01

    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

  7. Method of reducing contamination by radioactivity on a cask for spent fuel

    International Nuclear Information System (INIS)

    Irie, Hiromitsu.

    1976-01-01

    Object: To reduce a radioactivity level of an outer surface of a cask. Structure: After a cask body with a contamination preventive cover mounted thereon has been immersed into a water tank by means of a hanging member, a cask with the contamination preventive cover mounted thereon is suspended upwardly of the cask body. A spent fuel is stored within the cask body by the hanging member, and thereafter the cask lid is attached to the body and closed, and the hanging member is driven to move the cask to a decontamination station. Next, the contamination preventive cover is removed and the cask body and cask lid are closed and connected, and in this condition, decomtamination water is sprinkled on the outer surface to effect decontamination until the radioactivity reaches a level below rated value. (Yoshino, Y.)

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

    International Nuclear Information System (INIS)

    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

  9. 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.

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

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-09-01

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

  11. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    Parks, C.V.

    1989-01-01

    The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of k eff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

  12. Developing new transportable storage casks for interim dry storage

    International Nuclear Information System (INIS)

    Hayashi, K.; Iwasa, K.; Araki, K.; Asano, R.

    2004-01-01

    Transportable storage metal casks are to be consistently used during transport and storage for AFR interim dry storage facilities planning in Japan. The casks are required to comply with the technical standards of regulations for both transport (hereinafter called ''transport regulation'') and storage (hereafter called ''storage regulation'') to maintain safety functions (heat transfer, containment, shielding and sub-critical control). In addition to these requirements, it is not planned in normal state to change the seal materials during storage at the storage facility, therefore it is requested to use same seal materials when the casks are transported after storage period. The dry transportable storage metal casks that satisfy the requirements have been developed to meet the needs of the dry storage facilities. The basic policy of this development is to utilize proven technology achieved from our design and fabrication experience, to carry out necessary verification for new designs and to realize a safe and rational design with higher capacity and efficient fabrication

  13. The role of ORIGEN-S in the design of burnup credit spent fuel casks

    International Nuclear Information System (INIS)

    Brady, M.C.

    1991-01-01

    Current licensing practices for spent fuel pools, storage facilities, and transportation casks require a conservative ''fresh fuel assumption'' be used in the criticality analysis. Burnup credit refers to a new approach in criticality analyses for spent fuel handling systems in which reactivity credit is allowed for the depleted state of the fuel. Studies have shown that the increased cask capacities that can be achieved with burnup credit offer both economic and risk incentives. The US Department of Energy is currently sponsoring a program to develop analysis methodologies and establish a new generation of spent fuel casks using the principle of burnup credit. The key difference in this new approach is the necessity to accurately predict the isotopic composition of the spent fuel. ORIGEN-S was selected to satisfy this requirement because of the flexibility and user-friendly input offered via its usage in the Standardized Computer Analyses for Licensing and Evaluation (SCALE) code system. Specifically, through the Shielding Analysis Sequence 2H (SAS2H), ORIGEN-S is linked with cross-section processing codes and one-dimensional transport analyses to produce problem-specific cross-section data for the point-depletion calculation. The utility code COUPLE facilitates updating basic cross-section and fission-yield data for the calculations. This paper describes the fundamental role fulfilled by ORIGEN-S in the development of the analysis methodology, validation of the methods, definition of criticality safety margins and other licensing considerations in the design of a new generation of spent fuel casks. Particular emphasis is given to the performance of ORIGEN-S in comparisons with measurements of irradiated fuel compositions and in predicting isotopics for use in the calculation of reactor restart critical configurations that are performed as a part of the validation process

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

    International Nuclear Information System (INIS)

    1988-01-01

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

  15. The development of a transportable storage cask

    International Nuclear Information System (INIS)

    Stuart, I.F.

    1991-01-01

    There are a number of different technologies for implementing interim storage of spent fuel at reactor sites. It is generally accepted that, if possible, expanding the capacity of existing fuel pools through the installation of compact racks and the use of fuel rod consolidation are the most economical first steps. Once these have been carried out, other alternatives must be employed if further capacity expansion is required. It is not the purpose of this paper to discuss the relative economics of these alternatives, since under specific constraints and conditions each one can be shown to have an economic benefit. However, it is the reduction in plant operations, the minimising of radiation exposure, the inherent flexibility and corresponding overall favourable economics that have led to the development of the dual purpose storage and transport cask in the past few years. (author)

  16. TRANSPORTATION CASK RECEIPT AND RETURN FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

    Arakali, V.

    2005-01-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

  18. 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)

    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

  19. thermal power stations' reliability evaluation in a hydrothermal system

    African Journals Online (AJOL)

    Dr Obe

    A quantitative tool for the evaluation of thermal power stations reliability in a hydrothermal system is presented. A reliable power station is one which would supply the required power within its installed capacity at any time within the specified voltage and frequency limits. Required for this evaluation are the station's installed ...

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

    International Nuclear Information System (INIS)

    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

  1. Theoretical consideration and evaluation of thermal diffusivity ...

    African Journals Online (AJOL)

    The aim of this study was to theoretically consider and evaluate the transient temperature-time response of resin/ionomer materials between ~2EC (melting ice temperature) and ~50EC, and indicate their suitability as lining materials. Specimens of materials, cylindrical in geometry, (6 mm diameter and 10 mm in length), ...

  2. 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)

  3. Objective and subjective thermal comfort evaluation in Hungary

    Directory of Open Access Journals (Sweden)

    Kajtar Laszlo

    2017-01-01

    Full Text Available Thermal comfort sensation can be predicted in the most exact way based on Fanger’s predicted mean vote (PMV model. This evaluation method takes all the six influencing factors into consideration: air temperature and humidity, air velocity, mean radiant temperature of surrounding surfaces, clothing insulation, and occupants’ activities. Fanger’s PMV method was developed for temperate climate and European people, with the participation of university students as subjects. Many researchers had investigated its validity in different geographic locations (i. e. climatic conditions, people and under non-laboratory circumstances. The results were summarised by van Hoof which had been published in the scientific references. The articles gave us the idea to elaborate the former measurement results. During the last decades thermal comfort was evaluated by our research team using subjective scientific questionnaires and applying the objective Fanger’s model in several office buildings in Hungary. The relation between the PMV and actual mean vote values were analysed based on these results. Investigations were carried out under steady-state conditions in winter time. We performed objective thermal comfort evaluations based on instrumental measurements using the PMV theory. Parallel to this we assessed the subjective thermal sensation using scientific questionnaires. The mathematical relationship between the actual mean vote and PMV was defined according to the evaluated thermal environment: AMV = PMV + 0.275, (arg. –1.7 ≤ PMV ≤ +0.5.

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

    Science.gov (United States)

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

    2015-05-01

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

  5. Evaluation of algorithms for geological thermal-inertia mapping

    Science.gov (United States)

    Miller, S. H.; Watson, K.

    1977-01-01

    The errors incurred in producing a thermal inertia map are of three general types: measurement, analysis, and model simplification. To emphasize the geophysical relevance of these errors, they were expressed in terms of uncertainty in thermal inertia and compared with the thermal inertia values of geologic materials. Thus the applications and practical limitations of the technique were illustrated. All errors were calculated using the parameter values appropriate to a site at the Raft River, Id. Although these error values serve to illustrate the magnitudes that can be expected from the three general types of errors, extrapolation to other sites should be done using parameter values particular to the area. Three surface temperature algorithms were evaluated: linear Fourier series, finite difference, and Laplace transform. In terms of resulting errors in thermal inertia, the Laplace transform method is the most accurate (260 TIU), the forward finite difference method is intermediate (300 TIU), and the linear Fourier series method the least accurate (460 TIU).

  6. The evaluation of the overall thermal comfort inside a vehicle

    Science.gov (United States)

    Neacsu, Catalin; Tabacu, Ion; Ivanescu, Mariana; Vieru, Ionel

    2017-10-01

    The thermal comfort is one of the most important aspects of the modern vehicles that can influence the safety, the fuel consumption and the pollutions regulation. The objective of this paper is to compare the global and absolute thermal comfort indexes for two vehicles with different distribution air systems inside the car cockpit, one using only front air vents, and the other using both front and rear air vents. The methodology of calculus consists in using the 3D model of the interior vehicle, generally in a CAD format. Then, using a meshing software to create the finite element model of the interior surfaces inside the cockpit and the volume of internal air. Using the obtained finite element geometry, there will be conducted a Theseus FE calculus using the given boundary conditions. The results of the numerical simulation are presented in terms of graphs and figures and also PMV, PPD and DTS thermal comfort indexes. With the obtained results, we will then create the graphs that allows us to evaluate the global and absolute thermal comfort indexes. The results of the evaluation show us that the use of the method allow us to evaluate with a greater accuracy the thermal comfort for the whole vehicle, not only for each passenger, like the standard methods. This shows us that in terms of general and absolute thermal comfort, the vehicle that use front and rear systems is better than the version that use only a front system. The thermal comfort is an important aspect to be taken into account from the beginning of the design stage of a vehicle, by choosing the right air conditioning system. In addition, by using the numerical simulation, we are able to reduce the time needed for preliminary tests and be able to provide the vehicle to the market earlier, at a lower development cost.

  7. Evaluation of Polyesterimide Nanocomposites Using Methods of Thermal Analysis

    Science.gov (United States)

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

    2016-02-01

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

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

    International Nuclear Information System (INIS)

    Klann, R.T.; Picker, B.A. Jr.

    1993-01-01

    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

  9. Evaluation of thermal radiation simulator rectangular pulse characterization methods

    International Nuclear Information System (INIS)

    Loucks, R.B.

    1991-01-01

    This paper discusses the thermal output of an aluminum powder/liquid oxygen Thermal Radiation Simulator (TRS) which is approximated to that of a rectangular pulse. The output varies as a function of time. The rise and fall times are not relatively abrupt. The problem is how to quantify the thermal output of the TRS into terms of rectangular pulse. Within the nuclear weapons effects community, flux, or the transient intensity of thermal radiation energy onto a surface, and fluence, the total energy irradiated on a surface over a given time, are the determining parameters for specifying or evaluating an article's survivability in the thermal environment. Four methods are used to determine the TRS output for these parameters, assuming the output to be a perfect rectangular pulse. It was essential to determine which of the four methods best quantified the thermal output average flux and fluence. The four methods were compared by a computational experiment run on a personal computer. The experiment was a simulation of five actual TRS traces irradiated onto a fictitious aluminum plate

  10. ALTERNATIVE METHOD FOR ON SITE EVALUATION OF THERMAL TRANSMITTANCE

    Directory of Open Access Journals (Sweden)

    Aleksandar Janković

    2017-08-01

    Full Text Available Thermal transmittance or U-value is an indicator of the building envelope thermal properties and a key parameter for evaluation of heat losses through the building elements due to heat transmission. It can be determined by calculation based on thermal characteristics of the building element layers. However, this value does not take into account the effects of irregularities and degradation of certain elements of the envelope caused by aging, which may lead to errors in calculation of the heat losses. An effective and simple method for determination of thermal transmittance is in situ measurement, which is governed by the ISO 9869-1:2014 that defines heat flow meter method. This relatively expensive method leaves marks and damages surface of the building element. Furthermore, the final result is not always reliable, in particular when the building element is light or when the weather conditions are not suitable. In order to avoid the above mentioned problems and to estimate the real thermal transmittance value an alternative experimental method, here referred as the natural convection and radiation method, is proposed in this paper. For determination of thermal transmittance, this method requires only temperatures of inside and outside air, as well as the inner wall surface temperature. A detailed statistical analysis, performed by the software package SPSS ver. 20, shows several more advantages of this method comparing to the standard heat flow meter one, besides economic and non-destructive benefits.

  11. Development of transport and storage cask for high burn-up spent fuel

    International Nuclear Information System (INIS)

    Kuri, S.; Tamaki, H.; Hode, S.

    2004-01-01

    Mitsubishi Heavy Industries, LTD. (MHI) has been developing various transport and storage casks (MSF cask fleet) for high burn-up spent nuclear fuel (SNF). This paper outlines the specifications and describes the features of the newly developed casks and the advanced technology that enables the maximize number of the accommodated fuel assemblies of high burn-up and short cooling period

  12. Development of NUPAC 140B 100 ton rail/barge cask

    International Nuclear Information System (INIS)

    1990-04-01

    The 140-B Cask Ancillary Equipment includes all cask-related hardware necessary for a complete transportation package and for handling of the cask at shipping and receiving facilities. The transportation package equipment includes the cask tiedown system, the railcar and the sunshield/personnel barrier. The cask handling systems include both single and dual load path cask lifting fixtures, a cask uprighting system, an intermodal transfer system, and the cask drain and fill system. This document describes the individual systems in terms of their purpose, their function, and their mechanical features. Structural analyses are provided for the cask lifting and tiedown devices. The cask ancillary equipment will also include special tools and equipment such as seal surface protection device, special torque wrenches, leak test equipment, etc., for handling the cask at a reactor site. Although final design work remains to be completed, the ancillary equipment design information presented in this document ensures that the 140-B cask transportation package will meet or exceed all structural, functional, and operational requirements, within the specified gross vehicle weight limit. 18 figs

  13. Potential Multi-Canister Overpack (MCO) Cask Drop in the K West Basin South Loadout Pit

    International Nuclear Information System (INIS)

    POWERS, T.B.

    1999-01-01

    This calculation note documents the probabilistic calculation of a potential drop of a multi-canister overpack (MCO) cask or MCO cask and immersion pail at the K West Basin south loadout pit. The calculations are in support of the cask loading system (CLS) subproject alignment of CLS equipment in the K West Basin south loadout pit

  14. thermal power stations' reliability evaluation in a hydrothermal system

    African Journals Online (AJOL)

    Dr Obe

    R. C. OKONKWO. Department of Electrical Engineering. UNIVERSITY OF NIGERIA. NSUKKA. ABSTRACT. A quantitative tool for the evaluation of thermal power stations reliability in a hydrothermal system is presented. A reliable power station is one which would supply the required power within its installed capacity at any ...

  15. Application of liquid crystals in thermal nondestructive evaluation

    International Nuclear Information System (INIS)

    Panakal, J.P.; Mukherjee, S.; Ghosh, J.K.

    1983-01-01

    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)

  16. High energy neutron transmission analysis of dry cask storage

    Science.gov (United States)

    Greulich, Christopher; Hughes, Christopher; Gao, Yuan; Enqvist, Andreas; Baciak, James

    2017-12-01

    Since the U.S. currently only approves of storing used nuclear fuel in pools or dry casks, the demand for dry cask storage is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. With the rising demand comes additional pressure to ensure the integrity of dry cask systems. Visual inspection is costly and man-power intensive, so alternative nondestructive testing techniques are desired to insure the continued safe and effective storage of fuel. One such approach being investigated by the University of Florida is neutron based computed tomography. Simulations in MCNP are preformed where D-T energy neutrons are transmitted through the dry cask and measured on the opposite side. If the transmitted signal is clear enough, the interior of the cask can be reconstructed from the measurement of the alterations of neutron signal intensity using standard mathematical techniques developed for medical imaging. Preliminary efforts show a correlation between energy and number of scatters (which is an indication of retention of position information). Work is ongoing to quantify if the correlation is strong enough that an energy discriminator may be used as a filter in future image reconstruction. The calculated transmission probability suggests that an image could be reconstructed with a week of scanning.

  17. Shipping and storage cask data for spent nuclear fuel

    International Nuclear Information System (INIS)

    Johnson, E.R.; Notz, K.J.

    1988-11-01

    This document is a compilation of data on casks used for the storage and/or transport of commercially generated spent fuel in the US based on publicly available information. In using the information contained in the following data sheets, it should be understood that the data have been assembled from published information, which in some instances was not internally consistent. Moreover, it was sometimes necessary to calculate or infer the values of some attributes from available information. Nor was there always a uniform method of reporting the values of some attributes; for example, an outside surface dose of the loaded cask was sometimes reported to be the maximum acceptable by NRC, while in other cases the maximum actual dose rate expected was reported, and in still other cases the expected average dose rate was reported. A summary comparison of the principal attributes of storage and transportable storage casks is provided and a similar comparison for shipping casks is also shown. References to source data are provided on the individual data sheets for each cask

  18. Shipping and storage cask data for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, E.R.; Notz, K.J.

    1988-11-01

    This document is a compilation of data on casks used for the storage and/or transport of commercially generated spent fuel in the US based on publicly available information. In using the information contained in the following data sheets, it should be understood that the data have been assembled from published information, which in some instances was not internally consistent. Moreover, it was sometimes necessary to calculate or infer the values of some attributes from available information. Nor was there always a uniform method of reporting the values of some attributes; for example, an outside surface dose of the loaded cask was sometimes reported to be the maximum acceptable by NRC, while in other cases the maximum actual dose rate expected was reported, and in still other cases the expected average dose rate was reported. A summary comparison of the principal attributes of storage and transportable storage casks is provided and a similar comparison for shipping casks is also shown. References to source data are provided on the individual data sheets for each cask.

  19. Dry Cask Storage Inspection and Monitoring. Interim Report.

    Energy Technology Data Exchange (ETDEWEB)

    Bakhtiari, Susan [Argonne National Lab. (ANL), Argonne, IL (United States); Elmer, Thomas W. [Argonne National Lab. (ANL), Argonne, IL (United States); Koehl, Eugene R. [Argonne National Lab. (ANL), Argonne, IL (United States); Wang, Ke [Argonne National Lab. (ANL), Argonne, IL (United States); Raptis, Apostolos C. [Argonne National Lab. (ANL), Argonne, IL (United States); Kunerth, Dennis C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Birk, Sandra M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-03-04

    Recently, the U.S. Nuclear Regulatory Commission (NRC) issued the guidance on the aging management of dry storage facilities that indicates the necessity to monitor the conditions of dry cask storage systems (DCSSs) over extended periods of time.1 Part of the justification of the aging management plans is the requirement for inspection and monitoring to verify whether continued monitoring, inspection or mitigation are necessary. To meet this challenge Argonne National Laboratory (ANL) in collaboration with Idaho National Laboratory (INL) is conducting scoping studies on current and emerging nondestructive evaluation/examination (NDE) and online monitoring (OLM) technologies for DCSS integrity assessments. The scope of work plan includes identification and verification of technologies for long-term online monitoring of DCSSs’ crucial physical parameters such as temperature, pressure, leakage and structural integrity in general. Modifications have been made to the current technologies to accommodate field inspections and monitoring. A summary of the scoping studies and experimental efforts conducted to date as well as plans for future activities is provided below.

  20. 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.

  1. Vestibule and Cask Preparation Mechanical Handling Calculation

    International Nuclear Information System (INIS)

    Ambre, N.

    2004-01-01

    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-AC--28-01R 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-AC--28-01R 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

  2. Nuclear thermal rocket nozzle testing and evaluation program

    Science.gov (United States)

    Davidian, Kenneth O.; Kacynski, Kenneth J.

    1993-01-01

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

  3. Phase 1 study of metallic cask systems for spent fuel management from reactor to repository. Volume I. Phase 1 study summary

    Energy Technology Data Exchange (ETDEWEB)

    None

    1986-02-01

    It was proposed to perform a systems evaluation of metallic cask systems in order to define and examine the use of various metallic cask concepts or combination of concepts for the overall inventory management of spent fuel starting with its discharge from reactors to its emplacement in geologic repositories. This systems evaluation occurs in three phases. This three phase systems evaluation leads to a definition and recommendation of a sound and practical metallic cask system to accomplish efficient and effective management of spent fuel in the back end of the nuclear fuel cycle. Phase 1 Study objectives: establish system-wide functional criteria and assumptions; perform the systems engineering needed to define the metallic cask concepts and their feasibility; perform a screening evaluation of the technical and economic merits of the concepts; and recommend those to be included for a more detailed systems evaluation in Phase 2. Phase 2 Study objectives: refine the system-wide functional criteria and assumptions; perform the design engineering needed to enhance the validity and workability of those concepts recommended in Phase 1; and perform a more detailed systems evaluation. Phase 3 Study objectives: conclude the systems evaluation and develop an implementation plan. Volume I presents an overview of the detailed systems evaluation presented in Volume II.

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

    Science.gov (United States)

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

    2010-11-01

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

  5. Thermal management evaluation of the complex electro-optical system

    Directory of Open Access Journals (Sweden)

    Nijemčević Srećko S.

    2017-01-01

    Full Text Available The thermal management of a complex electro-optical system aimed for outdoor application is challenging task due to the requirement of having an air-sealed enclosure, harsh working environment, and an additional thermal load generated by sunlight. It is essential to consider the effect of heating loads in the system components, as well as the internal temperature distribution, that can have influence on the system life expectancy, operational readiness and parameters, and possibility for catastrophic failure. The main objective of this paper is to analyze internal temperature distribution and evaluate its influence on system component operation capability. The electro-optical system simplified model was defined and related thermal balance simulation model based on Solid Works thermal analysis module was set and applied for temperature distribution calculation. Various outdoor environment scenarios were compared to evaluate system temperature distribution and evaluate its influence on system operation, reliability, and life time in application environment. This work was done during the design process as a part of the electro-optical system optimization. The results show that temperature distribution will not be cause for catastrophic failure and malfunction operation during operation in the expected environment.

  6. Study of casks shielded with heavy metal to transport highly radioactive substances; Estudo de embalados com blindagem em metal pesado para transporte de substancias altamente radioativas

    Energy Technology Data Exchange (ETDEWEB)

    Lucchesi, R.F.; Hara, D.H.S.; Martinez, L.G.; Mucsi, C.S.; Rossi, J.L., E-mail: rflguimaraes@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2014-07-01

    Nowadays, Brazil relies on casks produced abroad for transportation in its territory of substances that are sources of high radioactivity, especially the Mo-99. The product of the radioactive decay of the Mo-99 is the Tc-99m, which is used in nuclear medicine for administration to humans in the form of injectable radioactive drugs for the image diagnosis of numerous pathologies. This paper aims to study the existing casks in order to propose materials for the construction of the core part as shielding against gamma radiation. To this purpose, the existing literature on the subject was studied, as well as evaluation of existing and available casks. The study was focused on the core of which is made of heavy metals, especially depleted uranium for shielding the emitted radiation. (author)

  7. 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.

  8. Cask dismantlement at the INEL for lead recovery and recycle

    International Nuclear Information System (INIS)

    Feldman, E.M.; Thurmond, S.M.

    1996-01-01

    The Idaho National Engineering Laboratory (INEL) is successfully recycling radiologically clean metallic scrap lead and steel separated from obsolete lead shielded casks and devices. These casks and devices have low level radiological surface contamination and, because of the characteristic toxicity of the lead shielding, were previously declared a mixed hazardous waste and placed in Resource Conservation Recovery Act (RCRA) storage. This paper highlights the unique capabilities available at the INEL to perform dismantlement and recycling of casks and devices weighing from a few pounds to over 40 tons. Issues and results associated with the INEL's recycling of lead as free-released scrap metal are presented and discussed. Dismantlement process methods and equipment are described

  9. On-site concrete cask storage system for spent nuclear fuel

    International Nuclear Information System (INIS)

    Craig, P.A.; Haelsig, R.T.; Kent, J.D.; Schmoker, D.S.

    1989-01-01

    A method is described of storing spent nuclear fuel assemblies including the steps of: transferring the fuel assemblies from a spent-fuel pool to a moveable concrete storage cask located outside the spent-fuel pool; maintaining a barrier between the fuel and the concrete in the cask to prevent contamination of the concrete by the fuel; maintaining the concrete storage cask containing the spent-fuel on site at the reactor complex for some predetermined period; transferring the fuel assemblies from the concrete storage cask to a shipping container; and, recycling the concrete storage cask

  10. Certifying the TN-BRP and TN-REG transportable storage demonstration casks

    International Nuclear Information System (INIS)

    Abbott, D.G.; Nolan, D.J.; Yoshimura, H.R.

    1991-01-01

    The US DOE has obtained US NRC certification to transport two transportable storage casks for a demonstration project. Because the casks had been built before the decision was made to obtain NRC certification, only limited modifications could be made to the casks. NRC's review resulted in several technical concerns that were subsequently resolved by design modifications, testing, and further analysis. Certification activities included qualifying the ferritic steel body material, modifying the borated stainless steel basket design, and extensive impact limiter testing. Recommendations for certifying future casks are presented based on experience with these casks

  11. 239Pu Resonance Evaluation for Thermal Benchmark System Calculations

    Science.gov (United States)

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

    2014-04-01

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

  12. Quality assurance in a cask fleet parts control system

    International Nuclear Information System (INIS)

    Fernandez, C.; Shappert, L.B.

    1992-01-01

    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

  13. Development of new type concrete for spent fuel storage cask

    International Nuclear Information System (INIS)

    Shimojo, J.; Mantani, K.; Owaki, E.; Sugihara, Y.; Hata, A.; Shimono, M.; Taniuchi, H.

    2004-01-01

    Heat resistant concrete has been developed to make it possible to design a new type cask that has been designed on the same concept of metal cask technologies for use in high temperature conditions. The allowable temperature of conventional concrete is limited to less than 100 degrees Celsius because most of its moisture is free water and therefore hydrogen, which is effective for neutron shielding, can be easily lost. Our newly developed concrete uses chemically bonded water and as a result can be used under high temperatures

  14. Benchmarking Data for the Proposed Signature of Used Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-23

    A set of benchmarking measurements to test facets of the proposed extended storage signature was conducted on May 17, 2016. The measurements were designed to test the overall concept of how the proposed signature can be used to identify a used fuel cask based only on the distribution of neutron sources within the cask. To simulate the distribution, 4 Cf-252 sources were chosen and arranged on a 3x3 grid in 3 different patterns and raw neutron totals counts were taken at 6 locations around the grid. This is a very simplified test of the typical geometry studied previously in simulation with simulated used nuclear fuel.

  15. Testing of ethylene propylene seals for the GA-4/GA-9 casks

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1993-01-01

    The primary O-ring seal of the GA-4 and GA-9 casks was tested for leakage with a full-scale mockup of the cask lid and flange. Tests were performed at temperatures of ambient, -41, 121, and 193 C. Shim plates between the lid and flange simulated gaps caused by thermal distortion. The testing used a helium mass spectrometer leak detector (MSLD). Results showed that the primary seal was leaktight for all test conditions. Helium permeation through the seal began in 13--23 minutes for the ambient tests and in 1--2 minutes for the tests at elevated temperatures. After each test several hours of pumping were typically required to reduce the MSLD background reading to an acceptable level for the next test, indicating that the seal had become saturated with helium. To verify that the test results showed permeation and not real leakage, several response checks were conducted in which a calibrated leak source was inserted in the detector line near the seal. When the leak source was activated the detector responded within seconds

  16. Structural challenges in the development of a truck shipping cask for the OCRWM cask systems development program

    International Nuclear Information System (INIS)

    Mello, R.M.; Severson, W.J.; Nair, B.R.

    1990-01-01

    The development of a spent fuel transportation cask design based on a structural material without licensing precedent presents many challenges. The U.S. Nuclear Regulatory Commission (NRC) requires that any new material be qualified to meet the design and fabrication requirements of the ASME Boiler and Pressure Vessel Code, Section III, Class 1. This paper discusses the strategy that is being implemented towards obtaining Code Acceptance of a titanium alloy (3A1-2.5V). This alloy has been chosen as the principal structural material for a Legal Weight Truck cask being developed by Westinghouse for the U.S. Department of Energy. The analysis approach used on some of the principal cask components is also presented

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

    International Nuclear Information System (INIS)

    Eslinger, L.E.; Schmitt, R.C.

    1989-01-01

    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

  18. Impact of more conservative cask designs of the CRWMS transportation system

    International Nuclear Information System (INIS)

    Joy, D.S.; Pope, R.B.; Johnson, P.E.

    1993-01-01

    The Office of Civilian Radioactive Waste Management has been working since the mid-1980s to develop a cask fleet, which will include legal weight truck and rail/barge casks for the transport of spent nuclear fuel (SNF) from reactors to Civilian Radioactive Waste Management System SNF receiving sites. The cask designs resulting from this effort have been identified as Initiative I casks. In order to maximize payloads, advanced technologies have been incorporated in the Initiative I cask designs, and some design margins have been reduced. Due to the wide range of the characteristics (age/burnup) of the spent fuel assemblies to be transported in the Initiative I casks, it has become apparent that a significant portion of the shipments of the Initiative I casks could not be loaded to their design capacity. Application of a more conventional cask design philosophy might result in new generation casks that would be easier to license, have more operational flexibility as to the range of age/burnup fuel that could be transported at full load, and be easier to fabricate. In general, these casks would have a lower capacity than the currently proposed Initiative I casks, thereby increasing the transportation impacts and the transportation costs

  19. Analysis of a hypothetical dropped spent nuclear fuel shipping cask impacting a floor mounted crush pad

    International Nuclear Information System (INIS)

    Hawkes, B.D.; Uldrich, E.D.

    1998-03-01

    A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped spent nuclear fuel shipping cask into a 44-ft. deep cask unloading pool at the Idaho Chemical Processing Plant. The 110-ton Large Cell Cask was assumed to be accidentally dropped onto the parapet of the unloading pool, causing the cask to tumble through the pool water and impact the floor mounted crush pad with the cask's top corner. The crush pad contains rigid polyurethane foam, which was modeled in a separate computer analysis to simulate the manufacturer's testing of the foam and to determine the foam's stress and strain characteristics. This computer analysis verified that the foam was accurately represented in the analysis to follow. A detailed non-linear, dynamic finite element analysis was then performed on the crush pad and adjacent pool structure to assure that a drop of this massive cask does not result in unacceptable damage to the storage facility. Additionally, verification was made that the crush pad adequately protects the cask from severe impact loading. At impact, the cask has significant vertical, horizontal and rotational velocities. The crush pad absorbs much of the energy of the cask through plastic deformation during primary and secondary impacts. After the primary impact with the crush pad, the cask still has sufficient energy to rebound and rotate until it impacts the pool wall. An assessment is made of the damage to the crush pad and pool wall and of the impact loading on the cask

  20. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    The traditional assumption used in evaluating criticality safety of spent fuel cask is that the spent fuel is as reactive as when it was fresh (new). This is known as the fresh fuel assumption. It avoids a number of calculational and verification difficulties, but could take a heavy toll in decreased efficiency. The alternative to the fresh fuel assumption is called burnup credit. That is, the reduced reactivity of spent fuel that comes about from depletion of fissile radionuclides and net increase in neutron absorbers (poisons) is taken into account. It is recognizable that the use of burnup credit will in fact increase the percentage of unacceptable or non-specification fuel available for misloading. This could reduce individual cask safety margins if current practices with respect to loading procedures are maintained. As such, additional operational, design, analysis, and validation requirements should be established that, as a minimum, compensate for any potential reduction in fuel loading safety margin. This method is based on a probabilistic (PRA) approach and is called a relative risk comparison. The method assumes a linear risk model, and uses a selected probability function to compare the system of interest and an acceptable reference system by varying the features of each to assess effects on system safety. While risk is the product of an event probability and its consequence, the consequences of criticality in a cask are considered to be both unacceptable and the same, regardless of the initiating sequence. Therefore, only the probability of the event is considered in a relative risk evaluation

  1. Experience with the loading and transport of fuel assembly transport casks, including CASTOR casks, and the radiation exposure of personnel

    International Nuclear Information System (INIS)

    Bentele, W.; Kinzelmann, T.

    1999-01-01

    In 1997 and 1998, six spent fuel assembly transports started from the nuclear power plant Gemeinschaftskernkraftwerk Neckar (GKN), using CASTOR-V19 casks. Professor Kuni of Marburg University challenged the statement made by the German Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz (BfS)) based on accepted scientific knowledge, according to which so-called CASTOR transports present no risk, either to the population or to the escorting police units. This paper shows that the collective dose during the loading of the CASTOR casks amounted to 4.5 mSv (gamma and neutrons) per cask at the most, and that the maximum individual dose amounted to 0.26 mSv. In addition to these doses, the collective dose during handling and transport must be considered: this amounted to 0.35 mSv (gamma and neutrons). The dose to the police escort was -2 (limit for surface contamination), presented degrees of contamination >4 Bq cm -2 upon reaching the Valognes/Cogema terminal. However, transport casks coming from French plants also revealed degrees of contamination >4 Bq cm -2 , as well as 'hot spots'. No such contamination was found on NTL 11 casks transported from the GKN to Sellafield. Neither was any increased contamination found upon the arrival of CASTOR-V19 casks transported from GKN to Gorleben or Ahaus. The partially sensationalist media reports were inversely proportional to the actual radiological relevance of the matter. The German Commission on Radiation Protection (SSK) confirmed that the radiological effect of such contaminated spent fuel transports is negligible. (author)

  2. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1992-09-01

    Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800 degree C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280 degree F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found

  3. AUTOCASK (AUTOmatic Generation of 3-D CASK models). A microcomputer based system for shipping cask design review analysis

    International Nuclear Information System (INIS)

    Gerhard, M.A.; Sommer, S.C.

    1995-04-01

    AUTOCASK (AUTOmatic Generation of 3-D CASK models) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for the structural analysis of shipping casks for radioactive material. Model specification is performed on the microcomputer, and the analyses are performed on an engineering workstation or mainframe computer. AUTOCASK is based on 80386/80486 compatible microcomputers. The system is composed of a series of menus, input programs, display programs, a mesh generation program, and archive programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests

  4. AUTOCASK (AUTOmatic Generation of 3-D CASK models). A microcomputer based system for shipping cask design review analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard, M.A.; Sommer, S.C. [Lawrence Livermore National Lab., CA (United States)

    1995-04-01

    AUTOCASK (AUTOmatic Generation of 3-D CASK models) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for the structural analysis of shipping casks for radioactive material. Model specification is performed on the microcomputer, and the analyses are performed on an engineering workstation or mainframe computer. AUTOCASK is based on 80386/80486 compatible microcomputers. The system is composed of a series of menus, input programs, display programs, a mesh generation program, and archive programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests.

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

    International Nuclear Information System (INIS)

    Ventura, Rodrigo; Ferreira, João; Filip, Iulian; Vale, Alberto

    2013-01-01

    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

  6. IMPACT ANALYSES AND TESTS OF CONCRETE OVERPACKS OF SPENT NUCLEAR FUEL STORAGE CASKS

    Directory of Open Access Journals (Sweden)

    SANGHOON LEE

    2014-02-01

    Full Text Available 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 [1], 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.

  7. Experimental evaluation of outer planets probe thermal insulation concepts

    Science.gov (United States)

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

    1976-01-01

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

  8. Evaluation of the thermal comfort of ceramic floor tiles

    Directory of Open Access Journals (Sweden)

    Carmeane Effting

    2007-09-01

    Full Text Available In places where people are bare feet, the thermal sensation of cold or hot depends on the environmental conditions and material properties including its microstructure and crustiness surface. The uncomforting can be characterized by heated floor surfaces in external environments which are exposed to sun radiation (swimming polls areas or by cold floor surfaces in internal environments (bed rooms, path rooms. The property named thermal effusivity which defines the interface temperature when two semi-infinite solids are putted in perfect contact. The introduction of the crustiness surface on the ceramic tiles interferes in the contact temperature and also it can be a strategy to obtain ceramic tiles more comfortable. Materials with low conductivities and densities can be obtained by porous inclusion are due particularly to the processing conditions usually employed. However, the presence of pores generally involves low mechanical strength. This work has the objective to evaluate the thermal comfort of ceramics floor obtained by incorporation of refractory raw materials (residue of the polishing of the porcelanato in industrial atomized ceramic powder, through the thermal and mechanical properties. The theoretical and experimental results show that the porosity and crustiness surface increases; there is sensitive improvement in the comfort by contact.

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

    Science.gov (United States)

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

    2012-11-01

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

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

    International Nuclear Information System (INIS)

    1983-11-01

    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

  11. Safety analysis report for packaging (onsite) for limited type Bmaterial in the CNS 14-215H cask

    International Nuclear Information System (INIS)

    Flanagan, B.D.

    1997-01-01

    The purpose of this Safety Analysis Report for Packaging is to provide the analyses and evaluations necessary to demonstrate that the CNS 14-215H cask provided by Chem-Nuclear Systems Inc. can safety transport greater than Type A quantities of radioactive material on the Hanford Site. The CNS 14-215H cask was chosen for its loading abilities, availability, and because it has a Certificate of Compliance (CoC) issued by the U.S. Nuclear Regulatory Commission (NRC) for transporting low specific activity in quantities greater than Type A material in commerce. Although the CDC does not cover greater than Type A material not meeting LSA requirements, it does allow for an established level of protection in determining the safety of transporting Type B material on the Hanford Site

  12. Evaluation of supercapacitors for space applications under thermal vacuum conditions

    Science.gov (United States)

    Chin, Keith C.; Green, Nelson W.; Brandon, Erik J.

    2018-03-01

    Commercially available supercapacitor cells from three separate vendors were evaluated for use in a space environment using thermal vacuum (Tvac) testing. Standard commercial cells are not hermetically sealed, but feature crimp or double seam seals between the header and the can, which may not maintain an adequate seal under vacuum. Cells were placed in a small vacuum chamber, and cycled between three separate temperature set points. Charging and discharging of cells was executed following each temperature soak, to confirm there was no significant impact on performance. A final electrical performance check, visual inspection and mass check following testing were also performed, to confirm the integrity of the cells had not been compromised during exposure to thermal cycling under vacuum. All cells tested were found to survive this testing protocol and exhibited no significant impact on electrical performance.

  13. Thermal analysis in the evaluation of sediment pollution.

    Science.gov (United States)

    Rodríguez-Barroso, M R; Ramírez-del Solar, M; Blanco, E; Quiroga, J M; García-Morales, J L

    2006-09-01

    Characterization of organic matter in four sediments in the infuence area of wastewater discharges was carried out by both chemical and thermal analysis in order to assess their pollution level. Oxidisable organic carbon and organic matter were calculated by the standard methodology in laboratory. Thermogravimetry (TG), between 50 and 900 degrees C, was simultaneously performed in oxidizing conditions on ground sediments samples after three different pretreatments. Linear regression adjust of thermal analysis results versus chemical parameters provides the better Pearson's coefficients, leading to the best coefficients for weight loss in 250-400 degree C temperature range versus oxidisable organic carbon and organic matter at 360 degrees C, respectively. These results demonstrated the utility of thermoanalysis technique for the evaluation of the organic matter content of fresh sediments.

  14. Design of casks: incorporating operational feedback from maintenance

    International Nuclear Information System (INIS)

    Bimet, F.; Hartenstein, M.

    2004-01-01

    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

  15. Software requirements definition Shipping Cask Analysis System (SCANS)

    International Nuclear Information System (INIS)

    Johnson, G.L.; Serbin, R.

    1985-01-01

    The US Nuclear Regulatory Commission (NRC) staff reviews the technical adequacy of applications for certification of designs of shipping casks for spent nuclear fuel. In order to confirm an acceptable design, the NRC staff may perform independent calculations. The current NRC procedure for confirming cask design analyses is laborious and tedious. Most of the work is currently done by hand or through the use of a remote computer network. The time required to certify a cask can be long. The review process may vary somewhat with the engineer doing the reviewing. Similarly, the documentation on the results of the review can also vary with the reviewer. To increase the efficiency of this certification process, LLNL was requested to design and write an integrated set of user-oriented, interactive computer programs for a personal microcomputer. The system is known as the NRC Shipping Cask Analysis System (SCANS). The computer codes and the software system supporting these codes are being developed and maintained for the NRC by LLNL. The objective of this system is generally to lessen the time and effort needed to review an application. Additionally, an objective of the system is to assure standardized methods and documentation of the confirmatory analyses used in the review of these cask designs. A software system should be designed based on NRC-defined requirements contained in a requirements document. The requirements document is a statement of a project's wants and needs as the users and implementers jointly understand them. The requirements document states the desired end products (i.e. WHAT's) of the project, not HOW the project provides them. This document describes the wants and needs for the SCANS system. 1 fig., 3 tabs

  16. Criticality safety of spent fuel casks considering water inleakage

    International Nuclear Information System (INIS)

    Osgood, N.L.; Withee, C.J.; Easton, E.P.

    2004-01-01

    A fundamental safety design parameter for all fissile material packages is that a single package must be critically safe even if water leaks into the containment system. In addition, criticality safety must be assured for arrays of packages under normal conditions of transport (undamaged packages) and under hypothetical accident conditions (damaged packages). The U.S. Nuclear Regulatory Commission staff has revised the review protocol for demonstrating criticality safety for spent fuel casks. Previous review guidance specified that water inleakage be considered under accident conditions. This practice was based on the fact that the leak tightness of spent fuel casks is typically demonstrated by use of structural analysis and not by physical testing. In addition, since a single package was shown to be safe with water inleakage, it was concluded that this analysis was also applicable to an array of damaged packages, since the heavy shield walls in spent fuel casks neutronically isolate each cask in the array. Inherent in this conclusion is that the fuel assembly geometry does not change significantly, even under drop test conditions. Requests for shipping fuel with burnup exceeding 40 GWd/MTU, including very high burnups exceeding 60 GWD/MTU, caused a reassessment of this assumption. Fuel cladding structural strength and ductility were not clearly predictable for these higher burnups. Therefore the single package analysis for an undamaged package may not be applicable for the damaged package. NRC staff developed a new practice for review of spent fuel casks under accident conditions. The practice presents two methods for approval that would allow an assessment of potential reconfiguration of the fuel assembly under accident conditions, or, alternatively, a demonstration of the water-exclusion boundary through physical testing

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

    International Nuclear Information System (INIS)

    Winsor, G.H.; Faletti, D.W.; DeSteese, J.G.

    1980-03-01

    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

  18. Anticipation of the needs linked with new generation reactors: COGEMA Logistics casks developments

    International Nuclear Information System (INIS)

    Issard, H.; Grygiel, J.M.

    2006-01-01

    to be used. To accompany the 'nuclear renaissance' and the related fuel design and fuel management evolution, it is crucial to anticipate the associated needs in terms of cask development notably for the spent fuel management. It is the reason why an ambitious R and D program has been set up at COGEMA Logistics. It aims at proposing innovative solutions oriented by the trends guiding this 'renaissance': the proposed systems have indeed to accommodate Spent Fuel Assemblies (SFAs) characterized by ever increasing burn-ups, fissile isotopes contents, and total inventory. Flexibility may potentially mean quick evacuation of UO 2 or MOX spent fuel with high thermal power to be dealt with. As described in the present paper, these evolutions directly guide the R and D actions on thermal and structural analysis, criticality and containment. The approach shall also include predictable licensing processes in an ever-demanding regulatory environment. The paper has the following structure: I. Evolutionary casks for evolving needs; II COGEMA Logistics R and D on evolutionary packagings; 1- The objectives of COGEMA Logistics R and D on evolutionary packagings; 2- Which R and D actions are currently implemented to anticipate evolving needs?; 2-1 Building an R and D program; 2-2 High performance design solutions for subcriticality; 2-3 Solutions for thermal and structural management; 2-4 Solution for enhanced shielding design; 2-5 Solutions for double containment systems; 2-6 Solutions for shock absorbers; 3- Examples of packaging developments; III Conclusion. To summarize, COGEMA Logistics is actively involved in Research and Development to accompany the improvement brought by Areva in fuel design and management linked either to extensive programs of nuclear plant life extension or to new constructions, such as Areva's European Pressurized Reactor. These evolutions coupled with needs for ever higher flexibility in terms of spent fuel management clearly guide the packaging

  19. PUNCTURE: a computer program for puncture analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

    In the drop puncture analyses for radioactive transport casks, it has become possible to perform them in detail by using interaction evaluation computer programs such as DYNA3D. However, the considerable cost and the computer time are necessitated to perform analyses by these programs. To decrease the computer cost and time, a simplified computer program PUNCTURE has been developed. The PUNCTURE is a static calculation computer program based on the Onat`s theory and Asada`s research. The PUNCTURE is capable of evaluating the acceleration of cask bodies, the deformation of punctured plates and, the stress and the deformation of puncture bars. Main features of the computer program are as follows; (1) three analysis models for punctured plate are used that are the fixed supported bending plate model, the simply supported bending plate model and the fixed supported membrane plate model, (2) it is capable of graphical representations for calculation results and (3) not only main frame computers (OS MSP) but also work stations (OS: UNIX) and personal computers (OS: Windows) are available for use of PUNCTURE. In the paper, brief illustration of calculation method using the Onat`s theory and Asada`s research is presented. The second section presents comparisons between calculation and experimental results. The third section provides a user`s guide for PUNCTURE. (author)

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

    International Nuclear Information System (INIS)

    Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H.

    2004-01-01

    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

  1. Development of heat-resistant neutron shielding resin for high payload metal cask

    International Nuclear Information System (INIS)

    Kamoshida, Mamoru; Hiranuma, Takeshi; Shimizu, Masashi

    2007-01-01

    A new neutron shielding resin has been developed for a dual-purpose metal cask. The resin is composed of a cycloaliphatic epoxy, anhydrous acid, catalyst, aluminum hydroxide and boron tetracarbide. Its long-term stability was verified by thermal degradation tests. Estimated weight loss of the resin during storage was about 1-2%. Because the curing reaction of epoxy and curing reagents was moderate at room temperature, a large amount of resin could be treated at one time which would lower fabrication cost. The fabrication process was verified by a full-scale mock-up test. No significant voids or cracks were found in the resin and uniform elemental composition was confirmed. (author)

  2. Multi-wall cask advantages with quarter-scale model drop test results for the NAC-STC Storable Transport Cask

    International Nuclear Information System (INIS)

    Danner, T.A.; Thompson, T.C.; Yaksh, M.C.

    1993-01-01

    Physical drop test results for a quarter-scale model multi-wall cask are presented for the 9 meter end, side and oblique drops with impact limiters, and for the 1 meter side and closure lid pin puncture drops. Lessons learned and final cask test qualification are presented. (J.P.N.)

  3. Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ross, Steven [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Grey, Carissa Ann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arviso, Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Garmendia, Rafael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Fernandez Perez, Ismael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Palacio, Alejandro [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Calleja, Guillermo [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Garrido, David [COORDINADORA, Madrid (Spain); Rodriguez Casas, Ana [COORDINADORA, Madrid (Spain); Gonzalez Garcia, Luis [COORDINADORA, Madrid (Spain); Chilton, Lyman Wes [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walz, Jacob [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gershon, Sabina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klymyshyn, Nicholas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pena, Ruben [Transportation Technology Center, Inc., Pueblo, CO (United States); Walker, Russell [Transportation Technology Center, Inc., Pueblo, CO (United States)

    2018-01-01

    This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

  4. Thermal characteristic of insulation for optimum design of RI transport package

    International Nuclear Information System (INIS)

    Lee, J. C.; Bang, K. S.; Seo, K. S.

    2002-01-01

    A package to transport the high level radioactive materials in required to withstand the hypothetical accident conditions as well as normal transport conditions according to IAEA and domestic regulations. The regulations require that the package should maintain the shielding, thermal and structural integrities to release no radioactive material. Thermal characteristics of insulations were evaluated and optimum insulation thickness was deduced for RI transport package. The package has a maximum capacity of 600 Curies for Ir-192 sealed source. The insulation thickness was decided with 10 mm of polyurethane form to maintain the thermal safety under fire accident condition. Thermal analysis was carried out for RI transport package, and it was shown that the thermal integrity of the package was maintained. The results obtained this study will be applied to a basic data for design of RI transport cask

  5. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1990-03-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing, These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale of neutron shield from the cask. The test article was heated in an environmental prescribed by NRC regulations. Results of this second testing phase showed that all three materials are thermally acceptable

  6. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.N.

    1990-01-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing. These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale section of neutron shield from the cask. The test article was heated in an environment prescribed by NRC regulations. Results of this second testing phase show that all three materials are thermally acceptable

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-11-01

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

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

    Science.gov (United States)

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

    1992-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Josten, N.E.

    1992-03-01

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

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

    International Nuclear Information System (INIS)

    Josten, N.E.

    1992-03-01

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

  11. 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.

  12. A preliminary assessment of system cost impacts of using transportable storage casks and other shippable metal casks in the utility/DOE spent fuel management system

    International Nuclear Information System (INIS)

    Johnson, E.R.

    1988-01-01

    In view of the foregoing, a study was conducted by E.R. Johnson Associates, Inc. and H and R Technical Associates, Inc. to determine the prospective viability of the use of TSCs and shippable SOCs in the combined utility/DOE system. This study considered costs, ALARA considerations and the logistics of the use and delivery of casks to the DOE system by utilities. It was intended that this study would result in a technical and cost resource base that could be used for evaluating various strategies and scenarios for deploying TSCs or SOCs in the combined utility/DOE spent fuel management system with respect to the prospective economic advantage that could be realized

  13. Application of linear-elastic fracture mechanics concepts to ferritic spent fuel shipping casks

    International Nuclear Information System (INIS)

    McConnell, P.; Wullaert, R.A.; Trujillo, A.

    1983-01-01

    A linear-elastic fracture mechanics analysis is an appropriate methodology for the fail-safe design of spent fuel shipping containers. It provides a quantitative basis by which to assess the margin of safety inherent to a particular cask. Required inputs are the maximum stresses to be expected from a nine-meter drop, the largest flaw size that may be expected to exist in the cask (based upon capabilities of NDE), and the fracture toughness of the cask material. It was analytically, and conservatively, demonstrated using linear-elastic fracture mechanics procedures that two typical shipping cask designs, constructed of candidate ferritic steel and nodular cast iron materials, can withstand a nine-meter drop. No crack initiation would occur even should a flaw of a size well above the detectability limit of conventional NDE exist in the cask. The fracture toughness of even relatively low toughness ferritic materials is sufficient to ensure the structural integrity of these casks

  14. Subjective evaluation to thermal environment. Onnetsu kankyo no shukanteki hyoka

    Energy Technology Data Exchange (ETDEWEB)

    Taya, H. (Human Factors Reserach Department, Tsukuba (Japan))

    1991-11-10

    This paper introduces a subjective evaluation on a thermal environment. Providing human being with an appropriate and pleasant environment requires considerations from three aspects of economy, health, and comfort. Related with these are the productively optimal temperature, the physiologically optimal temperature, and the subjectively optimal temperature. While a human being makes subjective judgments on his or her environment through physiological reactions, the kinds of these subjective judgments include those on perception, evaluation, preference, receptivity, and permissibility. The subjective evaluation is measured by declaration of subjects. An evaluation method using a category scale or an SD scale is used for the measurement. The evaluation on cold-hot feeling and comfort using the category scale and that using the SD scale are explained, whereas the evaluation on preference relative to environmental temperatures is carried out by rating that the present environment should be made warmer'', can remain unchanged'' or should be made cooler''. 22 refs., 5 figs., 5 tabs.

  15. Radiological source terms resulting from sabotage to transportation casks: Final report

    International Nuclear Information System (INIS)

    Miller, N.E.; Fentiman, A.W.; Kuhlman, M.R.; Ebersole, H.N.; Trott, B.D.; Orban, J.E.

    1986-11-01

    The Nuclear Regulatory Commission (NRC) promulgated a rule, 10 CFR 73.37, which established requirements for safeguarding shipments of spent fuel to reduce the risk from acts of sabotage of highly radioactive materials. After the rule became effective, experimental programs conducted by Battelle for the NRC and by Sandia for the DOE showed the consequences of an attack using explosives on a shipment of PWR spent fuel were significantly less than had been indicated by earlier analytical studies. As a result, NRC is considering modifying the safeguards requirements. In support of NRC's efforts to modify the rule, Battelle has conducted additional experimental studies to evaluate the consequences of attacks on shipments of high-temperature gas-cooled reactor (HTGR) spent fuel, nonpower reactor (NPR) spent fuel, and vitrified high-level waste (HLW). Model casks containing surrogates of the spent fuels or high-level waste were penetrated by the jet from a precision shaped charge. Air samples collected after each test were used to estimate the quantities of respirable material released after the cask was penetrated. Results of the tests were scaled by specially developed scaling factors to estimate the releases that may occur from attacks on full-sized shipments of the materials. It was concluded that the sabotage of shipments of HTGR spent fuel, NPR spent fuel, or vitrified HLW should have no greater consequences than those predicted for shipments of PWR spent fuel

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

    International Nuclear Information System (INIS)

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

    1990-01-01

    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)

    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)

  18. 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...

  19. Literature review of cask exterior surface contamination with application to a nuclear repository

    International Nuclear Information System (INIS)

    Adams, K.G.

    1990-04-01

    The characteristics of transportation cask surface contamination and the ''weeping'' phenomenon are reviewed. In addition, literature that pertains to the possible impact of surface contamination on repository operations is reviewed. This report consists of commentary on and rather extensive excerpts from the published literature on cask surface contamination that may have a bearing on repository risk assessment. In addition, estimates are made of the quantity of contamination that might be present on a cask. These estimates are used to calculate the direct exposure rates to personnel located at several distances from the cask. 15 refs., 15 tabs

  20. Transport experience of NH-25 spent fuel shipping cask for post irradiation examination

    International Nuclear Information System (INIS)

    Mori, Ryuji

    1982-01-01

    Since the Japan Atomic Energy Research Institute and Nippon Nuclear Fuel Development Co. hot laboratories are located far off from the port which can handle spent fuel shipping casks, it is necessary to use a trailer-mounted cask which can be transported by public roads, bridges and intersections for the transportation of spent fuel specimens to these hot laboratories. Model NH-25 shipping cask was designed, manufactured and oualification tested to meet Japanese regulations and was officially registered as a BM type cask. The NH-25 cask accomodates two BWR fuel assemblies, one PWR assembly or one ATR fuel assembly using interchangeable inner containers. The cask weight is 29.2 t. The cask has three concentric stainless steel shells. Gamma shielding is lead cast between the inner shell and the intermediate shell. Neutro n shielding consists of ethylene-glycol-aqueous solution layer formed between the intermediate shell and the outer shell. The NH-25 cask now has been in operation for 2.5 yr. It was used for the transportation of spent fuel assemblies from six LWR power plants to the port on shipping cask carrier ''Hinouramaru'' on the sea, as well as from the port to the hot laboratory on a trailer. The capability of safe handling and transporting of spent fuel assemblies has been well demonstrated. (author)

  1. Status of the Virginia Power/DOE Cooperative Cask Testing/Demonstration Program: A video presentation

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Creer, J.M.; Collantes, C.E.

    1990-01-01

    This paper is documentation of a video presentation and provides a brief summary of the Virginia power/US Department of Energy Cooperative Cask Testing/Demonstration Program. The program consists of two phases. The first phase has been completed and involved the unlicensed performance testing (heat transfer and shielding) of three metal spent fuel storage casks at the federally owned Idaho National Engineering Laboratory. The second phase is ongoing and consists of licensed demonstrations of standard casks from two different vendors and of one or two enhanced capacity casks. 6 refs., 1 tab

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

    International Nuclear Information System (INIS)

    Ferrell, P.C.

    1997-01-01

    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

  3. Impact response analysis of cask for spent fuel by dimensional analysis and mode superposition method

    International Nuclear Information System (INIS)

    Kim, Y. J.; Kim, W. T.; Lee, Y. S.

    2006-01-01

    Full text: Full text: Due to the potentiality of accidents, the transportation safety of radioactive material has become extremely important in these days. The most important means of accomplishing the safety in transportation for radioactive material is the integrity of cask. The cask for spent fuel consists of a cask body and two impact limiters generally. The impact limiters are attached at the upper and the lower of the cask body. The cask comprises general requirements and test requirements for normal transport conditions and hypothetical accident conditions in accordance with IAEA regulations. Among the test requirements for hypothetical accident conditions, the 9 m drop test of dropping the cask from 9 m height to unyielding surface to get maximum damage becomes very important requirement because it can affect the structural soundness of the cask. So far the impact response analysis for 9 m drop test has been obtained by finite element method with complex computational procedure. In this study, the empirical equations of the impact forces for 9 m drop test are formulated by dimensional analysis. And then using the empirical equations the characteristics of material used for impact limiters are analysed. Also the dynamic impact response of the cask body is analysed using the mode superposition method and the analysis method is proposed. The results are also validated by comparing with previous experimental results and finite element analysis results. The present method is simpler than finite element method and can be used to predict the impact response of the cask

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

    International Nuclear Information System (INIS)

    Fiarman, S.

    1987-01-01

    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)

  5. Thermal Conductivity of EB-PVD Thermal Barrier Coatings Evaluated by a Steady-State Laser Heat Flux Technique

    Science.gov (United States)

    Zhu, Dongming; Miller, Robert A.; Nagaraj, Ben A.; Bruce, Robert W.

    2000-01-01

    The thermal conductivity of electron beam-physical vapor deposited (EB-PVD) Zr02-8wt%Y2O3 thermal barrier coatings was determined by a steady-state heat flux laser technique. Thermal conductivity change kinetics of the EB-PVD ceramic coatings were also obtained in real time, at high temperatures, under the laser high heat flux, long term test conditions. The thermal conductivity increase due to micro-pore sintering and the decrease due to coating micro-delaminations in the EB-PVD coatings were evaluated for grooved and non-grooved EB-PVD coating systems under isothermal and thermal cycling conditions. The coating failure modes under the high heat flux test conditions were also investigated. The test technique provides a viable means for obtaining coating thermal conductivity data for use in design, development, and life prediction for engine applications.

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

    International Nuclear Information System (INIS)

    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 Caskclose 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

  7. Legal weight truck cask model impact limiter response

    International Nuclear Information System (INIS)

    Meinert, N.M.; Shappert, L.B.

    1989-01-01

    Dynamic and quasi-static quarter-scale model testing was performed to supplement the analytical case presented in the Nuclear Assurance Corporation Legal Weight Truck (NAC LWT) cask transport licensing application. Four successive drop tests from 9.0 meters (30 feet) onto an unyielding surface and one 1.0-meter (40-inch) drop onto a scale mild steel pin 3.8 centimeters (1.5 inches) in diameter, corroborated the impact limiter design and structural analyses presented in the licensing application. Quantitative measurements, made during drop testing, support the impact limiter analyses. High-speed photography of the tests confirm that only a small amount of energy is elastically stored in the aluminum honeycomb and that oblique drop slapdown is not significant. The qualitative conclusion is that the limiter protected LWT cask will not sustain permanent structural damage and containment will be maintained, subsequent to a hypothetical accident, as shown by structural analyses

  8. Preliminary design report for the NAC combined transport cask

    International Nuclear Information System (INIS)

    1990-04-01

    Nuclear Assurance Corporation (NAC) is under contract to the United States Department of Energy (DOE) to design, license, develop and test models, and fabricate a prototype cask transportation system for nuclear spent fuel. The design of this combined transport (rail/barge) transportation system has been divided into two phases, a preliminary design phase and a final design phase. This Preliminary Design Package (PDP) describes the NAC Combined Transport Cask (NAC-CTC), the results of work completed during the preliminary design phase and identifies the additional detailed analyses, which will be performed during final design. Preliminary analytical results are presented in the appropriate sections and supplemented by summaries of procedures and assumptions for performing the additional detailed analyses of the final design. 60 refs., 1 fig., 2 tabs

  9. Stress analysis of closure bolts for shipping casks

    International Nuclear Information System (INIS)

    Mok, G.C.; Fischer, L.E.; Hsu, S.T.

    1993-01-01

    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

  10. Impact analysis of spent nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Huerta, M.; Dennis, A.W.; Yoshimura, R.H.

    1978-07-01

    A full-scale testing program involving impact tests of spent-nuclear-fuel shipping systems is described. The program is being conducted by Sandia Laboratories for the Environmental Control Technology Division of the U.S. Energy Research and Development Administration. The paper describes the analytical and scale modeling techniques being employed to predict the response of the full-scale system in the very severe impact tests. The analytical techniques include lumped parameter modeling of the vehicle and cask system and finite modeling of isolated shipping casks. Some preliminary results from the mathematical analyses and scale model tests demonstrate close agreement between these two techniques. Scale models of the systems are also described and some results presented

  11. Conceptual design report for a remotely operated cask handling system

    International Nuclear Information System (INIS)

    Yount, J.A.; Berger, J.D.

    Recent advances in remote handling utilizing commercial robotics are conceptually applied to the problem of lowering operator cumulative dose and increasing throughput during cask handling operations in proposed nuclear waste container shipping and receiving facilities. The functional criteria for each subsystem are defined, and candidate systems are described. The report also contains a generic description of a waste receiving facility, to show possible deployment configurations for the equipment

  12. Brittle fracture tests at low temperature for transport cask materials

    International Nuclear Information System (INIS)

    Kosaki, Akio; Ito, Chihiro; Arai, Taku; Saegusa, Toshiari

    1993-01-01

    The IAEA Regulations for the Safe Transport of Radioactive Material were revised in 1985, and brittle fracture assessment at low temperature for transport packages are now required. This report discusses the applicability of the actual method for brittle fracture assessment of type-B transport cask materials used in JAPAN. The necessity of brittle fracture assessment at low temperature was estimated for each material of type-B transport casks used in Japan and the applicability was investigated. Dynamic fracture toughness values, K Id (J Id ), and RT NDT values of Low-Mn Carbon Steels, that are SA 350 Gr.LF1 Modify and SA 516 Gr.70 material which used in type-B transport cask body, were also obtained to check whether or not an easier and conventional test method, that prescribed in ASME CODE SECTION III, can be substituted for the dynamic fracture test method. And for bolt materials, which include 1.8Ni-0.8Cr-0.3Mo Carbon Steel and type 630 H Stainless Steel, toughness data were obtained for reference. (J.P.N.)

  13. Thermal Performance Evaluation of Walls with Gas Filled Panel Insulation

    Energy Technology Data Exchange (ETDEWEB)

    Shrestha, Som S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Desjarlais, Andre Omer [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Atchley, Jerald Allen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-11-01

    Gas filled insulation panels (GFP) are very light weight and compact (when uninflated) advanced insulation products. GFPs consist of multiple layers of thin, low emittance (low-e) metalized aluminum. When expanded, the internal, low-e aluminum layers form a honeycomb structure. These baffled polymer chambers are enveloped by a sealed barrier and filled with either air or a low-conductivity gas. The sealed exterior aluminum foil barrier films provide thermal resistance, flammability protection, and properties to contain air or a low conductivity inert gas. This product was initially developed with a grant from the U.S. Department of Energy. The unexpanded product is nearly flat for easy storage and transport. Therefore, transportation volume and weight of the GFP to fill unit volume of wall cavity is much smaller compared to that of other conventional insulation products. This feature makes this product appealing to use at Army Contingency Basing, when transportation cost is significant compared to the cost of materials. The objective of this study is to evaluate thermal performance of walls, similar to those used at typical Barracks Hut (B-Hut) hard shelters, when GFPs are used in the wall cavities. Oak Ridge National Laboratory (ORNL) tested performance of the wall in the rotatable guarded hotbox (RGHB) according to the ASTM C 1363 standard test method.

  14. 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)

    Johnson, P.E.; Joy, D.S.; Pope, R.B.; Shappert, L.B.; Wankerl, M.W.; Best, R.E.; Schmid, S.; Danese, F.L.

    1992-01-01

    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

  15. 76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-06-08

    ... Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear Regulatory Commission. ACTION: Direct final... regulations to add the Holtec HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage... Title 10 of the Code of Federal Regulations Section 72.214 to add the Holtec HI- STORM Flood/Wind cask...

  16. Feasibility study for scaling ferritic spent fuel casks for drop tests

    International Nuclear Information System (INIS)

    McConnell, P.; Jones, J.W.; Nickell, R.E.

    1983-01-01

    Within appropriate constraints, scale-model drop testing for demonstrating cask integrity is technically and practically feasible. Structural response and potential for crack initiation (fracture toughness) can be scaled. The flaw sizes assumed in this scaling analysis are a significant proportion of the cask wall thickness. The minimum depth considered was one-tenth the wall thickness. Detection of such flaw sizes are well within the capability of conventional NDE. The stress values analytically calculated are believed to realistically represent those a cask would experience in a 9-meter side-drop. Since this first phase of the EPRI/Sandia TTC cask scaling program analytically demonstrated the feasibility for scaling, it remains to actually drop test a properly instrumented scale model cask to experimentally demonstrate the approach. This model cask should have a defect machined into the highest stress location. The full-size cask should be analytically characterized prior to drop testing the model. Ideally, a drop test would eventually be conducted on the prototype to confirm the scaling concept. In addition to demonstrating the scaling concept, a successful drop test of a cask, be it model or prototype, would also demonstrate the potential of a linear-elastic fracture mechanics approach for ensuring structural integrity. 7 references, 3 figures

  17. 77 FR 9515 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule... regulations by revising the Holtec International HI-STORM 100 dry cask storage system listing within the... and safety will be adequately protected. This direct final rule revises the HI-STORM 100 listing in 10...

  18. CASK inhibits ECV304 cell growth and interacts with Id1

    International Nuclear Information System (INIS)

    Qi Jie; Su Yongyue; Sun Rongju; Zhang Fang; Luo Xiaofeng; Yang Zongcheng; Luo Xiangdong

    2005-01-01

    Calcium/calmodulin-dependent serine protein kinase (CASK) is generally known as a scaffold protein. Here we show that overexpression of CASK resulted in a reduced rate of cell growth, while inhibition of expression of endogenous CASK via RNA-mediated interference resulted in an increased rate of cell growth in ECV304 cells. To explore the molecular mechanism, we identified a novel CASK-interacting protein, inhibitor of differentiation 1 (Id1) with a yeast two-hybrid screening. Furthermore, endogenous CASK and Id1 proteins were co-precipitated from the lysates of ECV304 cells by immunoprecipitation. Mammalian two-hybrid protein-protein interaction assays indicated that CASK possessed a different binding activity for Id1 and its alternative splicing variant. It is known that Id proteins play important roles in regulation of cell proliferation and differentiation. Thus, we speculate that the regulation of cell growth mediated by CASK may be involved in Id1. Our findings indicate a novel function of CASK, the mechanism that remains to be further investigated

  19. Cask size and weight reduction through the use of depleted uranium dioxide-concrete material

    International Nuclear Information System (INIS)

    Lobach, S.Yu.; Haire, J.M.

    2007-01-01

    Newly developed depleted uranium (DU) composite materials enable fabrication of spent nuclear fuel (SNF) transport and storage casks that are smaller and lighter in weight than casks made with conventional materials. One such material is DU dioxide (DUO2)-concrete, so-called DUCRETE TM . This paper examines the radiation shielding efficiency of DUCRETE as compared with that of a conventional concrete cask that holds 32 pressurized-water-reactor SNF assemblies. In this analysis, conventional concrete shielding material is replaced with DUCRETE. The thickness of the DUCRETE shielding is adjusted to give the same radiation surface dose, 200 mrem/h (2 mSv/hr), as the conventional concrete cask. It was found that the concrete shielding thickness decreased from 71 to 20 cm and that the cask radial cross-section shielding area was reduced approx 50 %. The weight was reduced approx 21 %, from 154 to approx 127 tons. Should one choose to add an extra outer ring of SNF assemblies, the number of such assemblies would increase from 32 to 52. In this case, the outside cask diameter would still decrease, from 169 to 137 cm. However, the weight would increase somewhat from 156 to 177 tons. Neutron cask surface dose is only approx 10 % of the gamma dose. These reduced sizes and weights will significantly influence the design of next-generation SNF casks

  20. Man/machine interface for a nuclear cask remote handling control station: system design requirements

    International Nuclear Information System (INIS)

    Clarke, M.M.; Kreifeldt, J.G.; Draper, J.V.

    1984-01-01

    Design requirements are presented for a control station of a proposed semi-automated facility for remote handling of nuclear waste casks. Functional and operational man/machine interface: controls, displays, software format, station architecture, and work environment. In addition, some input is given to the design of remote sensing systems in the cask handling areas. 18 references, 9 figures, 12 tables

  1. Evaluation of the Thermal Effects in Tilting Pad Bearing

    Directory of Open Access Journals (Sweden)

    G. B. Daniel

    2013-01-01

    Full Text Available The analysis of thermal effects is of expressive importance in the context of rotordynamics to evaluate the behavior of hydrodynamic bearings because these effects can influence their dynamic characteristics under specific operational conditions. For this reason, a thermohydrodynamic model is developed in this work, in which the pressure distribution in the oil film and the temperature distribution are calculated together. From the pressure distribution, the velocity distribution field is determined, as well as the viscous dissipation, and consequently, the temperature distribution. The finite volume method is applied to solve the Reynolds equation and the energy equation in the thermohydrodynamic model (THD. The results show that the temperature is higher as the rotational speed increases due to the shear rate of the oil film. The maximum temperature in the bearing occurs in the overloaded pad, near the outlet boundary. The experimental tests were performed in a tilting pad journal bearing operating in a steam turbine to validate the model. The comparison between the experimental and numerical results provides a good correlation. The thermohydrodynamic lubrication developed in this assignment is promising to consistently evaluate the behavior of the tilting pad journal bearing operating in relatively high rotational speeds.

  2. 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.

  3. The external dose of lack fuel cask for analyses with MCNP

    International Nuclear Information System (INIS)

    Liu Liu; Qiu Xiaoping; Liao Lingyuan

    2009-01-01

    The transport vessel of lack fuel cask is a special facilities which is for reactor lack fuel transportation. MCNP4C is used to count the external dose rate of Westinghouse MC-10 lack fuel cask, it is based on mesh definition, to get the whole external dose rate of the cask, and in connection with the result from previous researcher Georgeta Radulescu, the outcome in consistency is good, using mesh causes long-playing machine hours and comes to some error, but it can get many data about external dose rate of the lack fuel cask roundly and at any rate. So it makes sense to the definition on the external dose rate of the lack fuel cask for missionary. (authors)

  4. A revision of the cask designers guide for the '90s

    International Nuclear Information System (INIS)

    Shappert, L.B.; Green, V.M.

    1993-01-01

    DOE has requested that ORNL initiate a revision to NSIC-68, A Guide for the Design, Fabrication, and Operation of Shipping Casks for Nuclear Applications, commonly called the Cask Designers Guide. This revision, called the Cask Handbook, has two goals: (1) to improve the quality of SARPs that are submitted to DOE, and (2) to provide up-to-date information on the design of spent fuel shipping casks, including information on fabrication, quality assurance, SARP preparation, certification, use, maintenance, and other general topics. The revision provides guidance that will help engineers through the cask licensing process, in part, by providing as much regulator-approved data and 'lessons-learned' information as possible. The effort is sponsored by DOE-Environmental, Safety and Health (EH), guided by Transportation Technology staff members at ORNL, and the information is being generated by experts in the various technical fields. (J.P.N.)

  5. Verification of maximum impact force for interim storage cask for the Fast Flux Testing Facility

    International Nuclear Information System (INIS)

    Chen, W.W.; Chang, S.J.

    1996-01-01

    The objective of this paper is to perform an impact analysis of the Interim Storage Cask (ISC) of the Fast Flux Test Facility (FFTF) for a 4-ft end drop. The ISC is a concrete cask used to store spent nuclear fuels. The analysis is to justify the impact force calculated by General Atomics (General Atomics, 1994) using the ILMOD computer code. ILMOD determines the maximum force developed by the concrete crushing which occurs when the drop energy has been absorbed. The maximum force, multiplied by the dynamic load factor (DLF), was used to determine the maximum g-level on the cask during a 4-ft end drop accident onto the heavily reinforced FFTF Reactor Service Building's concrete surface. For the analysis, this surface was assumed to be unyielding and the cask absorbed all the drop energy. This conservative assumption simplified the modeling used to qualify the cask's structural integrity for this accident condition

  6. Analysis of collective life-cycle dose for burnup credit shipping casks

    International Nuclear Information System (INIS)

    Brentlinger, L.A.; Peterson, R.W.; Hofmann, P.L.

    1989-01-01

    In 1987, several studies were conducted by Sandia National Laboratories (SNL) to investigate the feasibility of and the incentive to justify the consideration of spent fuel histories in the design of spent fuel shipping casks. Taking credit for reduction in fissile content of fuel elements resulting from burnup credit is not current practice in the design and certification of shipping casks. The general argument can be made, however, that if this were done cask capacities could be increased over the current shipping cask designs which do not take the benefit of such burnup credit. This paper deals specifically with the question of occupational and public dose reduction via the use of a series of postulated burnup-credit cask designs

  7. Influences in Thermal Conductivity Evaluation Using the Thermal Probe Method; some Practical Aspects

    OpenAIRE

    Strâmbu, Vasile

    2012-01-01

    The thermal probe is a device used for measuring the thermal conductivity of materials in the food industry, plastics industry, geotechnical engineering and studies of soft soils and rocks. The method also started being utilized in the field of construction materials with particularities that take into account their composition and the state they are in.

  8. Summary of the technical review of the safety analysis reports for packaging (SARP) for the transnuclear transport/storage casks: TN-BRP and TN-REG

    International Nuclear Information System (INIS)

    1986-07-01

    The Safety Analysis Reports for Packaging for two spent fuel shipping casks were technically reviewed by the Oak Ridge National Laboratory. The casks were designed by Transnuclear, Inc., for shipment of 85 Big Rock Point boiling water reactor fuel elements and 40 R.E. Ginna pressurized water reactor fuel elements from West Valley, New York, to Idaho Falls, Idaho. The intent of the review was to ensure compliance of the casks with the requirements the applicable Federal Regulations contained in 10 CFR Pt. 71 and allow issuance of Department of Energy Certificates of Compliance for transport by the Department of Energy Idaho Operations Office. The review was performed by a team of Oak Ridge National Laboratory staff assembled for their expertise in criticality analysis, shielding, metallurgy, nondestructive testing, thermal analysis, structural analysis, and containment. This report describes the review processes, the findings in each technical area, and the overall conclusion that a Certificate of Compliance could be issued for the proposed single shipment under the specified conditions and constraints

  9. Evaluation of the effectiveness of a thermal hygienization reactor

    Directory of Open Access Journals (Sweden)

    Daniel Borski

    2011-01-01

    Full Text Available For reasons of limiting the spread of serious transmissible diseases, with regard to the requirement for reducing landfill of biodegradable waste (which may or contains animal by-products and thus presents a potential risk to human and animal health and with a focus on supporting its separate collection, there has been created a legal framework for processing and hygienization of materials containing animal by-products. For the above reasons new technologies are being developed and implemented. These technologies are able to ensure the processing of biological waste containing animal by-products. As a practical result of the effort to ensure the hygienization of biowaste, a hygienization unit of own design, which uses the thermal way of hygienization, is presented in this work. The general part of the work defines a legislative framework for the assignment and gives technical parameters and minimum requirements for conversion that hygienization unit should be able to perform, including the limits for digestion residues and compost.In the experimental section there are described operational tests which document the technological process of hygienization depending on the aeration of the contents of the reactor. Experiment III outlines the validation process which uses contamination by indicator organisms, including subsequent checking of their occurrence as well as processing of the results of experiments and evaluation of the process of hygienization.

  10. Capability of environmental sampling to detect undeclared cask openings

    International Nuclear Information System (INIS)

    Beckstead, L.W.; Efurd, D.W.; Hemberger, P.H.; Abhold, M.E.; Eccleston, G.W.

    1997-01-01

    The goal of this study is to determine the signatures that would allow monitors to detect diversion of nuclear fuel (by a diverter) from a storage area such as a geological repository. Due to the complexity of operations surrounding disposal of spent nuclear fuel in a geologic repository, there are several places that a diversion of fuel could take place. After the canister that contains the fuel rods is breached, the diverter would require a hot cell to process or repackage the fuel. A reference repository and possible diversion scenarios are discussed. When a canister is breached, or during reprocessing to extract nuclear weapons material (primarily Pu), several important isotopes or signatures including tritium, 85 Kr, and 129 I are released to the surrounding environment and have the potential for analysis. Estimates of release concentrations of the key signatures from the repository under a hypothetical diversion scenario are presented and discussed. Gas analysis data collected from above-ground storage casks at Idaho National Engineering and Environmental Laboratory (INEEL) Test Area North (TAN) are included and discussed in the report. In addition, LANL participated in gas sampling of one TAN cask, the Castor V/21, in July 1997. Results of xenon analysis from the cask gas sample are presented and discussed. The importance of global fallout and recent commercial reprocessing activities and their effects on repository monitoring are discussed. Monitoring and instrumental equipment for analysis of the key signature isotopes are discussed along with limits of detection. A key factor in determining if diversion activities are in progress at a repository is the timeliness of detection and analysis of the signatures. Once a clandestine operation is suspected, analytical data should be collected as quickly as possible to support any evidence of diversion

  11. Capability of environmental sampling to detect undeclared cask openings

    Energy Technology Data Exchange (ETDEWEB)

    Beckstead, L.W.; Efurd, D.W.; Hemberger, P.H.; Abhold, M.E.; Eccleston, G.W.

    1997-12-01

    The goal of this study is to determine the signatures that would allow monitors to detect diversion of nuclear fuel (by a diverter) from a storage area such as a geological repository. Due to the complexity of operations surrounding disposal of spent nuclear fuel in a geologic repository, there are several places that a diversion of fuel could take place. After the canister that contains the fuel rods is breached, the diverter would require a hot cell to process or repackage the fuel. A reference repository and possible diversion scenarios are discussed. When a canister is breached, or during reprocessing to extract nuclear weapons material (primarily Pu), several important isotopes or signatures including tritium, {sup 85}Kr, and {sup 129}I are released to the surrounding environment and have the potential for analysis. Estimates of release concentrations of the key signatures from the repository under a hypothetical diversion scenario are presented and discussed. Gas analysis data collected from above-ground storage casks at Idaho National Engineering and Environmental Laboratory (INEEL) Test Area North (TAN) are included and discussed in the report. In addition, LANL participated in gas sampling of one TAN cask, the Castor V/21, in July 1997. Results of xenon analysis from the cask gas sample are presented and discussed. The importance of global fallout and recent commercial reprocessing activities and their effects on repository monitoring are discussed. Monitoring and instrumental equipment for analysis of the key signature isotopes are discussed along with limits of detection. A key factor in determining if diversion activities are in progress at a repository is the timeliness of detection and analysis of the signatures. Once a clandestine operation is suspected, analytical data should be collected as quickly as possible to support any evidence of diversion.

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

    International Nuclear Information System (INIS)

    Garcia, G.C.R.; Santos, E.M.B.; Ribeiro, S.; Rodrigues, J.A.

    2011-01-01

    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)

  13. Thermal fatigue evaluation of piping system Tee-connections

    International Nuclear Information System (INIS)

    Metzner, K.J.; Braillard, O.; Faidy, C.; Marcelles, I.; Solin, J.; Stumpfrock, L.

    2004-01-01

    Thermal fatigue is one significant long-term degradation mechanism 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 has been initiated to advance the accuracy and reliability of thermal fatigue load determination in engineering tools and research oriented approaches to outline a science based practical methodology for managing thermal fatigue risks in Tee-connections susceptible to high cyclic thermal fatigue. (orig.)

  14. Development testing of a nuclear waste cask remote handling system

    International Nuclear Information System (INIS)

    Gneiting, B.C.; Swannack, D.L.; Berger, J.D.; Allen, G.C. Jr.

    1985-01-01

    Radioactive waste shipping and receiving facilities presently planned for commercial and defense nuclear waste will handle waste packages at frequencies far in excess of those in common practice today. High radiation exposures and large personnel staffs would be necessary if current handling methods were used. To reduce personnel exposures and man-power requirements, alternate handling methods are being developed and demonstrated. Proof-of-principle testing of remote handling techniques using robotics demonstrated nearly all critical operations for cask receipt, preparation and unloading. 1 figure

  15. Design of a transportation cask for irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Nash, K.E.; Gavin, M.E.

    1983-01-01

    A major step in the development of a large-scale transportation system for irradiated CANDU fuel is being made by Ontario Hydro in the design and construction of a demonstration cask by 1988/89. The system being designed is based on dry transportation with the eventual fully developed system providing for dry fuel loading and unloading. Research carried out to date has demonstrated that it is possible to transport irradiated CANDU fuel in a operationally efficient and simple manner without any damage which would prejudice subsequent automated fuel handling

  16. 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

  17. 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....

  18. 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)

    1999-04-01

    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

  19. The buckling of fuel rods in transportation casks under hypothetical accident conditions

    International Nuclear Information System (INIS)

    Bjorkman, G.S.

    2004-01-01

    The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations following a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding the higher the inertia loads on the cladding, and, therefore, the lower the ''g'' value at which buckling occurs. Current published solutions do not consider displacement compatibility between the fuel and the cladding. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading

  20. Mechanical properties used for the qualification of transport casks: Prototype development and extension to serial production

    International Nuclear Information System (INIS)

    Salzbrenner, R.; Crenshaw, T.B.; Sorenson, K.B.

    1991-01-01

    A thorough understanding of the mechanical behavior of material in a specific cask is required to properly analyze the structural response of the cask. An appropriate way to establish this understanding is through laboratory testing of cask material. The laboratory testing that was done to support the MOSAIK Drop Test Program is summarized as an example of how mechanical properties can be mapped for a prototype cask. The broad range behavior to be understood. This is necessary for the proper application of fracture mechanics, and focuses on fracture toughness as the inherent materials property which quantifies the fracture resistance of a material. The understanding established by a mechanics to a particular prototype, behavior of a prototype must be correctly associated with parameters which can be measured on production casks. Since the production casks cannot be destructively tested, measurements are commonly made on sub-size specimens. This may prevent direct measurement of valid design properties. An additional database may then be required to establish the correlation between sub-size specimen measurements and valid design properties. This is illustrated by outlining the additional testing which would be necessary to allow the successful verification of the MOSAIK Drop Test Program to be extended from the prototype to serially produced casks

  1. Verification of heat removal capability of a concrete cask system for spent fuel storage

    International Nuclear Information System (INIS)

    Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatugu

    2001-01-01

    The reprocessing works comprising of a center of nuclear fuel cycle in Japan is now under construction at Rokkasho-mura in Aomori prefecture, which is to be operated in 2005. However, as reprocessing capacity of the works is under total forming amount of spent nuclear fuels, it has been essential to construct a new facility intermediately to store them at a period before reprocessing them because of prediction to reach limit of pool storage in nuclear power stations. There are some intermediate storage methods, which are water pool method for wet storage, and bolt method, metal cask method, silo method and concrete cask method for dry storage. Among many methods, the dry storage is focussed at a standpoint of its operability and economy, the concrete cask method which has a lot of using results in U.S.A. has been focussed as a method expectable in its cost reduction effect among it. The Ishikawajima-Harima Heavy Industries Co., Ltd. produced, in trial, a concrete cask with real size to confirm productivity when advancing design work on concrete cask. By using the trial product, a heat removal test mainly focussing temperature of concrete in the cask was carried out to confirm heat conductive performances of the cask. And, analysis of heat conductivity was also carried out to verify validity of its analysis model. (G.K.)

  2. Analysis of radiation characteristics for casks loaded with spent RBMK-1500 nuclear fuel

    International Nuclear Information System (INIS)

    Smaizys, A.; Poskas, P.

    2001-01-01

    The objective of this paper is to present the analysis of radiation characteristics for the ductile cast iron CASTOR RBMK-1500 and heavy concrete CONSTOR RBMK-1500 casks loaded with spent nuclear fuel from Ignalina NPP RBMK-1500 reactors. These casks are designed for an interim storage (up to 50 years) of spent nuclear fuel at Ignalina NPP. Computer calculations have been performed using SCALE4.3 computer codes system. The dose rate calculations have been performed on the sidelong, upper and lower surface of the casks and for certain distance at the beginning of spent nuclear fuel storage in the casks and after 50 years of interim dry storage. The results obtained results show that dose rate values on the surface of the cask are much less than the design criteria value 1000 μSv/h when the average burn-up of fuel assembly is 20 GWd/tU. It was revealed that CONSTOR RBMK- 1500 cask has better shielding characteristics than CASTOR RBMK-1500 cask.(author)

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

    International Nuclear Information System (INIS)

    Fournel, B.; Turchet, J.P.; Faure, S.; Allinei, P.G.; Briquet, L.; Baubet, D.

    2002-01-01

    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)

    Thurmond, S.M.

    1995-08-01

    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 6O Co) were detected in the lead rejected for free release

  5. Numerical assessment of spent fuel casks impacting real targets

    Energy Technology Data Exchange (ETDEWEB)

    Ballheimer, V.; Probst, A.; Droste, B

    2000-07-01

    A reference container of high capacity was analysed for loads beyond those it has to withstand during a 9 m IAEA drop test onto an unyielding target. In doing this a lid-end drop with shock absorber onto a real target was simulated. This is a possible accident for the rail transport of such casks. In this case the most critical components of the containment system are the primary lid bolts. The behaviour of the lid system and its sealing function were investigated with finite element (FE) analysis. To correlate the findings with a corresponding impact velocity onto real targets an analytical method was used. Despite the conservative assumptions made in this study a two-fold safety factor compared to the 9 m drop tests onto the unyielding target could be shown. The quantification of the additional safety the cask might provide requires further basic investigations on the behaviour of the real targets considered as well as the reduction of the conservatism included in the assumptions made up to now. (author)

  6. Closure system for a spent fuel storage cask

    International Nuclear Information System (INIS)

    Kessinger, B.A.

    1987-01-01

    A closure system is described for temporarily sealing a cask base element having a mouth region with first and second steps and for permitting the cask base element to be permanently sealed, the closure system comprising: a primary cover having a bottom side with a first peripheral region configured for placement on the first step and having a top side with a second peripheral region, an upwardly open annular recess being provided in the primary cover at the secondary peripheral region; first means disposed between the first peripheral region and the first step for creating a mechanical seal to seal the cash base element at least temporarily; a generally hand-shaped element which is disposed in the annular recess and which has first and second edges and a curved cross section, the first edge being sealingly affixed to the primary cover and the second edge being sealingly affixed to the cash base element only if the cash base element is to be permanently sealed; a secondary cover having a bottom side with a peripheral region configured for placement on the second step; and second means disposed between the peripheral region of the secondary cover and the second step for creating an additional mechanical seal to seal the cast base element at least temporarily

  7. Transport and Storage Cask Safety Assessment - Drop Tests and Numerical Calculations -

    International Nuclear Information System (INIS)

    Voelzke, H.; Wille, F.; Wieser, G.; Quercetti, Th.

    2006-01-01

    BAM (the German Federal Institute for Materials Research and Testing) has been performing cask design testing for more than 30 years with a large number of prototype casks of original dimensions and of 1:2 or 1:3 scales. In 2004 a brand new drop test facility was built at the new BAM test facility at Horstwalde about 80 km to the south of Berlin. In September 2004 first demonstration tests with 2 different cask designs were performed in connection with the PATRAM 2004 conference held in Berlin. The dropped prototype casks had gross masses of 141 and 181 metric tons. Since that time BAM has been performing a lot of more drop tests with new cask designs developed by different international cask manufacturers for getting German Type B(U) transport licenses. Current safety assessments especially for mechanical accident scenarios require a combination of experimental and analytical/numerical proofs commonly, because both methods offer specific options and advantages with respect to more and more detailed structural analyses. That again is a consequence of the permanent cask design optimisation for commercial reasons leading to higher stress levels in general. For that reason BAM also improves its numerical analyses capacities including the operation of different software codes. A general BAM guideline describing basic requirements for numerical safety assessment reports gives a good orientation for both applicants and inspectors. But different details of any cask design and safety assessment have to be taken into account and lead to specific questions, investigations and experiences. This paper gives an overview about the new BAM drop test facility and the ongoing drop testing there and it presents current experiences and results of numerical cask analyses and the specific methods developed and used by BAM. In this context special attention is turned to the correlation between experimental and numerical results and an outlook to future developments is given. (authors)

  8. Mechanical properties used for the qualification of transport casks: Prototype development and extension to serial production

    International Nuclear Information System (INIS)

    Salzbrenner, R.; Crenshaw, T.B.; Sorenson, K.B.

    1992-01-01

    A thorough understanding of the mechanical behavior of material in a specific cask is required to properly analyze the structural response of the cask. An appropriate way to establish this understanding is through laboratory testing of cask material. The laboratory testing that was done to support the MOSAIK Drop Test Program is summarized as an example of how mechanical properties can be mapped for a prototype cask. The broad range of measured properties allows the critical aspects of mechanical behavior to be understood. This is necessary for the proper application of fracture mechanics, and focuses on fracture toughness as the inherent materials property which quantifies the fracture resistance of a material. The general fracture mechanics approach and its application to specific cask designs are described elsewhere (Salzbrenner et al. 1990, Sorenson et al. 1992a, Sorenson et al. 1992b). The understanding established by a thorough mapping of the mechanical properties is necessary to apply fracture mechanics to a particular prototype, but it is not sufficient for qualifying serially produced casks. The mechanical behavior of a prototype must be correctly associated with parameters which can be measured on production casks. Since the production casks cannot be destructively tested, measurements are commonly made on sub-size specimens. This may prevent direct measurement of valid design properties. An additional database may then be required to establish the correlation between sub-size specimen measurements and valid design properties. This is illustrated by outlining the additional testing which would be necessary to allow the successful verification of the MOSAIK Drop Test Program to be extended from the prototype to serially produced casks

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

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1991-08-01

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

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

    International Nuclear Information System (INIS)

    2007-01-01

    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. Development of on-site accident criteria for waste transfer casks

    International Nuclear Information System (INIS)

    Uldrich, E.D.

    1989-01-01

    Removal of radioactive waste must withstand the scrutiny of the public and various regulatory offices. Currently, there is no standard accident criteria or methodology for intra-site shipments at the Idaho National Engineering Laboratory (INEL). Since the radioactive waste transfer casks only carry material within the INEL site boundaries and are not used for normal over-the-road transport, the requirements of 10 CFR 71 Packaging and Transportation of Radioactive Material, do not provide suitable requirements for cask design or safety analyses. The objective is to develop realistically conservative accident scenarios consistent with the limited uses at the INEL for which the cask is approved

  12. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  13. ALARA for cask MRS by remote operations

    International Nuclear Information System (INIS)

    Wells, A.H.; Vick, D.E.

    1985-01-01

    Radiation dose rates in a monitored retrievable storage (MRS) facility are high enough to warrant the evaluation of robotic systems to achieve personnel dose reductions. Robots with sufficient mobility and dexterity to perform Health Physics surveys and maintenance are currently in use. The addition of artificial intelligence computer methods to the robot removes the need for a human operator for normal surveillance activities. Use of an Expert System creates a robot with sufficient flexibility to recognize and respond to off-normal conditions such as radiation leaks

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

    International Nuclear Information System (INIS)

    Leite da Silva, Luiz; Rodrigues Mansur, Tanius; Cimini Junior, Carlos Alberto

    2011-01-01

    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.

  15. Activation measurements for thermal neutrons. Part J. Evaluation of thermal neutron transmission factors

    International Nuclear Information System (INIS)

    Egbert, Stephen D.

    2005-01-01

    In order to relate thermal neutron activation measurements in samples to the calculated free-in-air thermal neutron activation levels given in Chapter 3, use is made of sample transmission factors. Transmission factors account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. The procedures for calculation of TF's and example results are presented in this section. (author)

  16. 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...

  17. Nondestructive Examination Guidance for Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suter, Jonathan D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lareau, John P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhuge, Jing Wei [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Moran, Traci L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-30

    In this report, an assessment of NDE methods is performed for components of NUHOMS 80 and 102 dry storage system components in an effort to assist NRC staff with review of license renewal applications. The report considers concrete components associated with the horizontal storage modules (HSMs) as well as metal components in the HSMs. In addition, the report considers the dry shielded canister (DSC). Scope is limited to NDE methods that are considered most likely to be proposed by licensees. The document, ACI 349.3R, Evaluation of Existing Nuclear Safety-Related Concrete Structures, is used as the basis for the majority of the NDE methods summarized for inspecting HSM concrete components. Two other documents, ACI 228.2R, Nondestructive Test Methods for Evaluation of Concrete in Structures, and ORNL/TM-2007/191, Inspection of Nuclear Power Plant Structure--Overview of Methods and Related Application, supplement the list with additional technologies that are considered applicable. For the canister, the ASME B&PV Code is used as the basis for NDE methods considered, along with currently funded efforts through industry (Electric Power Research Institute [EPRI]) and the U.S. Department of Energy (DOE) to develop inspection technologies for canisters. The report provides a description of HSM and DSC components with a focus on those aspects of design considered relevant to inspection. This is followed by a brief description of other concrete structural components such as bridge decks, dams, and reactor containment structures in an effort to facilitate comparison between these structures and HSM concrete components and infer which NDE methods may work best for certain HSM concrete components based on experience with these other structures. Brief overviews of the NDE methods are provided with a focus on issues and influencing factors that may impact implementation or performance. An analysis is performed to determine which NDE methods are most applicable to specific

  18. Performance evaluation of a thermal Doppler Michelson interferometer system.

    Science.gov (United States)

    Mani, Reza; Dobbie, Steven; Scott, Alan; Shepherd, Gordon; Gault, William; Brown, Stephen

    2005-11-20

    The thermal Doppler Michelson interferometer is the primary element of a proposed limb-viewing satellite instrument called SWIFT (Stratospheric Wind Interferometer for Transport studies). SWIFT is intended to measure stratospheric wind velocities in the altitude range of 15-45 km. SWIFT also uses narrowband tandem etalon filters made of germanium to select a line out of the thermal spectrum. The instrument uses the same technique of phase-stepping interferometry employed by the Wind Imaging Interferometer onboard the Upper Atmosphere Research Satellite. A thermal emission line of ozone near 9 microm is used to detect the Doppler shift due to winds. A test bed was set up for this instrument that included the Michelson interferometer and the etalon filters. For the test bed work, we investigate the behavior of individual components and their combination and report the results.

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

    International Nuclear Information System (INIS)

    Wattez, L.; Marguerat, Y.; Hoesli, C.

    2004-01-01

    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

  20. CFD simulation of a cabin thermal environment with and without human body - thermal comfort evaluation

    Science.gov (United States)

    Danca, Paul; Bode, Florin; Nastase, Ilinca; Meslem, Amina

    2018-02-01

    Nowadays, thermal comfort became one of the criteria in choosing a vehicle. In last decades time spent by people in vehicles had risen substantially. During each trip, thermal comfort must to be ensured for a good psychological and physical state of the passengers. Also, a comfortable environment leads to a higher power concentration of the driver thereby to a safe trip for vehicle occupants and for all traffic participants. The present study numerically investigated the effect of human body sited in the driver's place, over the air velocity distribution and over the thermal comfort in a passenger compartment. CFD simulations were made with different angles of the left inlet grill, in both cases, with and without driver presence. In majority of the actual vehicles environment studies, are made without consideration of human body geometry, in this case, the results precision can be affected. The results show that the presence of human body, lead to global changing of the whole flow pattern inside the vehicular cabin. Also, the locations of the maximum velocities are changing with the angle of the guiding vanes. The thermal comfort PMV/PPD indexes were calculated for each case. The presence of human body leads to a more comfortable environment.

  1. Thermal Evaluation of Scorched Graphite-Epoxy Panels by Infrared Scanning

    OpenAIRE

    Slifka, A. J.; Hall, T.; Boltz, E. S.

    2003-01-01

    A simple measurement system is described for evaluating damage to graphite-epoxy panels, such as those used in high-performance aircraft. The system uses a heating laser and infrared imaging system to measure thermal performance. Thermal conductivity or diffusivity is a sensitive indicator of damage in materials, allowing this thermal measurement to show various degrees of damage in graphite-epoxy composites. Our measurements track well with heat-flux damage to graphite epoxy panels. This mea...

  2. 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...

  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 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...... we applied different thermodynamic models, such as the Soave-Redlich-Kwong and the Peng-Robinson equations of state. The necessity to try different thermo-dynamic models is caused by the high sensitivity of the thermal diffusion factors to the values of the partial molar properties. Two different...

  5. Computer simulations of a generic truck cask in a regulatory fire using the Container Analysis Fire Environment (CAFE) code

    International Nuclear Information System (INIS)

    Ju, H.; Greiner, M.; Suo-Anttila, A.

    2002-01-01

    The Container Analysis Fire Environment (CAFE) computer code is designed to predict accurately convection and radiation heat transfer to a thermally massive object engulfed in a large pool fire. It is well suited for design and risk analyses of spent nuclear fuel transport systems. CAFE employs computational fluid dynamics and several fire and radiation models. These models must be benchmarked using experimental results. In this paper, a set of wind velocity conditions are determined which allow CAFE accurately to reproduce recent heat transfer measurements for a thick walled calorimeter in a ST-1 regulatory pool fire. CAFE is then used to predict the response of an intack (thin walled) generic legal weight truck cask. The maximum temperatures reached by internal components are within safe limits. A simple 800 deg. C, grey-radiation fire model gives maximum component temperatures that are somewhat below those predicted by CAFE. (author)

  6. Evaluation of a fuzzy logic controller for laser thermal therapy

    Science.gov (United States)

    Choy, Vanessa; Sadeghian, Alireza; Sherar, Michael D.; Whelan, William M.

    2002-06-01

    Laser thermal therapy (LTT) is a minimally invasive surgical technique used to destroy solid tumors while minimizing damage to adjacent normal tissues. Optical energy, delivered through fibers implanted into the target volume, raises tissue temperatures above 60 degree(s)C resulting in coagulative necrosis (thermal damage). Thermal damage volumes, however, can be irregular and unpredictable, resulting from dynamic changes in the tissue properties during treatment. A closed-loop feedback fuzzy logic controller for LTT was developed with the tissue treated as a black-box system. Preliminary testing was conducted for simulated LTT with a single spherically emitting source fiber at the center of 5 mm and 10 mm diameter target tissues. Dynamic changes in blood perfusion and tissue optical properties due to heating were incorporated into the LTT simulator. Input laser power was modulated to control the temperature field in an attempt to reach target temperatures at the source (90 degree(s)C to avoid tissue charring) and at the target boundary (55 degree(s)C). In all simulations, thermal damage based on Arrhenius formulation ((Omega) equals 1) was reached at the target boundary. The controller also responded efficiently to unexpected, rapid temperature changes.

  7. Synthesis and evaluation of effective parameters in thermal ...

    Indian Academy of Sciences (India)

    In solid oxide fuel cells (SOFCs) the cathode functions as the site for thee lectrochemical reduction of oxygen. There- fore, thec athode must have high electronic conductivity, ade- quate porosity, stability under an oxidizing atmosphere and high catalytic activity. In addition, it is crucial for the cathode to have matched thermal ...

  8. Evaluation of Electrical and Thermal Conductivity of Polymeric ...

    African Journals Online (AJOL)

    PROF HORSFALL

    ABSTRACT: This work being gingered by the big menace being posed on our environment by polymeric waste and it's rechanneling involved the studying of the electrical and thermal conductivities of the polymers PP, PE, PS and nylon66 doped with charcoal and graphite. Five grams of each polymer was mixed with ...

  9. Evaluation of electrical and thermal conductivity of polymeric wastes ...

    African Journals Online (AJOL)

    This work being gingered by the big menace being posed on our environment by polymeric waste and it's rechanneling involved the studying of the electrical and thermal conductivities of the polymers PP, PE, PS and nylon66 doped with charcoal and graphite. Five grams of each polymer was mixed with varying ...

  10. Thermal assault and polyurethane foam-evaluating protective mechanisms

    International Nuclear Information System (INIS)

    Williamson, C.L.; Iams, Z.L.

    2004-01-01

    Rigid polyurethane foam utilizes a variety of mechanisms to mitigate the thermal assault of a ''regulatory burn''. Polymer specific heat and foam k-factor are of limited usefulness in predicting payload protection. Properly formulated rigid polyurethane foam provides additional safeguards by employing ablative mechanisms which are effective even when the foam has been crushed or fractured as a result of trauma. The dissociative transitions from polymer to gas and char, and the gas transport of heat from inside the package out into the environment are also thermal mitigators. Additionally, the in-situ production of an intumescent, insulative, carbonaceous char, confers thermal protection even when a package's outer steel skin has been breached. In this test program, 19 liter, ''Five gallon'' steel pails are exposed on one end to the flame of an ''Oil Burner'' as described in the US Federal Aviation Administration (FAA) ''Aircraft Materials Fire Test Handbook''. When burning 2 diesel at a nominal rate of 8.39 kg (18.5 pounds)/hr, the burner generates a high emissivity flame that impinges on the pail face with the thermal intensity of a full scale pool-fire environment. Results of these tests, TGA and MDSC analysis on the subject foams are reported, and their relevance to full size packages and pool fires are discussed

  11. Evaluation of Electrical and Thermal Conductivity of Polymeric ...

    African Journals Online (AJOL)

    PROF HORSFALL

    the best available conductor, that is, silver. At room temperature, the conductivity of polyacetylene ... cell batteries. Electrical weighing balance from mettler. Toledo 2007 mode serial no 021-64852350. ENGLAND, Stirring rods, Beakers (PYREX). Electro thermal Heater of about 250oc from Barnstaed 2006 model, serial no ...

  12. Methods of evaluation of thermal tolerance of cyclic sports athletes

    Directory of Open Access Journals (Sweden)

    Kish А.А.

    2017-12-01

    Full Text Available The purpose of the study is investigation of thermal stability in athletes of cyclic sports and assessment of its influence on physical working capacity under hyperthermia. Material and methods. 15 male athletes of cyclic sports who had the senior degree as minimum were included in the study Middle age 24,2±1,1 years. Work was performed in the climatic camera of the Center of sports medicine and rehabilitation in several stages, on each of which assessment of physical working capacity and a thermal condition of athletes was carried out. Results. In the real work the burdening action of the heating climate on indicators of physical working capacity and a thermal condition of athletes of cyclic sports is shown; the operating ranges of high temperatures are determined. Conclusion. The data obtained by means of the offered technique, confirm importance of definition of individual thresholds of shipping of a thermal state at athletes and the burdening action of a heat on their physical working capacity.

  13. Evaluating Thermal Properties of Rock | Bello | Journal of the ...

    African Journals Online (AJOL)

    The coefficients of the generalized model give the thermal properties of each rock type. The chi-square test showed that there was no significant difference (p>0.05) between the expected and observed data for all the models. The model developed in this work enabled us to use simulation prediction as the basis for rock ...

  14. Comparative Batch and Column Evaluation of Thermal and Wet ...

    African Journals Online (AJOL)

    batch adsorption experiments and continous flow (fixed bed) column experiment to study the mechanism of dye removal by the commercial ... compared with those obtained for thermal regeneration method (qmax = 61.73mgg-1) and wet oxidative method (qmax .... Carbon hardness was determined using a wet attrition test ...

  15. Evaluation and verification of thermal stratification models for was

    African Journals Online (AJOL)

    USER

    form bacteria and verification. INTRODUCTION ... bottom. The thermal stratification can be stable-persisting for months or intermittent, appearing for a few hours in the day (Dor et al. 1993; Pedahzur et al., 1993; Torres et al.,. 1997). In waste stabilization ponds the .... water depth of 0.2 and a thick sludge deposits. The WSP ...

  16. The application of fracture mechanics to the safety assessment of transport casks for radioactive material

    International Nuclear Information System (INIS)

    Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H.

    2004-01-01

    BAM is the responsible authority in Germany for the assessment of the mechanical and thermal design safety of packages for the transport of radioactive materials. The assessment has to cover the brittle fracture safety 'proof of package' for 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 will be permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof using the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA's 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 conclude the paper. (author)

  17. Thermal Margin Evaluation Report for COLSS of APR 1400 using the Virtual Incore Detector

    Energy Technology Data Exchange (ETDEWEB)

    Lee, E.K.; Cha, K.H. [Korea Electric Power Research Institute, Taejon (Korea)

    2001-07-01

    The objective of this analysis is to evaluate thermal margin for COLSS, using the virtual incore detector, of Advanced Power Reactor 1400 according to the {sup R}esearch and Development on Next Generation Reactor (Phase III) - Core/Thermal Hydraulics Design{sup .} (author) 7 refs., 18 figs., 7 tabs.

  18. Cordierite production with natural and synthetic raw materials, and evaluation of resistance to thermal shock

    International Nuclear Information System (INIS)

    Buoso, Alberto; Bergmann, Carlos Perez

    1997-01-01

    This work presents a study on the formation of cordierite from raw materials and pure oxides. For this proposal, different formulations and sintering curves were developed. The formation of cordierite was analysed by means of both linear thermal expansion and X-ray diffraction. The performance of these materials under thermal shock was also evaluated. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Koji Shirai

    2006-04-01

    The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

  20. Development of NUPAC 140B 100 ton rail/barge cask

    International Nuclear Information System (INIS)

    1990-04-01

    The NuPac 140-B 100 Ton Rail/Barge Shipping Cask Preliminary Design Report (PDR) presents a general introduction to, and description of, the NuPac 140-B Cask and its fuel payload. The NuPac 140-B Cask, Model: NuPac 140-B, is being designed by Nuclear Packaging, Inc., to meet or exceed all NRC and Department of Transportation regulations governing the shipment of radioactive material. Specifically the Cask is being developed as a safe means of transporting spent light-water-reactor (LWR) fuels from existing and proposed reactor facilities to a repository and/or a monitored retrievable storage (MRS) facility. The primary transportation mode is by railroad, although the shipping package is designed to be transported by barge and by truck shipment on a special overweight basis for short distances. This feature allows the servicing of reactor sites and other facilities which lack direct railroad access

  1. 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.

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

    International Nuclear Information System (INIS)

    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

  3. GA-4 and GA-9 legal weight truck shipping cask development

    International Nuclear Information System (INIS)

    Grenier, R.M.; Meyer, R.J.; Jensen, M.F.

    1989-02-01

    We are developing two new legal weight truck spent fuel shipping casks that will carry four PWR or nine BWR spent fuel assemblies. They are being developed to meet requirements to dispose of nuclear wastes at a permanent disposal site. Our primary goal is to maximize the number of fuel elements of each fuel type that a legal weight truck (LWT) cask can carry, while ensuring that the design meets all NRC licensing requirements. 1 ref., 4 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Daling, P.M.

    1984-11-01

    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.

  5. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    International Nuclear Information System (INIS)

    Tang, J.S.

    1980-11-01

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm

  6. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations

    International Nuclear Information System (INIS)

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

    1991-12-01

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 x 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990

  7. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations; Yucca Mountain Site Characterization Project

    Energy Technology Data Exchange (ETDEWEB)

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States)

    1991-12-01

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 {times} 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990.

  8. Developments in shielding and criticality assessment for cask design

    International Nuclear Information System (INIS)

    Watmough, M.H.; Cooper, A.J.; Croxford, R.W.

    1993-01-01

    This paper presents recent highlights from the shielding and criticality methods development programme that are of relevance to cask design. Specifically, the following points emerge: 1) the preparation of a licence application based upon UK methods and data used in a standardized fissile depletion and plutonium production model has been completed; 2) the assumptions used in the modelling of granules of broken fuel within the transport package following a postulated impact accident have been revised thereby allowing less pessimistic assessments to be performed; 3) enhancements are being made to the software used for shielding and criticality analysis enabling a more cost effective design service to be provided. These ongoing developments clearly show the activity to extend the scope of assessments while increasing the physical realism of the models. Through these developments BNFL continues to offer a comprehensive and cost effective shielding and criticality analysis service as part of its worldwide fuel transport business. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Hill, C.V.; Loud, G.C.; Heitzman, A.C.

    1993-01-01

    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

  10. COBRA-SFS modifications and cask model optimization

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. Optical Property Evaluation of Next Generation Thermal Control Coatings

    Science.gov (United States)

    Jaworske, Donald A.; Deshpande, Mukund S.; Pierson, Edward A.

    2010-01-01

    Next generation white thermal control coatings were developed via the Small Business Innovative Research program utilizing lithium silicate chemistry as a binder. Doping of the binder with additives yielded a powder that was plasma spray capable and that could be applied to light weight polymers and carbon-carbon composite surfaces. The plasma sprayed coating had acceptable beginning-of-life and end-of-live optical properties, as indicated by a successful 1.5 year exposure to the space environment in low Earth orbit. Recent studies also showed the coating to be durable to simulated space environments consisting of 1 keV and 10 keV electrons, 4.5 MeV electrons, and thermal cycling. Large scale deposition was demonstrated on a polymer matrix composite radiator panel, leading to the selection of the coating for use on the Gravity Recovery And Interior Laboratory (GRAIL) mission.

  13. 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.

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

    International Nuclear Information System (INIS)

    Alekseev, O.G.; Matveev, V.Z.; Morenko, A.I.; Il'kaev, R.I.; Shapovalov, V.I.

    2004-01-01

    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

  15. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    Science.gov (United States)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-07-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  16. Synthesis and evaluation of effective parameters in thermal ...

    Indian Academy of Sciences (India)

    Abstract. In this study, oxides Ln0.6Sr0.4Co0.2M0.8O3−δ (Ln = La,Nd and M = Mn,Fe) have been prepared and characterized to study the influence of the different cations on thermal expansion coefficient (TEC). TEC decreases favourably with replacement of Nd3+ and Mn3+ ions in the lattice. Nd3+ leads to decreasing ...

  17. Three-dimensional finite element impact analysis of a nuclear waste truck cask

    International Nuclear Information System (INIS)

    Miller, J.D.

    1985-01-01

    This paper presents a three-dimensional finite element impact analysis of a hypothetical accident event for the preliminary design of a shipping cask which is used to transport radioactive waste by standard tractor-semitrailer truck. The nonlinear dynamic structural analysis code DYNA3D run on Sandia's Cray-1 computer was used to calculate the effects of the cask's closure-end impacting a rigid frictionless surface on an edge of its external impact limiter after a 30-foot fall. The center of gravity of the cask (made of 304 stainless steel and depleted uranium) was assumed to be directly above the impact point. An elastic-plastic material constitutive model was used to calculate the nonlinear response of the cask components to the transient loading. Interactive color graphics (PATRAN and MOVIE BYU) were used throughout the analysis, proving to be extremely helpful for generation and verification of the geometry and boundary conditions of the finite element model and for interpretation of the analysis results. Results from the calculations show the cask sustained large localized deformations. However, these were almost entirely confined to the impact limiters built into the cask. The closure sections were determined to remain intact, and leakage would not be expected after the event. As an example of a large three-dimensional finite element dynamic impact calculation, this analysis can serve as an excellent benchmark for computer aided design procedures

  18. Evaluation of the Efficiency of Liquid Cooling Garments using a Thermal Manikin

    National Research Council Canada - National Science Library

    Xu, Xiaojiang; Endrusick, Thomas; Gonzalez, Julio; Laprise, Brad; Teal, Walter; Santee, William; Kolka, Margaret

    2005-01-01

    .... personal protective equipment), and environmental conditions. Thermal manikins (TM) have been used to evaluate the performance of LCG systems and to determine the amount of heat that a LCG can extract from a TM...

  19. Flaw evaluation of thermally aged cast stainless steel in light-water reactor applications

    International Nuclear Information System (INIS)

    Lee, S.; Kuo, P.T.; Wichman, K.; Chopra, O.

    1997-01-01

    Cast stainless steel may be used in the fabrication of the primary loop piping, fittings, valve bodies, and pump casings in light-water reactors. However, this material is subject to embrittlement due to thermal aging at the reactor temperature, that is 290 o C (550 o F). The Argonne National Laboratory (ANL) recently completed a research program and the results indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). Thus, the US Nuclear Regulatory Commission (NRC) staff has accepted the use of SAW flaw evaluation procedures in IWB-3640 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to evaluate flaws in thermally aged cast stainless steel for a license renewal evaluation. Alternatively, utilities may estimate component-specific fracture toughness of thermally aged cast stainless steel using procedures developed at ANL for a case-by-case flaw evaluation. (Author)

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

    International Nuclear Information System (INIS)

    Abadie, P.

    2004-01-01

    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 22 at/cm 3 for hydrogen and 9.10 20 at/cm 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

  1. Thermoregulatory strategies in an aquatic ectotherm from thermally-constrained habitats: An evaluation of current approaches.

    Science.gov (United States)

    Piasečná, Karin; Pončová, Alena; Tejedo, Miguel; Gvoždík, Lumír

    2015-08-01

    Many ectotherms employ diverse behavioral adjustments to effectively buffer the spatio-temporal variation in environmental temperatures, whereas others remain passive to thermal heterogeneity. Thermoregulatory studies are frequently performed on species living in thermally benign habitats, which complicate understanding of the thermoregulation-thermoconformity continuum. The need for new empirical data from ectotherms exposed to thermally challenging conditions requires the evaluation of available methods for quantifying thermoregulatory strategies. We evaluated the applicability of various thermoregulatory indices using fire salamander larvae, Salamandra salamandra, in two aquatic habitats, a forest pool and well, as examples of disparate thermally-constrained environments. Water temperatures in the well were lower and less variable than in the pool. Thermal conditions prevented larvae from reaching their preferred body temperature range in both water bodies. In contrast to their thermoregulatory abilities examined in a laboratory thermal gradient, field body temperatures only matched the mean and range of operative temperatures, showing thermal passivity of larvae in both habitats. Despite apparent thermoconformity, thermoregulatory indices indicated various strategies from active thermoregulation, to thermoconformity, and even thermal evasion, which revealed their limited applicability under thermally-constrained conditions. Salamander larvae abandoned behavioral thermoregulation despite varying opportunities to increase their body temperature above average water temperatures. Thermoconformity represents a favored strategy in these ectotherms living in more thermally-constrained environments than those examined in previous thermoregulatory studies. To understand thermal ecology and its impact on population dynamics, the quantification of thermoregulatory strategies of ectotherms in thermally-constrained habitats requires the careful choice of an appropriate

  2. CASTOR(r) and CONSTOR(r) type transport and storage casks for spent fuel and high active waste

    International Nuclear Information System (INIS)

    Kuehne, B.; Sowa, W.

    2002-01-01

    The German company GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and high-level waste. CASTOR(r) casks are used at 18 sites on three continents. Spent fuel assemblies of the types PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste (HAW) containers are stored in these kinds of casks. More than 600 CASTOR(r) casks have been loaded for long-term storage. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in each case. There is no indication that problems will arise in the future. Of course, the experience of 20 years has resulted in improvements of the cask design. One basic improvement is GNB's development since the mid 1990s of a sandwich cask design using heavy concrete and steel as basic materials, for economical and technical reasons. This CONSTOR(r) cask concept also fulfils all design criteria for transport and storage given by the IAEA recommendations and national authorities. By May 2002 40 CONSTOR(r) casks had been delivered and 15 had been successfully loaded and stored. In this paper the different types of casks are presented. Experiences gained during the large number of cask loadings and more than 4000 cask-years of storage will be summarised. The presentation of recent and future development shows the optimisation potential of the CASTOR(r) and CONSTOR(r) cask families for safe and economical management of spent fuel. (author)

  3. Thermally-Induced Crack Evaluation in H13 Tool Steel

    Directory of Open Access Journals (Sweden)

    Hassan Abdulrssoul Abdulhadi

    2017-11-01

    Full Text Available This study reported the effect of thermal wear on cylindrical tool steel (AISI H13 under aluminum die-casting conditions. The AISIH13 steels were immersed in the molten aluminum alloy at 700 °C before water-quenching at room temperature. The process involved an alternating heating and cooling of each sample for a period of 24 s. The design of the immersion test apparatus stylistically simulated aluminum alloy dies casting conditions. The testing phase was performed at 1850, 3000, and 5000 cycles. The samples were subjected to visual inspection after each phase of testing, before being examined for metallographic studies, surface crack measurement, and hardness characteristics. Furthermore, the samples were segmented and examined under optical and Scanning Electron Microscopy (SEM. The areas around the crack zones were additionally examined under Energy Dispersive X-ray Spectroscopy (EDXS. The crack’s maximum length and Vickers hardness profiles were obtained; and from the metallographic study, an increase in the number of cycles during the testing phase resulted in an increase in the surface crack formation; suggesting an increase in the thermal stress at higher cycle numbers. The crack length of Region I (spherically shaped was about 47 to 127 µm, with a high oxygen content that was analyzed within 140 µm from the surface of the sample. At 700 °C, there is a formation of aluminum oxides, which was in contact with the surface of the H13 sample. These stresses propagate the thermal wear crack length into the tool material of spherically shaped Region I and cylindrically shape Region II, while hardness parameters presented a different observation. The crack length of Region I was about 32% higher than the crack length of Region II.

  4. TMI-2 transportation program - Design considerations for the NUPAC 125-B cask handling and loading/unloading equipment

    International Nuclear Information System (INIS)

    Schmoker, D.S.; Schmitt, R.C.; Barkanic, R.J.

    1987-01-01

    Removal, transport and receipt of core debris from the damaged reactor at Three Mile Island Unit 2 (TMI-2) required the design of transport cask handling and ''dry'' loading and unloading equipment. The system for ''dry'' (not underwater) loading of the transport cask includes; 1) equipment for handling the cask, 2) equipment for loading core debris canisters into the cask at TMI, and 3) equipment for removing the canisters in a hot cell facility. This paper reviews the technical design considerations operational parameters and summarizes lessons learned in the design, testing, startup and use of the equipment provided for the TMI-2 Transportation Program

  5. Investigation on structural analysis computer program of spent nuclear fuel shipping cask, (2)

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results (II) done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for drop impact of the axisymmetric cylinder and plate were adopted the calculational evaluations for taking into account the strain rate effect of material properties, effect of artificial viscosity and effect of time integration step size were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  6. Investigation on structural analysis computer program of spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods such as HONDO, STEALTH and DYNA-3D, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for 9 m vertical drop impact of the axisymmetric lead cylinder with and without stainless steel clads were adopted where the calculational evaluations for taking into account the strain rate effect were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  7. Numerical simulation of thermal behaviors of a clothed human body with evaluation of indoor solar radiation

    International Nuclear Information System (INIS)

    Mao, Aihua; Luo, Jie; Li, Yi

    2017-01-01

    Highlights: • Solar radiation evaluation is integrated with the thermal transfer in clothed humans. • Thermal models are developed for clothed humans exposed in indoor solar radiation. • The effect of indoor solar radiation on humans can be predicted in different situations in living. • The green solar energy can be efficiently utilized in the building development. - Abstract: Solar radiation is a valuable green energy, which is important in achieving a successful building design for thermal comfort in indoor environment. This paper considers solar radiation indoors into the transient thermal transfer models of a clothed human body and offers a new numerical method to analyze the dynamic thermal status of a clothed human body under different solar radiation incidences. The evaluation model of solar radiation indoors and a group of coupled thermal models of the clothed human body are developed and integrated. The simulation capacities of these integrated models are validated through a comparison between the predicted results and the experimental data in reference. After that, simulation cases are also conducted to show the influence of solar radiation on the thermal status of individual clothed body segments when the human body is staying indoors in different seasons. This numerical simulation method provides a useful tool to analyze the thermal status of clothed human body under different solar radiation incidences indoors and thus enables the architect to efficiently utilize the green solar energy in building development.

  8. 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.

  9. Research on technology of evaluating thermal property data of nuclear power materials

    International Nuclear Information System (INIS)

    Imai, Hidetaka; Baba, Tetsuya; Matsumoto, Tsuyoshi; Kishimoto, Isao; Taketoshi, Naoyuki; Arai, Teruo

    1997-01-01

    For the materials of first wall and diverter of nuclear fusion reactor, in order to withstand steady and unsteady high heat flux load, excellent thermal characteristics are required. It is strongly demanded to measure such thermal property values as heat conductivity, heat diffusivity, specific heat capacity, emissivity and so using small test pieces up to higher than 2000degC. As the materials of nuclear reactors are subjected to neutron irradiation, in order to secure the long term reliability of the materials, it is very important to establish the techniques for forecasting the change of the thermal property values due to irradiation effect. Also the establishment of the techniques for estimating the thermal property values of new materials like low radioactivation material is important. In National Research Laboratory of Metrology, the research on the advancement of the measuring technology for high temperature thermal properties has resulted in the considerably successful development of such technologies. In this research, the rapid measurement of thermal property values up to superhigh temperature with highest accuracy, the making of thermal property data set of high level, the analysis and evaluation of the correlation of material characters and thermal property values, and the development of the basic techniques for estimating the thermal property values of solid materials are aimed at and advanced. These are explained. (K.I.)

  10. Thermal stress evaluation of the Viking RTG heat shield

    International Nuclear Information System (INIS)

    Stadter, J.T.; Weiss, R.O.

    1976-03-01

    Thermal stress analyses of the Viking RTG heat shield are presented. The primary purpose of the analyses was to determine the effects of the end cap and the finite length of the heat shield on the peak tensile stress in the barrel wall. The SAAS III computer code was used to calculate the thermal stresses; axisymmetric and plane section analyses were performed for a variety of temperature distributions. The study consisted of three parts. In the first phase, the influence of the end cap on the barrel wall stresses was examined by parametrically varying the modulus of elasticity of the contact zone between the end cap and the barrel. The second phase was concerned with stresses occurring as a result of an orbital decay reentry trajectory, and the effects of the magnitude and shape of the axial temperature gradient. The final part of the study was concerned with the circumferentially nonuniform temperature distribution which develops during a side-on stable reentry. The last part includes a comparison of stresses generated for a hexagonal cross section with those generated for a circular cross section

  11. 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.

  12. Structural changes of wood during hydro-thermal and thermal treatments evaluated through NIR spectroscopy and principal component analysis

    Science.gov (United States)

    Popescu, Carmen-Mihaela; Navi, Parviz; Placencia Peña, María Inés; Popescu, Maria-Cristina

    2018-02-01

    Spruce wood samples were subjected to different conditions of thermal and hydro-thermal treatment by varying the temperature, relative humidity and period of exposure. The obtained treated samples were evaluated using near infrared spectroscopy (NIR), principal component analysis (PCA) and hierarchical cluster analysis (HCA) in order to evidence the structural changes which may occur during the applied treatment conditions. Following this, modification in all wood components were observed, modifications which were dependent on the temperature, amount of relative humidity and also the treatment time. Therefore, higher variations were evidenced for samples treated at higher temperatures and for longer periods. At the same time, the increase in the amount of water vapours in the medium induced a reduced rate of side chains and condensation reactions occurring in the wood structure. Further, by PCA and HCA was possible to discriminate the modifications in the wood samples according to treatment time and amount of relative humidity.

  13. Performance evaluation of cost-optimized thermal cycler.

    Science.gov (United States)

    Park, Chan-Young; Park, Young-Hyun; Kim, Yu-Seop; Song, Hye-Jeong; Kim, Jong-Dae

    2015-01-01

    A polymerase chain reaction is a test method currently used in almost all process steps of a genetic manipulation experiment. It involves the amplification of the given genetic material targeted by the detection test. In consideration of the graphical user interface development environment or user accessibility, if a PC with the windows operating system or its embedded version can be employed as a host, it will contribute significantly to resource saving, including development-related human resources and time, along with enabling a broad use of the product. In this study, we focus on the low cost implementation of a PCR thermal cycler for the personal usage. It is aimed to drastically reduce the product development time and maintenance/repair costs. To achieve this, we implement the functions for biochemical process in a local embedded system, and the functions of data management, including the PCR protocol, and user-interface management are implemented on a PC.

  14. Recommendations for cask features for robotic handling from the Advanced Handling Technology Project

    International Nuclear Information System (INIS)

    Drotning, W.

    1991-02-01

    This report describes the current status and recent progress in the Advanced Handling Technology Project (AHTP) initiated to explore the use of advanced robotic systems and handling technologies to perform automated cask handling operations at radioactive waste handling facilities, and to provide guidance to cask designers on the impact of robotic handling on cask design. Current AHTP tasks have developed system mock-ups to investigate robotic manipulation of impact limiters and cask tiedowns. In addition, cask uprighting and transport, using computer control of a bridge crane and robot, were performed to demonstrate the high speed cask transport operation possible under computer control. All of the current AHTP tasks involving manipulation of impact limiters and tiedowns require robotic operations using a torque wrench. To perform these operations, a pneumatic torque wrench and control system were integrated into the tool suite and control architecture of the gantry robot. The use of captured fasteners is briefly discussed as an area where alternative cask design preferences have resulted from the influence of guidance for robotic handling vs traditional operations experience. Specific robotic handling experiences with these system mock-ups highlight a number of continually recurring design principles: (1) robotic handling feasibility is improved by mechanical designs which emphasize operation with limited dexterity in constrained workspaces; (2) clearances, tolerances, and chamfers must allow for operations under actual conditions with consideration for misalignment and imprecise fixturing; (3) successful robotic handling is enhanced by including design detail in representations for model-based control; (4) robotic handling and overall quality assurance are improved by designs which eliminate the use of loose, disassembled parts. 8 refs., 15 figs

  15. 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.

  16. Thermal Protection Systems Nondestructive Evaluation Tool, Phase I

    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)...

  17. Signature Evaluation for Thermal Infrared Countermine and IED Detection Systems

    National Research Council Canada - National Science Library

    Peters, John F; Howington, Stacy E; Ballard, Jr., Jerry; Lynch, Larry N

    2007-01-01

    .... These simulations provide information ranging from simple temperature contrasts to high-resolution images comparable to actual sensor images that can be used to evaluate or train automatic target recognition (ATR) systems...

  18. Comparative study of linear and nonlinear ultrasonic techniques for evaluation thermal damage of tube like structures

    International Nuclear Information System (INIS)

    Li, Weibin; Cho, Younho; Li, Xianqiang

    2013-01-01

    Ultrasonic guided wave techniques have been widely used for long range nondestructive detection in tube like structures. The present paper investigates the ultrasonic linear and nonlinear parameters for evaluating the thermal damage in aluminum pipe. Specimens were subjected to thermal loading. Flexible polyvinylidene fluoride (PVDF) comb transducers were used to generate and receive the ultrasonic waves. The second harmonic wave generation technique was used to check the material nonlinearity change after different heat loadings. The conventional linear ultrasonic approach based on attenuation was also used to evaluate the thermal damages in specimens. The results show that the proposed experimental setup is viable to assess the thermal damage in an aluminum pipe. The ultrasonic nonlinear parameter is a promising candidate for the prediction of micro damages in a tube like structure

  19. Evaluation of Strategies to Improve the Thermal Performance of Steel Frames in Curtain Wall Systems

    Directory of Open Access Journals (Sweden)

    Ji Hyun Oh

    2016-12-01

    Full Text Available Recently, metal curtain wall systems have been widely used in high-rise buildings due to many advantages, including being lightweight, rapid construction, and aesthetic features. Since the metal frame may lead to lower energy performance, thermal discomfort, and condensation risk due to the high thermal conductivity, its thermal performance can be important for the improvement of the overall thermal performance of the curtain wall system, as well as the energy efficiency of the building envelope. This study aims to evaluate variety of design strategies to improve the thermal performance of steel curtain wall frames. Five base cases and three further steps were selected for two different head profile shapes based on a state-of-the art technology review, and their thermal transmittances were calculated through simulations according to the ISO 12631 standard which is an international standard for calculating thermal transmittance of curtain wall system. Measured results that were obtained from hot-box tests were compared with the calculated results to validate the simulation method of this study. The shape of the head profile did not strongly influence the overall thermal transmittance, and the choice of strategies for the rabbet space was more important. More effective strategies could be decided according to the steps for variation development. This result can serve as a guideline for the design of high-performance curtain wall frames.

  20. 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.

  1. A model for managing and evaluating solar radiation for indoor thermal comfort

    Energy Technology Data Exchange (ETDEWEB)

    La Gennusa, Maria; Rizzo, Gianfranco [Dipartimento di Ricerche Energetiche ed Ambientali (DREAM), Universita degli Studi di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Nucara, Antonino; Pietrafesa, Matilde [Dipartimento di Informatica, Matematica, Elettronica e Trasporti (DIMET), Universita Mediterranea di Reggio Calabria, Feo di Vito, 89060 Reggio Calabria (Italy)

    2007-05-15

    Thermal comfort of people occupying indoor spaces depends, to a large extent, on the direct component of solar radiation incident on the human body. In turn, even the diffuse component of the solar radiation could affect the thermal sensations of people. Despite this evidence, at the present there is a lack in the availability of simple and reliable methods capable of taking into account the influence of the solar radiation on thermal balance in the human body. In this work a comprehensive method is presented for the computation of the mean radiant temperature of people in thermal moderate indoor environments in the presence of solar radiation. The effects produced on the amount of solar radiation entering rooms in the presence of shadowing devices are also analysed. Finally, an application of the method is provided for a non-parallelepiped room equipped with a south window: results are shown in terms of the mean radiant temperature. A simple evaluation of thermal comfort conditions, referring to the present international standards, is also provided. The model can be easily linked to the computerized methods for analyzing the thermal behaviour of buildings, and is intended as a support for the thermal comfort evaluation methods. (author)

  2. Occupant feedback based model predictive control for thermal comfort and energy optimization: A chamber experimental evaluation

    International Nuclear Information System (INIS)

    Chen, Xiao; Wang, Qian; Srebric, Jelena

    2016-01-01

    Highlights: • This study evaluates an occupant-feedback driven Model Predictive Controller (MPC). • The MPC adjusts indoor temperature based on a dynamic thermal sensation (DTS) model. • A chamber model for predicting chamber air temperature is developed and validated. • Experiments show that MPC using DTS performs better than using Predicted Mean Vote. - Abstract: In current centralized building climate control, occupants do not have much opportunity to intervene the automated control system. This study explores the benefit of using thermal comfort feedback from occupants in the model predictive control (MPC) design based on a novel dynamic thermal sensation (DTS) model. This DTS model based MPC was evaluated in chamber experiments. A hierarchical structure for thermal control was adopted in the chamber experiments. At the high level, an MPC controller calculates the optimal supply air temperature of the chamber heating, ventilation, and air conditioning (HVAC) system, using the feedback of occupants’ votes on thermal sensation. At the low level, the actual supply air temperature is controlled by the chiller/heater using a PI control to achieve the optimal set point. This DTS-based MPC was also compared to an MPC designed based on the Predicted Mean Vote (PMV) model for thermal sensation. The experiment results demonstrated that the DTS-based MPC using occupant feedback allows significant energy saving while maintaining occupant thermal comfort compared to the PMV-based MPC.

  3. Durability evaluation method for contact component interconnections in printed circuit boards under thermal loads

    Science.gov (United States)

    Azin, Anton; Zhukov, Andrey A.; Ponomarev, Sergey A.; Ponomarev, Sergey V.

    2017-11-01

    In designing sophisticated printed circuit boards, required evaluation of the product life cycle is relevant in terms of presumably applied load conditions during operation. The paper describes the durability evaluation method of printed circuit board components with contact output such as ball grid array (BGA) and pitch grid array (PGA) under thermal loads. Experiment data and numerical simulation results of soldered connections have been obtained. This method is demonstrated by a practical application example-evaluating of printed circuit board durability under cyclic thermal loads. The application of proposed method for printed circuit boards would make it possible to predict the service life of these designed products.

  4. Thermal Evaluation of Scorched Graphite-Epoxy Panels by Infrared Scanning.

    Science.gov (United States)

    Slifka, A J; Hall, T; Boltz, E S

    2003-01-01

    A simple measurement system is described for evaluating damage to graphite-epoxy panels, such as those used in high-performance aircraft. The system uses a heating laser and infrared imaging system to measure thermal performance. Thermal conductivity or diffusivity is a sensitive indicator of damage in materials, allowing this thermal measurement to show various degrees of damage in graphite-epoxy composites. Our measurements track well with heat-flux damage to graphite epoxy panels. This measurement system, including analysis software, could easily be used in the field, such as on the deck of an aircraft carrier or at remote air strips.

  5. Evaluation of polarization of embedded piezoelectrics by the thermal wave method.

    Science.gov (United States)

    Suchaneck, Gunnar; Eydam, Agnes; Hu, Wenguo; Kranz, Burkhart; Drossel, Welf-Guntram; Gerlach, Gerald

    2012-09-01

    This work demonstrates the benefit of the thermal wave method for the evaluation of the polarization state of embedded piezoelectrics. Two types of samples were investigated: A low-temperature co-fired ceramics (LTCC)/lead zirconate titanate (PZT) sensor-actuator and a macro-fiber composite (MFC) actuator. At modulation frequencies below 10 Hz, the pyroelectric response was governed by thermal losses to the embedding layers. Here, the sample behavior was described by a harmonically heated piezoelectric plate exhibiting heat losses to the environment characterized by a single thermal relaxation time.

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

    OpenAIRE

    Ebadian B.; Parkan MA.

    2002-01-01

    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 res...

  7. Performance evaluation of low concentrating photovoltaic/thermal systems: A case study from Sweden

    OpenAIRE

    Bernardo, Ricardo; Perers, Bengt; Håkansson, Håkan; Karlsson, Björn

    2011-01-01

    Some of the main bottlenecks for the development and commercialization of photovoltaic/thermal hybrids are the lack of an internationally recognized standard testing procedure as well as a method to compare different hybrids with each other and with conventional alternatives. A complete methodology to characterize, simulate and evaluate concentrating photovoltaic/thermal hybrids has been proposed and exemplified in a particular case study. By using the suggested testing method, the hybrid par...

  8. 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.

  9. 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 ...

  10. 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.

  11. 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)

    LANGEVIN, A.S.

    1999-01-01

    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

  12. Design Of Dry Cask Storage For Serpong Multipurpose Reactor Spent Nuclear Fuel

    Directory of Open Access Journals (Sweden)

    Dyah Sulistyani Rahayu

    2018-03-01

    Full Text Available DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL. The spent nuclear fuel (SNF from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF. At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decayat a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method basedon SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister.  The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i

  13. CHARACTERISTICS OF NEXT-GENERATION SPENT NUCLEAR FUEL (SNF) TRANSPORT AND STORAGE CASKS

    Energy Technology Data Exchange (ETDEWEB)

    Haire, M.J.; Forsberg, C.W.; Matveev, V.Z.; Shapovalov, V.I.

    2004-10-03

    The design of spent nuclear fuel (SNF) casks used in the present SNF disposition systems has evolved from early concepts about the nuclear fuel cycle. The reality today is much different from that envisioned by early nuclear scientists. Most SNF is placed in pool storage, awaiting reprocessing (as in Russia) or disposal at a geologic SNF repository (as in the United States). Very little transport of SNF occurs. This paper examines the requirements for SNF casks from today's perspective and attempts to answer this question: What type of SNF cask would be produced if we were to start over and design SNF casks based on today's requirements? The characteristics for a next-generation SNF cask system are examined and are found to be essentially the same in Russia and the United States. It appears that the new depleted uranium dioxide (DUO2)-steel cermet material will enable these requirements to be met. Depleted uranium (DU) is uranium in which a portion of the 235U isotope has been removed during a uranium enrichment process. The DUO2-steel cermet material is described. The United States and Russia are cooperating toward the development of a next-generation, dual-purpose, storage and transport SNF system.

  14. 76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ...-0007] RIN 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY... or the Commission) is proposing to amend its spent fuel storage cask regulations to add the HI-STORM...: June 13, 2011. SAR Submitted by: Holtec International, Inc. SAR Title: Safety Analysis Report on the HI...

  15. Effects of temperature on concrete cask in a dry storage facility for spent nuclear fuels

    International Nuclear Information System (INIS)

    Huang Weiqing; Wu Ruixian; Zheng Yukuan

    2011-01-01

    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)

  16. Shielding effectiveness of the modified cask loading station in FFTF/RSB

    International Nuclear Information System (INIS)

    Simons, R.L.

    1994-01-01

    The Cask Loading Station (CLS) designed for the Maintenance and Storage Facility (MASF) was modified so that it could be placed in the Fast Flux Test Facility Reactor Service Building (FFTF/RSB). The modification to the CLS design required to adapt the CLS t the FFTF/RSB included an aluminum cover plate between the CLS cover plate and cell walls. The purpose of the present analysis was to determine if scattered radiation through the aluminum covered gap significantly increased the radiation dose rate at the 550 foot level. No significant dose rates by this pathway were found. The CLS design for the MASF excluded a supplemental shield block due to low calculated dose rates. A review of the shielding evaluation for the MASF CLS concluded that extra shielding was included around the core component container (CCC) that caused the low calculated dose rate and the Co60 gamma ray source was not considered. Inclusion of these two factors caused the dose rate to increase substantially around the sides of the floor valve and adapter plates. However, either inclusion of the supplemental shield blocks or restricting the accessibility to the area during the CCC transfer reduces the dose rates below the 200 mrem/hr transient limit

  17. Simulated Solar Flare X-Ray and Thermal Cycling Durability Evaluation of Hubble Space Telescope Thermal Control Candidate Replacement Materials

    Science.gov (United States)

    deGroh, Kim K.; Banks, Bruce A.; Sechkar, Edward A.; Scheiman, David A.

    1998-01-01

    During the Hubble Space Telescope (HST) second servicing mission (SM2), astronauts noticed that the multilayer insulation (MLI) covering the telescope was damaged. Large pieces of the outer layer of MLI (aluminized Teflon fluorinated ethylene propylene (Al-FEP)) were torn in several locations around the telescope. A piece of curled up Al-FEP was retrieved by the astronauts and was found to be severely embrittled, as witnessed by ground testing. Goddard Space Flight Center (GSFC) organized a HST MLI Failure Review Board (FRB) to determine the damage mechanism of FEP in the HST environment, and to recommend replacement insulation material to be installed on HST during the third servicing mission (SM3) in 1999. Candidate thermal control replacement materials were chosen by the FRB and tested for environmental durability under various exposures and durations. This paper describes durability testing of candidate materials which were exposed to charged particle radiation, simulated solar flare x-ray radiation and thermal cycling under load. Samples were evaluated for changes in solar absorptance and tear resistance. Descriptions of environmental exposures and durability evaluations of these materials are presented.

  18. Evaluation of Thermal and Thermo-mechanical Behavior of Full-scale Energy Foundations

    Science.gov (United States)

    Murphy, Kyle D.

    This study focuses on the thermo-mechanical and thermal behavior of full-scale energy foundations installed as part of two buildings recently constructed in Colorado. The soil stratigraphy at each of the sites differed, but both foundations were expected to function as primarily end-bearing elements with a tip socketed into rock. The heat exchanger configurations were also different amongst the foundations at both sites, permitting evaluation of the role of heat exchange. A common thread for both energy foundation case histories was the monitoring of the temperature and axial strain within the foundations during heat exchange operations. The first case study involves an evaluation of the long-term thermo-mechanical response of two full-scale energy foundations installed at the new Denver Housing Authority (DHA) Senior Living Facility at 1099 Osage St. in Denver, Colorado. Due to the construction schedule for this project, the thermal properties of the foundations and surrounding subsurface could not be assessed using thermal response tests. However, instrumentation was incorporated into the foundations to assess their long-term heat exchange response as well as the thermo-mechanical strains, stresses, and displacements that occurred during construction and operation of the ground-source heat pump system. The temperature changes within the foundations during heating and cooling operations over a period of approximately 600 days ranged from 9 to 32 °C, respectively. The thermal axial stresses in the foundations were calculated from the measured strains, and ranged from 3.1 MPa during heating to --1.0 MPa during cooling. These values are within reasonable limits for reinforced concrete structures. The maximum thermal axial stress was observed near the toe of both foundations, which is consistent with trends expected for end-bearing toe boundary conditions. The greatest thermal axial strains were observed near the top of the foundations (upward expansion during

  19. Design and Evaluation of a Thermal Tactile Display for Colour Rendering

    Directory of Open Access Journals (Sweden)

    Zhen Jia

    2015-11-01

    Full Text Available This paper proposes a novel method of manipulating both thermal change rate and thermal intensity to convey colour information by using a thermal tactile display. The colour-space transformation from {red, green, blue} to {hue, saturation, intensity} is introduced, and the mapping between colour and temperature is established based on warm and cold colours. Considering the lower resolution of the tactile channel, six limited stimulation levels are generated to represent colours. Based on the semi-infinite body model, the thermal response within the skin for each stimulation form is investigated. The Peltier element of the display is designed to convey different thermal stimuli to the human finger. Two experiments are performed to evaluate the performance of the display: colour identification and discrimination. Experimental results indicate that there is a response bias among the perceived colours for the traditional method of only employing thermal intensity, but there is no response bias for the proposed method; subjects’ mean recognition accuracy with the proposed method is significantly higher than that gained using the traditional method. Furthermore, colour information of the captured images can be reliably discriminated by using this devised thermal tactile display.

  20. Ground-based thermal imaging of stream surface temperatures: Technique and evaluation

    Science.gov (United States)

    Bonar, Scott A.; Petre, Sally J.

    2015-01-01

    We evaluated a ground-based handheld thermal imaging system for measuring water temperatures using data from eight southwestern USA streams and rivers. We found handheld thermal imagers could provide considerably more spatial information on water temperature (for our unit one image = 19,600 individual temperature measurements) than traditional methods could supply without a prohibitive amount of effort. Furthermore, they could provide measurements of stream surface temperature almost instantaneously compared with most traditional handheld thermometers (e.g., >20 s/reading). Spatial temperature analysis is important for measurement of subtle temperature differences across waterways, and identification of warm and cold groundwater inputs. Handheld thermal imaging is less expensive and equipment intensive than airborne thermal imaging methods and is useful under riparian canopies. Disadvantages of handheld thermal imagers include their current higher expense than thermometers, their susceptibility to interference when used incorrectly, and their slightly lower accuracy than traditional temperature measurement methods. Thermal imagers can only measure surface temperature, but this usually corresponds to subsurface temperatures in well-mixed streams and rivers. Using thermal imaging in select applications, such as where spatial investigations of water temperature are needed, or in conjunction with stationary temperature data loggers or handheld electronic or liquid-in-glass thermometers to characterize stream temperatures by both time and space, could provide valuable information on stream temperature dynamics. These tools will become increasingly important to fisheries biologists as costs continue to decline.