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Sample records for shipping cask analysis

  1. 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

  2. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  3. Impact analysis of shipping casks

    International Nuclear Information System (INIS)

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

    1989-01-01

    This 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 calculation. 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: a thirty foot axial drop on either end, a thirty foot oblique angel drop with the cask having several different orientations from the vertical with impact on the top end cask corner, and a thirty foot side drop with simultaneous impact on one of the lifting trunnions and the bottom end. Prevention of damage hinges on the strength of the various components that comprises the cask. The predicted levels of deformation and stresses in the cask are used to assess the potential damage level

  4. Free drop impact analysis of shipping cask

    International Nuclear Information System (INIS)

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

    1989-01-01

    The WHAMS-2D and WHAMS-3D codes were used to analyze the dynamic response of the RAS/TREAT shielded shipping cask subjected to transient leadings for the purpose of assessing potential damage to the various components that comprise the the cask. The paper describes how these codes can be used to provide and intermediate level of detail between full three-dimensional finite element calculations and hand calculations which are cost effective for design purposes. Three free drops were adressed: (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 cask corner; and (3) a thirty foot side drop with simultaneous impact on the lifting trunnion and the bottom end. Results are presented for two models of the side and oblique angle drops; one model includes only the mass of the lapped sleeves of depleted uranium (DU) while the other includes the mass and stiffness of the DU. The results of the end drop analyses are given for models with and without imperfections in the cask. Comparison of the analysis to hand calculations and simplified analyses are given. (orig.)

  5. Criticality analysis of a spent fuel shipping cask

    International Nuclear Information System (INIS)

    Pena, J.

    1984-01-01

    Criticality analysis for a system yields to the determination of the multiplication factor. Should such analysis be performed for a spent fuel shipping cask some standards must be accomplished. In this study a sample design is analyzed and criticality results are presented. (author)

  6. Structural analysis of closure bolts for shipping casks

    International Nuclear Information System (INIS)

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

    1993-04-01

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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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 presentation made at the CUBE (Computer Use By Engineers) Symposium, October 1976, in Albuquerque, New Mexico is summarized. A full-scale testing program involving impact tests of spent-nuclear-fuel shipping systems is described. This program is being conducted by Sandia Laboratories for the Environmental Control Technology Division of the U.S. Energy Research and Development Administration. The analytical and scale modeling techniques being employed to predict the response of the full-scale system in the very severe impact tests are described. 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

  12. 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

  13. Safety analysis report for packaging: neutron shipping cask, model 0.5T

    International Nuclear Information System (INIS)

    Peterson, R.T.

    1976-01-01

    The Safety Analysis Report for Packaging demonstrates that the neutron shipping cask can safely transport, in solid or powder form, all isotopes of uranium, plutonium, americium, curium, berkelium, californium, einsteinium, and fermium. The shipping cask and its contents are described. It also evaluates transport conditions, structural parameters (e.g., load resistance, pressure and impact effects, lifting and tiedown devices), and shielding. Finally, it discusses compliance with Chapter 0529 of the Energy Research and Development Administration Manual

  14. 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

  15. 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

  16. 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.

  17. Prototypic fabrication of TRIGA irradiated fuel shipping casks

    International Nuclear Information System (INIS)

    Kim, B.K.; Lee, Y.W.; Whang, C.K.; Lee, J.B.

    1980-01-01

    This is the safety analysis report on the prototypic fabrication of ''TRIGA Irradiated Fuel Shipping Cask'' conducted by KAERI in 1980. The results of the evaluation show that the shipping cask is in compliance with the applicable regulation for the normal conditions of transport as well as hypothetical accident conditions. The prototypic fabrication of the shipping cask (type B) was carried out for the first time in Korea after getting technical experience from fabrication of the ''TRIGA Spent Fuel Shipping Cask'' and ''the KO-RI Unit 1 surveillance capsule shipping cask'' in 1979. This report contains structural evaluation, thermal evaluation, shielding, criticality, quality assurance, and handling procedures of the shipping cask

  18. Safety analysis report: packages cobalt-60 shipping cask (packaging of radioactive and fissile materials)

    International Nuclear Information System (INIS)

    Evans, J.E.; Langhaar, J.W.

    1973-07-01

    Safety Analysis Report DPSPU-73-124-1 replaces DPSPU-69-124-1 and Supplement 1 to permit shipment of 350,000 curies of 60 Co (maximum) in cobalt-60 shipping casks in compliance with 10 CFR Part 71, Packaging of Radioactive Materials for Transport

  19. Safety analysis report for packaging: neutron shipping cask, model 4T

    International Nuclear Information System (INIS)

    Peterson, R.T.

    1977-01-01

    This Safety Analysis Report for Packaging demonstrates that the neutron shipping cask can safely transport, in solid or powder form, all isotopes of uranium, plutonium, americium, curium, berkelium, californium, einsteinium, and fermium. The cask and its contents are described. It also evaluates transport conditions, structural parameters (e.g., load resistance, pressure and impact effects, lifting and tiedown devices), and shielding. Finally, it discusses compliance with Chapter 0529 of the Energy Research and Development Administration Manual, Safety Standards for the Packaging of Fissile and Other Radioactive Materials

  20. 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)

  1. 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

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

    International Nuclear Information System (INIS)

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

    1977-11-01

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

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

    International Nuclear Information System (INIS)

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

    1977-11-01

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

  4. Spent fuel shipping cask sealing concepts

    International Nuclear Information System (INIS)

    Sonnier, C.S.

    1989-05-01

    In late 1985, the International Atomic Energy Agency (IAEA) requested the US Program for Technical Assistance to IAEA Safeguards (POTAS) to provide a study which examined sealing concepts for application to spent fuel shipping casks. This request was approved, and assigned to Sandia National Laboratories (Sandia). In the course of this study, discussions were held with personnel in the International Safeguards Community who were familiar with the shipping casks used in their States. A number of shipping casks were examined, and discussions were held with two shipping cask manufacturers in the US. As a result of these efforts, it was concluded that the shipping casks provided an extremely good containment, and that many of the existing casks can be effectively sealed by applying the seal to the cask closure bolts/nuts

  5. Thermal analysis of a one-element PWR spent fuel shipping cask

    International Nuclear Information System (INIS)

    Fields, S.R.

    1979-06-01

    The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

  6. Comparison of elastic and inelastic analysis and test results for the defense high level waste shipping cask

    International Nuclear Information System (INIS)

    Zimmer, A.; Koploy, M.A.; Madsen, M.M.

    1991-01-01

    In the early 1980s, the US DOE/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as fully complies with all applicable DOE, Nuclear Regulatory Commission, and DOT regulations. General Atomics designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Labs. (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This paper will compare the results of the two analytical approaches and with model testing results. The purpose of this work is to provide data to support licensing of the DHLW cask and to support the acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing

  7. 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

  8. Comparison of structural computer programs used for the analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Friley, J.R.

    1984-09-01

    Several structural analysis computer programs were selected and used in analyses relevant to the hypothetical impact requirements for spent fuel shipping cask designs. The objective of the study was to evaluate the computer codes by performing a series of analyses and comparing results. The code evaluation efforts treated end and side impact situations only. As a result, the models were either one or two dimensional. Both clad lead and solid wall construction types were considered. For clad lead models, frictionless sliding between the lead and cladding was assumed. General agreement was achieved between the codes for problems involving non-clad models. For clad models, agreement between the codes was poor. This work was sponsored by the Department of Energy through the Transportation Technology Center at Sandia National Laboratories, Albuquerque, New Mexico. 16 references, 18 figures, 9 tables

  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. Nondestructive evaluation of LWR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructive testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined

  11. 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

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

    International Nuclear Information System (INIS)

    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

  13. 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.

  14. 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)

  15. 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)

  16. 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)

  17. Analysis, scale modeling, and full-scale test of a railcar and spent-nuclear-fuel shipping cask in a high-velocity impact against a rigid barrier

    International Nuclear Information System (INIS)

    Huerta, M.

    1981-06-01

    This report describes the mathematical analysis, the physical scale modeling, and a full-scale crash test of a railcar spent-nuclear-fuel shipping system. The mathematical analysis utilized a lumped-parameter model to predict the structural response of the railcar and the shipping cask. The physical scale modeling analysis consisted of two crash tests that used 1/8-scale models to assess railcar and shipping cask damage. The full-scale crash test, conducted with retired railcar equipment, was carefully monitored with onboard instrumentation and high-speed photography. Results of the mathematical and scale modeling analyses are compared with the full-scale test. 29 figures

  18. Development of the nuclear ship MUTSU spent fuel shipping cask

    International Nuclear Information System (INIS)

    Ishizuka, M.; Umeda, M.; Nawata, Y.; Sato, H.; Honami, M.; Nomura, T.; Ohashi, M.; Higashino, A.

    1989-01-01

    After the planned trial voyage (4700 MWD/MTU) of the nuclear ship MUTSU in 1990, her spent fuel assemblies, initially made of two types of enriched UO 2 (3.2wt% and 4.4wt%), will be transferred to the reprocessing plant soon after cooling down in the ship reactor for more than one year. For transportation, the MUTSU spent fuel shipping casks will be used. Prior to transportation to the reprocessing plant, the cooled spent fuel assemblies will be removed from the reactor to the shipping casks and housed at the spent fuel storage facility on site. In designing the MUTSU spent fuel shipping cask, considerations were given to make the leak-tightness and integrity of the cask confirmable during storage. The development of the cask and the storage function demonstration test were performed by Japan Atomic Energy Research Institute (JAERI) and Mitsubishi Heavy Industries, Ltd. (MHI). One prototype cask for the storage demonstration test and licensed thirty-five casks were manufactured between 1987 and 1988

  19. Nondestructive evaluation of LWR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study

  20. Feasibility and incentives for the consideration of spent fuel operating histories in the criticality analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Jones, R.H.

    1987-08-01

    Analyses have been completed that indicate the consideration of spent fuel histories (''burnup credit'') in the design of spent fuel shipping casks is a justifiable concept that would result in cost savings and public risk benefits in the transport of spent nuclear fuel. Since cask capacities could be increased over those of casks without burnup credit, the number of shipments necessary to transport a given amount of fuel could be reduced. Reducing the number of shipments would increase safety benefits by reducing public and occupational exposure to both radiological and nonradiological risks associated with the transport of spent fuel. Economic benefits would include lower in-transit shipping, reduced transportation fleet capital costs, and reduced numbers of cask handling operations at both shipping and receiving facilities. 44 refs., 66 figs., 28 tabs

  1. 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

  2. Safety analysis report: packages 238Pu oxide shipping cask (packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Evans, J.E.; Gates, A.A.

    1975-06-01

    Plutonium-238 (as PuO 2 powder) is shipped in triple-container stainless steel shipping casks in compliance with ERDA Manual Chapter 0529 (ERDAM 0529), Safety Standards for the Packaging of Fissile and Other Radioactive Materials. (U.S.)

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

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1991-08-01

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

  4. Capabilities for processing shipping casks at spent fuel storage facilities

    International Nuclear Information System (INIS)

    Baker, W.H.; Arnett, L.M.

    1978-01-01

    Spent fuel is received at a storage facility in heavily shielded casks transported either by rail or truck. The casks are inspected, cooled, emptied, decontaminated, and reshipped. The spent fuel is transferred to storage. The number of locations or space inside the building provided to perform each function in cask processing will determine the rate at which the facility can process shipping casks and transfer spent fuel to storage. Because of the high cost of construction of licensed spent fuel handling and storage facilities and the difficulty in retrofitting, it is desirable to correctly specify the space required. In this paper, the size of the cask handling facilities is specified as a function of rate at which spent fuel is received for storage. The minimum number of handling locations to achieve a given throughput of shipping casks has been determined by computer simulation of the process. The simulation program uses a Monte Carlo technique in which a large number of casks are received at a facility with a fixed number of handling locations in each process area. As a cask enters a handling location, the time to process the cask at that location is selected at random from the distribution of process time. Shipping cask handling times are based on experience at the General Electric Storage Facility, Morris, Illinois. Shipping cask capacity is based on the most recent survey available of the expected capability of reactors to handle existing rail or truck casks

  5. 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.

  6. Dry cask handling system for shipping nuclear fuel

    International Nuclear Information System (INIS)

    Jones, C.R.

    1975-01-01

    A nuclear facility is described for improved handling of a shipping cask for nuclear fuel. After being brought into the building, the cask is lowered into a tank mounted on a transporter, which then carries the tank into a position under an auxiliary well to which it is sealed. Fuel can then be loaded into or unloaded from the cask via the auxiliary well which is flooded. Throughout the procedure, the cask surface remains dry. (U.S.)

  7. Spent and fresh fuel shipping cask considerations

    International Nuclear Information System (INIS)

    Shappert, L.B.; Unger, W.E.; Freedman, J.M.

    1975-01-01

    A program to provide basic information for cask design and safety has been conducted for over ten years at Oak Ridge National Laboratory. Principal problem areas in Liquid Metal Fast Breeder Reactor (LMFBR) casks are identified as heat transfer, structures and containment, criticality and shielding. Solutions in the problem areas, as well as the need for future work, are addressed by describing an LMFBR conceptual design cask. A new program, which is underway at Sandia Laboratories, Albuquerque, New Mexico, is aimed at producing technology useful to industry and government. Technologies are being developed in areas of hazards analysis, heat transfer, shielding, structures and containment, and spent fuel characterization, substantiated by hot laboratory verification. Particular emphasis will be placed on establishing qualification tests based on accident experience. Handling requirements and limitations are discussed. (auth)

  8. 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

  9. Prototypical fabrication of PWR spent fuel shipping cask

    International Nuclear Information System (INIS)

    Kwack, Eun Ho; Kim, Byung Ku; Kang, Hee Yung; Lee, Chung Young; Jeon, Kyeong Lak; Lee, Bum Soo

    1985-02-01

    This report describes about the safety analysis for the spent fuel shipping cask, which is used to transfer a single fuel assembly discharged from PWR in operation in Korea. The contents cover the methods and the results of structural, thermal, thermo-hydraulic, radiation shield and criticality detail analysis. The safety evaluation has been made under the normal transportation and hypothetical accident conditions such as 30ft free drop, puncture, fire, immersion, penetration, corner drop, etc,. Some corrections in design are made, and a brief information for fabrication and transportation are obtained by the use of a 1/6 scale model. The design is based on one year cooling time of the spent fuel with 40,000 MWT/MTU maximum burnup, which gives 7.2KW decay heat and 1.6x10 6 ci/hr radiation intensity. The cask is composed of main body with the double closures, impact limiter and fuel basket. The inner shell, inner closure and valves constitute the pressure boundary of the containment. The inner, intermediate and outer shells, upper and lower forgings are made of stainless steel which compose the main body with lead for gamma shield and 50% ethylene glycol for neutron shield. The impact limiters are made of balsa wood on both end sides of the cask to protect the cask from a sudden shocks in accident during the transportation. The analysis results show that the cask is proved to retain its structural integrity within allowable stress and to be safe under the normal and hypothetical accident conditions, and the maximum dose rates of radiation at 2m distance from the surface of the cask are less than the required values. The weight will be 23.2tons in dry and 27.8 tons in wet with fuel loaded. All the design data, calculated results for the structural integrity, shield and thermal analysis are shown in this report with the basic drawings. (Author)

  10. Status of the Beneficial Uses Shipping System cask (BUSS)

    International Nuclear Information System (INIS)

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

    1994-01-01

    The Beneficial Uses Shipping System cask is a Type B packaging developed by Sandia National Laboratories for the U.S. Department of Energy. The cask is designed to transport special form radioactive source capsules (cesium chloride and strontium fluoride) produced by the Department of Energy's Hanford Waste Encapsulation and Storage Facility. This paper describes the cask system and the analyses performed to predict the response of the cask in impact, puncture, and fire accident conditions as specified in the regulations. The cask prototype has been fabricated and Certificates of Compliance have been obtained

  11. Benchmark study of some thermal and structural computer codes for nuclear shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Kanae, Yoshioki; Shimada, Hirohisa; Shimoda, Atsumu; Halliquist, J.O.

    1984-01-01

    There are many computer codes which could be applied to the design and analysis of nuclear material shipping casks. One of problems which the designer of shipping cask faces is the decision regarding the choice of the computer codes to be used. For this situation, the thermal and structural benchmark tests for nuclear shipping casks are carried out to clarify adequacy of the calculation results. The calculation results are compared with the experimental ones. This report describes the results and discussion of the benchmark test. (author)

  12. Structural code benchmarking for the analysis of impact response of nuclear material shipping casks

    International Nuclear Information System (INIS)

    Glass, R.E.

    1984-01-01

    The Transportation Technology Center at Sandia National Laboratories has initiated a program to benchmark thermal and structural codes that are available to the nuclear material transportation community. The program consists of the following five phrases: (1) code inventory and review, (2) development of a cask-like set of problems, (3) multiple independent numerical analyses of the problems, (4) transfer of information, and (5) performance of experiments to obtain data for comparison with the numerical analyses. This paper will summarize the results obtained by the independent numerical analyses. The analyses indicate the variability that can be expected both due to differences in user-controlled parameters and from code-to-code differences. The results show that in purely elastic analyses, differences can be attributed to user controlled parameters. Model problems involving elastic/plastic material behavior and large deformations, however, have greater variability with significant differences reported for implicit and explicit integration schemes in finite element programs. This variability demonstrates the need to obtain experimental data to properly benchmark codes utilizing elastic/plastic material models and large deformation capability

  13. Structural Analysis of Shipping Casks, Vol. 9. Energy Absorption Capabilities of Plastically Deformed Struts Under Specified Impact Loading Conditions (Thesis)

    International Nuclear Information System (INIS)

    Davis, F.C.

    2001-01-01

    The purpose of this investigation was to determine the energy absorption characteristics of plastically deformed inclined struts under impact loading. This information is needed to provide a usable method by which designers and analysts of shipping casks for radioactive or fissile materials can determine the energy absorption capabilities of external longitudinal fins on cylindrical casks under specified impact conditions. A survey of technical literature related to experimental determination of the dynamic plastic behavior of struts revealed no information directly applicable to the immediate problem, especially in the impact velocity ranges desired, and an experimental program was conducted to obtain the needed data. Mild-steel struts with rectangular cross sections were impacted by free-falling weights dropped from known heights. These struts or fin specimens were inclined at five different angles to simulate different angles of impact that fins on a shipping cask could experience under certain accident conditions. The resisting force of the deforming strut was measured and recorded as a function of time by using load cells instrumented with resistance strain gage bridges, signal conditioning equipment, an oscilloscope, and a Polaroid camera. The acceleration of the impacting weight was measured and recorded as a function of time during the latter portion of the testing program by using an accelerometer attached to the drop hammer, appropriate signal conditioning equipment, the oscilloscope, and the camera. A digital computer program was prepared to numerically integrate the force-time and acceleration-time data recorded during the tests to obtain deformation-time data. The force-displacement relationships were then integrated to obtain values of absorbed energy with respect to deformation or time. The results for various fin specimen geometries and impact angles are presented graphically, and these curves may be used to compute the energy absorption capacity of

  14. 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

  15. Scale-4 shipping cask shielding applications

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Parks, C.V.

    1991-01-01

    This paper reports the application of the SCALE-4 shielding sequences SAS1 and SAS4 to the problem set distributed by the Organization for Economic Cooperation and Development (OECD) Working Group on Shielding Assessment of Transportation Packages. In many cases, additional comparison are made with MCNP and QADS solutions to provide a complete cross-check of methods, cross sections, geometry, etc. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask

  16. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    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

  17. CASKCODES, Program CAPSIZE Scope KWIKDOSE for Shipping Cask Shielding

    International Nuclear Information System (INIS)

    1988-01-01

    1 - Description of program or function: CAPSIZE is an interactive program to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel casks designed to meet those objectives. 2 - Method of solution: Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the load cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium- shielded cask meeting those objectives. Using optimal packing arrangements and shielding requirements input by the user, SCOPE will design a cask to carry a single fuel assembly and then continue incrementing the number of assemblies until one or more of the design limits can no longer be met. KWIKDOSE queries the user for the number of PWR fuel assemblies in a cask, the type of cask and thickness of the shield. Upon getting the necessary input, KWIKDOSE prints out the total dose rate, 10 feet from the centerline of the cask, as a function of the burnup and cooling time of the spent fuel. 3 - Restrictions on the complexity of the problem: The restrictions are subject to the shielding requirements of the shipping cask

  18. Spent fuel shipping cask development status

    International Nuclear Information System (INIS)

    Henry, K.H.; Lattin, W.C.

    1989-01-01

    The Nuclear Waste Policy Act of 1982 (NWPA) authorized the US Department of Energy (DOE) to establish a national system for the disposal of spent nuclear fuel and high-level radioactive waste from commercial power generation, and established the Office of Civilian Radioactive Waste Management (OCRWM) within the DOE-Headquarters (DOE-HQ) to carry out these duties. A 1985 presidential decision added the disposal of high-level radioactive waste generated by defense programs to the national disposal system. A primary element of the disposal program is the development and operation of a transportation system to move the waste from its present locations to the facilities that will be included in the waste management system. The primary type of disposal facility to be established is a geologic repository; a Monitored Retrievable Storage (MRS) facility may also be included as an intermediate step in the nuclear waste disposal process. This paper focuses on the progress and status of one facet of the transportation program--the development of a family of shipping casks for transporting spent fuel from nuclear power reactor sites to the repository of MRS facility

  19. 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

  20. DOE procurement activities for spent fuel shipping casks

    International Nuclear Information System (INIS)

    Callaghan, E.F.; Lake, W.H.

    1988-01-01

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

  1. Design criteria for the structural analysis of shipping cask containment vessels

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    10 CFR Part 71, Sections 71.35 and 71.36, require that packages used to transport radioactive materials meet specified normal and hypothetical accident conditions. Acceptable design criteria are presented for use in the structural analysis of the containment vessels of Type B packages used to transport irradiated nuclear fuel. Alternative design criteria meeting the structural requirements of 10 CFR Part 71, Section 71.35 and 71.36, may also be used

  2. Thermal Evaluation of a KRI-BGM Shipping Cask

    International Nuclear Information System (INIS)

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

    2007-01-01

    Radioactive isotopes are used extensively in the fields of industry, medical treatment, food and agriculture. Use of radioactive isotopes is expected to increase continuously with the growth of each field. In order to safely transport radioactive isotopes from the place of manufacture to the place of use, a shipping package is required. Therefore KAERI is developing the KRI-BGM shipping cask to transport the Ir-192 bulk radioactive material, which is produced at the HANARO research reactor. The shipping package should satisfy the requirements which are prescribed in the Korea MOST Act 2001-23, IAEA Safety Standard Series No. TS-R-1, US 10 CFR Part 71 and the US 49 CFR Part 173. These regulatory classify the KRI-BGM shipping cask as a Type B package, and their regulatory guidelines state that the Type B package for transporting radioactive materials should be able to withstand a period of 30 minutes under a thermal condition of 800 .deg. C. However, the polyurethane, which is to be used as the filling within the cavity of the KRIBGM shipping cask, has a very weak characteristic in a high temperature. Therefore it is difficult for the depleted uranium(hereafter DU), which is used as shielding material, to be protected under a thermal condition of 800 .deg. C. Accordingly, the KRI-BGM shipping cask, which applied non-combustible polyurethane and fireproof materials as the filling, was fabricated. The thermal tests by using prototype cask have been performed to estimate the thermal integrity of the KRI-BGM shipping cask under a thermal condition of 800 .deg. C

  3. BR-100 spent fuel shipping cask development

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  4. 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

  5. PRE-CASKETSS: an input data generation 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 PRE-CASKETSS has been developed for the purpose of input data generation 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 PRE-CASKETSS are as follow; (1) Function of input data generation for thermal and structural analysis computer programs is provided in the program. (2) Two- and three-dimensional mesh generation for finite element and finite difference programs are available in the program. (3) The capacity of the material input data generation are provided in the program. (4) The boundary conditions, the load conditions and the initial conditions are capable in the program. (5) This computer program operate 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)

  6. High-burnup/low-cooling-time fuel carrying capacity of the GA-4 and GA-9 spent fuel shipping casks

    International Nuclear Information System (INIS)

    Boshoven, J.K.; Hopf, J.E.

    1994-01-01

    In response to utilities' projected needs to ship higher burnup spent fuel, General Atomics (GA) has performed shielding and thermal analysis for the GA-4 and GA-9 legal weight shipping casks to determine the minimum cooling times for various burnup levels for fully loaded GA-4 and GA-9 casks and reduced payloads for the casks. Tables are provided in the paper which show the minimum cooling time for a given burnup and payload for each of the casks. The analyses show that the GA-4 and GA-9 casks can carry at least as many high-burnup and/or short-cooling-time spent fuel assemblies as present day shipping casks. In addition, the GA casks are able to carry at least twice as many assemblies as the present day shipping casks if the spent fuel burnup levels and/or cooling times are open-quotes coolerclose quotes or open-quotes as coolclose quotes as their design basis fuels. The increased shipping capacity for these more common open-quotes coolerclose quotes assemblies allows fewer shipments and therefore increases the efficiency and lowers predicted risks of the transport system

  7. Incentives for the allowance of burnup credit in the design of spent nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Jones, R.H.

    1987-01-01

    An analysis has been completed which indicates that the consideration of spent fuel histories ('burnup credit') in the criticality design of spent fuel shipping casks could result in considerable public risk benefits and cost savings in the transport of spent nuclear fuel. Capacities of casks could be increased considerably in some cases. These capacity increases result in lower public and occupational exposures to ionizing radiation due to the reduced number of shipments necessary to transport a given amount of fuel. Additional safety benefits result from reduced non-radiological risks to both public and occupational sectors. In addition, economic benefits result from lower in-transit shipping costs, reduced transportation fleet capital costs, and fewer cask handling requirements at both shipping and receiving facilities

  8. 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.

    1984-12-01

    The PuO 2 cask or 5320-3 cask is designed for shipment of americium or plutonium by surface transportation modes. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, dated 12/21/76, which invokes Title 10 Code of Federal Regulations, Part 71 (10 CFR 71) ''Packaging of Radioactive Materials for Transport,'' and Title 49 CFR Parts 171.179 ''Hazardous Materials Regulations.'' (US DOE Order 4580.1A, Chapter III, superseded manual chapter 0529 effective May 1981, but it retained the same 10 CFR 71 and 49 CFR 171-179 references

  9. 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

  10. 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)

  11. Documentation for first annual testing and inspections of Benificial Uses Shipping System (BUSS) Cask

    International Nuclear Information System (INIS)

    Lundeen, J.E.

    1994-01-01

    The purpose of this report is to compile date generated during the first annual tests and inspections of the Benificiai Uses Shipping System (BUSS) Cask. In addition, this report will verify that the testing criteria identified in chapter 8 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met. 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 lifting lugs; and torque test on all bolts; impact limiter inspection and weight test. The first annual inspections and testing of the BUSS Cask were completed on May 5, 1994, and met the SARP criteria

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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%

  17. Temperature and neutron dose rate measurements at a spent fuel shipping cask

    International Nuclear Information System (INIS)

    Krause, F.

    1982-01-01

    Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

  18. Research and development of spent fuel shipping casks and the criteria for seagoing vessel carrying casks

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Considering that the transportation of spent fuel will increase rapidly and extensively in the near future, Japanese Atomic Energy Committee enacted ''Technical Standard for Transportation of Radioactive Materials'' based on ''IAEA Regulation for the Safe Transport of Radioactive Materials 1973 Revised Edition''. Coping with the recommendation of AEC, Atomic Energy Bureau in Science and Technology Agency and other authorities concerned started to review the former ordinances for transportation of radioactive materials and to consolidate a unified system of relevant laws and standards. On the other hand, Atomic Energy Bureau has invested in research and development since ten years ago in order to obtain the data for design and licensing work of spent fuel shipping casks. In those studies some different scale models of a prototype of 80 t in weight have been used to make clear the scale effect at the drop, pucture and fire tests, which are one of the features of Japanese research and development. And also the immersion test in high pressure water up to about 500 bars is now carried out to investigate the integrity of cask body and sealing structure to prevent leakage of radioactive contents to the ambient when the cask falls into deep sea. In Japan, depending on the site conditions of nuclear plants, almost all transportations of unirradiated and spent fuels are done on the sea. Therefore, in order to secure the safety of transportation, the design criteria of the seagoing vessels for exclusive transportation of spent fuel shipping casks, namely full load shipping, has been enacted, which aims to make minimum the probability of sinking at collison, stranding and other unforeseen accidents at sea and also to restrain radiation exposure of the crew as low as possible

  19. Management plan for the procurement of shipping casks required to service proposed federal waste repositories

    International Nuclear Information System (INIS)

    Renken, J.H.; Dupree, S.A.; Allen, G.C. Jr.; Freedman, J.M.

    1978-08-01

    Development of transportation systems to move radioactive waste and unreprocessed spent fuel to proposed federal waste repositories is an integral part of the National Waste Terminal Storage Program. To meet this requirement, shipping casks must be designed, licensed, and fabricated. To assist the manager charged with this responsibility, a Cask Procurement Plan has been formulated. This plan is presented as a logic diagram that is suitable for computer analysis. In addition to the diagram, narrative material that describes various activities in the plan is also included. A preliminary computer analysis of the logic diagram indicates that, depending on the result of several decisions which must be made during the course of the work, the latest start dates which will allow prototype delivery of all types of casks by December 1985, range from November 1977 to March 1982

  20. Simulation of the mechanical behavior of a spent fuel shipping cask in a rail accident environment

    International Nuclear Information System (INIS)

    Fields, S.R.

    1977-02-01

    A preliminary mathematical model has been developed to simulate the dynamic mechanical response of a large spent fuel shipping cask to the impact experienced in a hypothetical rail accident. The report was written to record the status of the development of the mechanical response model and to supplement an earlier report on spent fuel shipping cask accident evaluation

  1. Application of a viscoplastic constitutive law to lead in the impact analysis of radioactive material shipping casks

    International Nuclear Information System (INIS)

    Wang, Zhibi; Turula, P.; Popper, G.F.

    1990-01-01

    Perzyna's viscoplastic material model is selected to consider the strain rate effect of lead used in radioactive material shipping packages. The model is checked using data from two scale-model tests and the deformations are found to be within 10 percent. 3 refs., 4 figs

  2. A comparison of recent results from HONDO III with the JSME nuclear shipping cask benchmark calculations

    International Nuclear Information System (INIS)

    Key, S.W.

    1985-01-01

    The results of two calculations related to the impact response of spent nuclear fuel shipping casks are compared to the benchmark results reported in a recent study by the Japan Society of Mechanical Engineers Subcommittee on Structural Analysis of Nuclear Shipping Casks. Two idealized impacts are considered. The first calculation utilizes a right circular cylinder of lead subjected to a 9.0 m free fall onto a rigid target, while the second calculation utilizes a stainless steel clad cylinder of lead subjected to the same impact conditions. For the first problem, four calculations from graphical results presented in the original study have been singled out for comparison with HONDO III. The results from DYNA3D, STEALTH, PISCES, and ABAQUS are reproduced. In the second problem, the results from four separate computer programs in the original study, ABAQUS, ANSYS, MARC, and PISCES, are used and compared with HONDO III. The current version of HONDO III contains a fully automated implementation of the explicit-explicit partitioning procedure for the central difference method time integration which results in a reduction of computational effort by a factor in excess of 5. The results reported here further support the conclusion of the original study that the explicit time integration schemes with automated time incrementation are effective and efficient techniques for computing the transient dynamic response of nuclear fuel shipping casks subject to impact loading. (orig.)

  3. 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

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

    International Nuclear Information System (INIS)

    Frost, R.L.

    1999-01-01

    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

  5. 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)

  6. Gamma ray benchmark on the spent fuel shipping cask TN 12

    International Nuclear Information System (INIS)

    Blum, P.; Cagnon, R.; Cladel, C.; Ermont, G.; Nimal, J.C.

    1983-05-01

    The purpose of this benchmark is to compare measurements and calculation of gamma-ray dose rates around a shipping cask loaded with 12 spent fuel elements of FESSENHEIM PWR type. The benchmark provides a means to verify gamma-ray sources and gamma-ray transport calculation methods in shipping cask configurations. The comparison between measurements and calculations shows a good agreement except near the fuel element top where the discrepancy reaches a factor 2

  7. 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

  8. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  9. Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask

    International Nuclear Information System (INIS)

    Zimmer, A.; Koploy, M.

    1991-12-01

    In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing

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

    International Nuclear Information System (INIS)

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

    1994-04-01

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

  11. Equivalency relations for mixtures of nuclides in shipping casks 9972-9975

    International Nuclear Information System (INIS)

    Niemer, K.A.; Frost, R.L.; Williamson, T.G.

    1994-01-01

    Equivalence relations required to determine mass limits for mixtures of nuclides for the Safety Analysis Report for Packaging (SARP) of the Savannah River Site 9972, 9973, 9974, and 9975 shipping casks were calculated. The systems analyzed included aqueous spheres, homogeneous metal spheres, and metal ball-and-shell configurations, all surrounded by an effectively infinite stainless steel or water reflector. Comparison of the equivalence calculations with the rule-of-fractions showed conservative agreement for aqueous solutions, both conservative and non-conservative agreement for the metal homogenous sphere systems, and non-conservative agreement for the majority of metal ball-and-shell systems. Equivalence factors for the aqueous solutions and homogeneous metal spheres were calculated. The equivalence factors for the non-conservative metal homogeneous sphere systems were adjusted so that they were conservative. No equivalence factors were calculated for the ball-and-shell systems since the SARP assumes that only homogeneous or uniformly distributed material will be shipped in the 9972-9975 shipping casks, and an unnecessarily conservative critical mass may result if the ball-and-shell configurations are included

  12. Calculation of radiation dose rates from a spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask and the presence of the ground surface, and the effects of these factors on the calculated dose rates are largely unknown. Moreover, similar empirical equations for buildup factors currently do not exist for neutrons. The objective of this study is to use a more accurate approach in calculating radiation dose rates for both neutrons and gamma photons from a spent fuel cask. The calculation utilizes the more sophisticated transport method and takes into account the geometry of the cask and the presence of the ground surface. The results of a detailed study of dose rates in the near field (within 20 meters) are presented and, for easy application, the cask centerline dose rates are fitted into empirical equations at cask centerline distances up to 2000 meters from the surface of the cask

  13. Transport experience with the NAC-1 radioactive materials shipping cask

    International Nuclear Information System (INIS)

    Rollins, J.D.; Hoffman, C.C.

    1976-01-01

    During the first one and one-half years of operation of Nuclear Assurance Corporation's (NAC) four (4) second-generation NAC-1 truck casks, shipments of spent fuel assemblies, fuel rods, and other highly irradiated reactor components have involved over 300,000 cask miles of travel by land, and cask handling at some ten different nuclear facilities. This on-site experience has included the use of various types of auxiliary lifting devices, operational problems with which have identified the need to establish related Quality Assurance procedures in the area of post-fabrication testing. During the course of pre-shipment checkout and testing of the casks minor defects in the upper impact limiter and lower cask shielding wall have been detected and repaired according to procedure. One enroute occurrence with the cask in which an emergency response was implemented has emphasized the need for rigid adherence to procedural checkout before shipment. Periodic inspection and testing are performed as part of the cask license requirement whereby cask components are inspected and/or replaced. During such test periods leaking ball valves and a leaking neutron shield tank have been detected and repaired. (author)

  14. 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

  15. Postmortem metallurgical examination of a fire-exposed spent fuel shipping cask

    International Nuclear Information System (INIS)

    Rack, H.J.; Yoshimura, H.R.

    1980-04-01

    A potmortem examination of a large fire-exposed rail-transported spent fuel shipping container has revealed the presence of two macrofissures in the outer cask shell. The first, a part-thru crack located within the seam weld fusion zone of the outer cask shell, was typical of hot cracks that may be found in stainless steel weldments. The second, located within the stainless steel base metal, apparently originated at microcracks formed during the welding of a copper-stainless steel dissimilar metal joint. The latter microcrack then propagated during the fire-test, ultimately penetrating the outer shall of the cask. 18 figures, 2 tables

  16. A truck cask design for shipping defense high-level waste

    International Nuclear Information System (INIS)

    Madsen, M.M.; Zimmer, A.

    1985-01-01

    The Defense High-Level Waste (DHLW) cask is a Type B packaging currently under development by the U.S. Department of Energy (DOE). This truck cask has been designed to initially transport borosilicate glass waste from the Defense Waste Processing Facility (DWPF) to the Waste Isolation Pilot Plant (WIPP). Specific program activities include designing, testing, certifying, and fabricating a prototype legal-weight truck cask system. The design includes such state-of-the-art features as integral impact limiters and remote handling features. A replaceable shielding liner provides the flexibility for shipping a wide range of waste types and activity levels

  17. Fire resistivity of irradiated nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Shimada, Hirohisa

    1975-01-01

    The fire resistance of lead-lined casks was examined and compared with that of a cask without lead lining. Three cask models with 1/8 radius of actual casks and one with 1/4 radius were used, each one is composed of three layers, i.e. steel outer shell, lead shield, and stainless steel inner shell. The models were heated in an oil furnace only from their side at 800 0 C and cooled in the furnace. During the experiment, the temperature in the furnace and of the models were recorded continuously. The lead shield of the models started to melt 5--7 min after the start of heating. The temperature difference between the outer shell and the lead shield of the models was larger in case of the model without lead lining treatment than the models with it, and it is attributable to the low heat conductivity of the gap between the outer shell and the lead shield. The heat transfer property of casks was affected by the fabricating method of the casks. The temperature at the outer shell and that at the lead shield were calculated, and the results agreed considerably well with the experimental values, when 180 and 1800 kcal/m 2 h 0 C were employed as the heat conductivity of the gaps of the models. The gaps were estimated as 0.23 mm and 0.023 mm, respectively. In order to dissipate effectively the heat generated by contained fuel, lead lining treatment is necessary before pouring molten lead for shielding, but when the casks with the lead lining treatment are exposed to fire, the lead shield cannot keep its integrity. (Kako, I.)

  18. XSDRNPM-S biasing of MORSE-SGC/S shipping-cask calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Tang, J.S.

    1982-06-01

    This report describes implementation of a systematic approach for biasing a Monte Carlo radiation transport calculation. In particular, the adjoint fluxes from a one-dimensional discrete ordinates calculation with the XSDRNPM-S code are used to generate biasing parameters for the multigroup Monte Carlo code, MORSE-SGC/S. Application of this biasing procedure to several deep penetration spent fuel shipping cask problems is also reported. The results obtained for neutron and gamma-ray transport indicate that relatively inexpensive Monte Carlo calculations are possible for dry and water filled shipping cask problems using these procedures. 5 tables

  19. 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

  20. 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

  1. 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.

  2. Effectiveness of shield materials in the design of the PFBR irradiated fuel subassembly shipping cask

    International Nuclear Information System (INIS)

    Radhakrishnan, G.

    2003-01-01

    Fuel subassemblies are irradiated inside the reactor core till they achieve the required burn up and after that they are cooled to permissible decay power level in in-vessel and ex-vessel storage places. Subsequently they are transported to reprocessing plants by means of shipping casks. Shield for the shipping cask has to be designed such a way that it has to comply with the ICRP recommended dose levels of less than 2 mSv/h on contact at the outer surface of the cask and less than 100 mSv/h at 1 m distance from the outer surface of the cask. In this paper, shield design of a typical PFBR irradiated fuel subassembly, which can transport three subassemblies at a time, is narrated. Considering the neutron and fission product and induced gamma rays emitted by typical PFBR irradiated core central subassembly subjected to a maximum burn up, as the source term shield design optimizations have been done. One-dimensional discrete ordinates transport theory computer code ANISN and point kernel computer code QAD-CGGP have been used in complement to carry out the shield design optimizations. Cast-iron, carbon steel, stainless steel 304 and lead and permali have been considered for shield materials. Shield requirements on top, bottom and along the axial height of the shipping cask have also been estimated. (author)

  3. 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 US Nuclear Regulatory Commission (NRC) requires that any new material be qualified to meet the design and fabrication requirements of the ASME Boiler ampersand 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 US Department of Energy. The analysis approach used on some of the principal cask components is also presented. 5 refs., 8 figs., 3 tabs

  4. 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

  5. Safety assessment technology on the free drop impact and puncture analysis of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    Lee, Dew Hey; Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Ho Chul; Hong, Song Jin; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun

    2001-03-01

    In this study, the regulatory condition and analysis condition is analyzed for the free drop and puncture impact analysis to develop the safety assessment technology. Impact analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. LS-DYNA3D and ABAQUS is suitable for the free drop and the puncture impact 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 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. Safety assessment technology on the free drop impact and puncture analysis of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-15

    In this study, the regulatory condition and analysis condition is analyzed for the free drop and puncture impact analysis to develop the safety assessment technology. Impact analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. LS-DYNA3D and ABAQUS is suitable for the free drop and the puncture impact 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 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.

  7. Shipping cask demand associated with United States Government storage of commercial spent fuel

    International Nuclear Information System (INIS)

    Daling, P.M.; Engel, R.L.

    1983-05-01

    There were primarily two objectives of this study. The first was to develop estimates of the shipping cask fleet size that will be needed in the United States in the near future. These estimates were compared with current US spent fuel cask fleet size to determine its adequacy to provide the transportation services. The second objective was to develop estimates of the transportation costs associated with future movements of spent fuel. The results of this study were based on assumptions that were made prior to passage of the Nuclear Waste Policy Act of 1982 which authorizes the Department of Energy (DOE) to provide Federal Interim Storage of spent fuel from commercial reactors. The Act requires that the DOE is responsible for transportation of the fuel, although private industry is to provide these services. This paper examined the impacts of various spent fuel management strategies on spent fuel transportation hardware requirements and transportation costs. Conclusions related to optimization of the spent fuel transportation system can be drawn from the results of this study. The conclusions can be affected by changing the given set of assumptions used in this analysis. 3 tables

  8. 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

  9. Analysis of DCI cask drop test onto reinforced concrete pad

    International Nuclear Information System (INIS)

    Ito, C.; Kato, Y.; Hattori, S.; Shirai, K.; Misumi, M.; Ozaki, S.

    1993-01-01

    In a cask-storage facility, a cask may be subjected to an impact load as a result of a free drop onto the floor because of cask mishandling. We performed drop tests of casks onto a reinforced concrete (RC) slab representing the floor of a facility as well as simulation analysis [Kato et al]. This paper describes the details of the FEM analysis and calculated results and compares them with the drop test results. (J.P.N.)

  10. A status report on the development and certification of the Beneficial Uses Shipping System (BUSS) cask

    International Nuclear Information System (INIS)

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

    1996-01-01

    In the early 1980s, the US Department of Energy (DOE) implemented a program to encourage beneficial uses of nuclear byproduct materials, such as cesium-137 and strontium-90, created during the production of defense materials. Potential uses of the cesium-137 ( 137 CS) isotope included sterilizing medical products, maintaining the quality of certain food products, and disinfecting municipal sewage sludge. Strontium-90 ( 90 Sr) is a good heat source and has been used in thermoelectric generators and other products that require a constant supply of heat. During that same period, a proposed facility in Albuquerque, New Mexico, was designed to use cesium-137 to sterilize sewage sludge. To support the sewage sludge treatment facility, Sandia National Laboratories was funded by the DOE to develop a Nuclear Regulatory Commission (NRC)-certified Type B shipping container to transport cesium chloride (CsCl) or strontium fluoride (SrF 2 ) capsules produced by the Hanford Waste Encapsulation and Storage Facility (WESF) in the State of Washington. The primary purpose of the Beneficial Uses Shipping System (BUSS) cask is to provide shielding and confinement, as well as impact, puncture, and thermal protection for certified, special form contents during transport under normal and hypothetical accident conditions. The BUSS cask was designed to meet dimensional and weight constraints of the WESF and user facilities. Attaining as-low-as-reasonably-achievable (ALARA) radiation exposures in the design and operation of the transport system was a major design goal. Another goal was to obtain regulatory approval of the design by preparing a safety analysis report for packaging (SARP) (Yoshimura et al. 1993)

  11. Estimate of the crud contribution to shipping cask containment requirements

    International Nuclear Information System (INIS)

    Sandoval, R.P.; Einziger, R.E.; Jordan, H.; Malinauskas, A.P.; Mings, W.J.

    1992-01-01

    This paper reports that a methodology is developed to relate U.S. Code of Federal Regulations, Title 10, Part 71 (10CFR71) containment requirements to leak rates for the special case in which the only radioactive species having a potential for escape form the cask is that associated with debris (crud) contained on the fuel assemblies being transported. The methodology accounts for the characteristics of the crud and for attenuation of the gas-borne crud particulates once they become suspended within the cask. Calculations are performed for typical spent-fuel transport cask geometries and the normal and accident conditions prescribed in 10CFR71. The most current published data are used for crud composition and structure, specific activity, spallation mechanics and fractions, and crud particle size. The containment criteria leak rates are calculated assuming 5-yr-old spent fuel. In each accident case, the containment leak rate criteria are well in excess of 10 cm 3 /s. Under normal conditions of transport, the regulatory containment requirements are met by leak rates ranging from 1.5 x 10 -3 cm 3 /s to 1.5 x 10 -4 cm 3 /s for the transport of boiling water reactor fuel assemblies and form 1.8 x 10 -2 cm 3 /s to 1.3 x 10 -3 cm 3 /s for pressurized water reactor fuel assemblies. The calculated leak rates are most sensitive to the cask design, type of fuel, and particle size distribution. Conservatism of the limiting leak rates is discussed

  12. GA-4 and GA-9 legal weight truck shipping cask development

    International Nuclear Information System (INIS)

    Grenier, R.; Meyer, R.; Jensen, M.

    1989-01-01

    General Atomics (GA), under contract to the Idaho Operations Office of the U.S. Department of Energy, is 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 for the Office of Civilian Radioactive Waste Management (OCRWM) to meet its mission to dispose of nuclear wastes at a permanent disposal site. This paper discusses the primary goal, to maximize the number of fuel elements of each fuel type that a LWT cask can carry, while ensuring that the design meets all NRC licensing requirements

  13. Neutron multiplication and shielding problems in pressurized water reactor spent fuel shipping casks

    International Nuclear Information System (INIS)

    Devillers, C.; Blum, P.

    1977-01-01

    To evaluate the degree of accuracy of computational methods used in the shield design of spent fuel shipping casks, comparisons have been made between biological dose-rate calculations and measurements at the surface of a cask carrying three pressurized water reactor fuel assemblies. Neutron dose-rate measurements made with the fuel-carrying region successively wet and dry are also used to derive an experimental value of the k/sub eff/ of the wet fuel assemblies. Results obtained by this method are shown to be consistent with criticality calculations, taking into account fuel depletion

  14. A fuel response model for the design of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Duffey, T.A.; Einziger, R.E.; Hobbins, R.R.; Jordon, H.; Rashid, Y.R.; Barrett, P.R.; Sanders, T.L.

    1989-01-01

    The radiological source terms pertinent to spent fuel shipping cask safety assessments are of three distinct origins. One of these concerns residual contamination within the cask due to handling operations and previous shipments. A second is associated with debris (''crud'') that had been deposited on the fuel rods in the course of reactor operation, and a third involves the radioactive material contained within the rods. Although the lattermost source of radiotoxic material overwhelms the others in terms of inventory, its release into the shipping cask, and thence into the biosphere, requires the breach of an additional release barrier, viz., the fuel rod cladding. Hence, except for the special case involving the transport of fuel rods containing previously breached claddings, considerations of the source terms due to material contained in the fuel rods are complicated by the need to address the likelihood of fuel cladding failure during transport. The purpose of this report is to describe a methodology for estimating the shipping cask source terms contribution due to radioactive material contained within the spent fuel rods. Thus, the probability of fuel cladding failure as well as radioactivity release is addressed. 8 refs., 2 tabs

  15. 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

  16. Study on the evaluation method of radiation dose rate around spent fuel shipping casks

    International Nuclear Information System (INIS)

    Yamakoshi, Hisao

    1986-01-01

    This study aims at developing a simple calculation method which can evaluate radiation dose rate around casks with high accuracy in a short time. The method is based on a concept of the radiation shielding characteristics of cask walls. The concept was introduced to replace for ordinary radiation shielding calculation which requires a long calculation time and a large memory capacity of a computer in the matrix calculation. For the purpose of verifying the accuracy and reliability of the new method, it was applied to the analysis of the dose rate distribution around actual casks, which had been measured. The results of the analysis revealed that the newly proposed method was excellent for the forecast of radiation dose rate distribution around casks in view of the accuracy and calculation time. The short calculation time and high accuracy by the proposed method were attained by dividing the whole procedure of ordinary fine radiation shielding calculation into the calculation of radiation dose rate on a cask surface by the matrix expression of the characteristic function and the calculation of dose rate distribution using the simple analytical expression of dose rate distribution around casks. The effect of the heterogeneous array of spent fuel in different burnup state on dose rate distribution around casks was evaluated by this method. (Kako, I.)

  17. 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

  18. Status of radiation shield design for liquid metal fast breeder reactor spent fuel shipping cask application

    International Nuclear Information System (INIS)

    Dupree, S.A.; Rack, H.J.

    1976-09-01

    Neutron and gamma-ray transport calculations in one-dimensional cylindrical geometry have been performed on a trial reference LMFBR spent-fuel shipping cask that could transport one CRBR subassembly. In the study it was assumed that a layer of depleted U and a layer of neutron shielding materials were sandwiched between 5.08-cm-thick (2-in.) layers of stainless steel. The thicknesses of the internal layers were adjusted until a balanced dose rate (50 percent neuton and 50 percent gamma-ray) of 5 mrem/hr was achieved at a point 1.83 m (6 ft) from the cask surface. Neutron-shield materials considered were LiH, Be, B 4 C, DiH 2 . 5 , and C (graphite). Of these materials, LiH provided the smallest, lightest, and least expensive cask; however, its use would be contigent on expansion of production facilities for LiH and development of a canning or cladding procedure. The B 4 C shielded cask would offer the best alternative if the designs were limited to those using currently available materials

  19. 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

  20. Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

    1976-01-01

    A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

  1. 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

  2. Neutron multiplication and shielding problems in PWR spent-fuel shipping casks

    International Nuclear Information System (INIS)

    Devillers, C.

    1976-01-01

    In order to evaluate the degree of accuracy of computational methods used for the shield design of spent-fuel shipping casks, comparisons were made between biological dose rate calculations and measurements at the surface of a cask carrying three PWR fuel assemblies (the fuel being successively wet and dry). The experimental methods used provide ksub(eff) with an accuracy of 0.024. Neutron multiplication coefficients provided by the APOLLO and DOT-3 codes are located within the uncertainty range of the experimentally derived values. The APOLLO plus DOT codes for neutron source calculations and ANISN plus DOT codes for neutron transmission calculations provide neutron dose rate predictions in agreement with measurements to within 10%. The PEPIN 76 code used for deriving fission product γ-rays and the point kernel code MERCURE 4 treating the γ-ray transmission give γ dose rate predictions that generally differ from measurements by less than 25%

  3. Assessment of LMFBR spent fuel shipping cask concepts for the CRBRP and the US conceptual design study

    International Nuclear Information System (INIS)

    Pope, R.B.; Ortman, J.M.; Eakes, R.G.; Leisher, W.B.; Dupree, S.A.

    1980-01-01

    Study of conceptual shipping systems for CRBRP and CDS spent fuel has shown that systems significantly different from those used for LWR spent fuel will be required. In the conceptual design, liquid sodium was assumed to be the coolant in canisters containing the spent fuel assemblies, and multiple levels of containment were provided by canisters, an inner cask lid and an outer cask lid. Cask cooling at the reactor site during loading, and cooldown at the receiving site prior to unloading are significant but tractable problems

  4. Automated-biasing approach to Monte Carlo shipping-cask calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Tang, J.S.; Parks, C.V.; Childs, R.L.

    1982-01-01

    Computer Sciences at Oak Ridge National Laboratory, under a contract with the Nuclear Regulatory Commission, has developed the SCALE system for performing standardized criticality, shielding, and heat transfer analyses of nuclear systems. During the early phase of shielding development in SCALE, it was established that Monte Carlo calculations of radiation levels exterior to a spent fuel shipping cask would be extremely expensive. This cost can be substantially reduced by proper biasing of the Monte Carlo histories. The purpose of this study is to develop and test an automated biasing procedure for the MORSE-SGC/S module of the SCALE system

  5. Developmental testing of partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Allen, G.C.; Rack, H.J.; Joseph, B.J.; Dupree, S.A.

    1980-01-01

    Results of the phase one tests have demonstrated that the neutron-shielding concept described in this paper is a viable design option for spent fuel shipping casks. The tests have shown that the Boro-silicone 236 shield is superior to the other shield materials considered. Repeated TGA, aging and fire tests demonstrated the reliability of the data. A second phase of the test program is now being pursued where the Boro-silicone 236 is injected into all-steel slab sections, and cured in place. 5 tables

  6. Monte Carlo simulation of radiation streaming from a radioactive material shipping cask

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Schwarz, R.A.; Tang, J.S.

    1996-01-01

    Simulated detection of gamma radiation streaming from a radioactive material shipping cask have been performed with the Monte Carlo codes MCNP4A and MORSE-SGC/S. Despite inherent difficulties in simulating deep penetration of radiation and streaming, the simulations have yielded results that agree within one order of magnitude with the radiation survey data, with reasonable statistics. These simulations have also provided insight into modeling radiation detection, notably on location and orientation of the radiation detector with respect to photon streaming paths, and on techniques used to reduce variance in the Monte Carlo calculations. 13 refs., 4 figs., 2 tabs

  7. 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)

  8. Actinide partitioning-transmutation program. V. Preconceptual designs and costs of partitioning facilities and shipping casks, Appendix 4. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    This Appendix contains cost estimate documents for the Fuels Fabrication Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contributing to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed. Shipping cask costs are provided.

  9. Research and development of spent-fuel shipping casks and the criteria for sea-going vessels carrying them

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Since the transport of spent fuel will increase rapidly and extensively in the near future, the Japanese Atomic Energy Committee enacted the Technical Standard for Transportation of Radioactive Materials, based on the IAEA Regulation for the Safe Transport of Radioactive Materials, 1973 Revised Edition. The authorities concerned have begun to review the former ordinances for transporting radioactive materials and to develop a unified system of relevant laws and standards. For ten years the Atomic Energy Bureau has invested in research and development to obtain data for the design and licence of a spent-fuel shipping cask. Different scale models of a prototype weighing 80t were used to clarify the scale effect of drop, puncture and fire tests, which are a feature of Japanese research and development. Also an immersion test in water at pressures up to about 500 bar is now carried out to investigate the integrity of the cask body and sealing structure to prevent leakage of radioactive contents to the surroundings should the cask fall into deep sea. In Japan, depending on the site of nuclear plants, almost all transport of unirradiated and spent fuels is by sea. Therefore, to secure safe transport, the design criteria of ships for the exclusive transport of spent-fuel shipping casks, namely full-load shipping, have been enacted, which aim to make the likelihood of sinking on collision, stranding, and other unforeseen accidents at sea highly improbable and also to keep radiation exposure of the crew as low as possible. (author)

  10. Actinide partitioning-transmutation program. V. Preconceptual designs and costs of partitioning facilities and shipping casks, Appendix 4. Final report

    International Nuclear Information System (INIS)

    1980-06-01

    This Appendix contains cost estimate documents for the Fuels Fabrication Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contributing to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed. Shipping cask costs are provided

  11. Value/impact of design criteria for cast ductile iron shipping casks

    International Nuclear Information System (INIS)

    1983-01-01

    The ductile failure criteria proposed in the Base report appear appropriate except that stress intensity values, S/sub m/ should be based on lower safety factors and ductility should be added as a criterion. A safety factor for stress intensity, s/sub m/ of 4 is recommended rather than 3 on minimum ultimate tensile strength, S/sub u/ in accordance with ASME code philosophy of assigning higher safety factors to cast ductile iron than to steel. This more conservative approach has no impact on costs since the selection of wall thickness is controlled by shielding rather than by stress considerations. The addition of a ductility criterion is recommended because of the problems associated with the selection of appropriate brittle failure criteria and the potential for cast ductile iron to have extremely low elongation at failure. Neither a materials nor a linear elastic fracture mechanics (LEFM) approach appear to be viable for demonstrating the prevention of brittle failure in cast ductile iron shipping casks. It is possible that the analytic methods predict brittle failure because of extremely conservative assumptions whereas real casks may not fail. Model drop tests could be used to demonstrate containment integrity. It is estimated that a risk committment of at least $1,000,000 would be required for engineering, design, model fabrication and testing. Before taking such risks, a mechanism should be found to obtain concurrence from NRC that the results of the test would be acceptable. Probabilistic approaches or model testing could be used to demonstrate the acceptability of cast ductile iron casks from a brittle failure point of view. Before probabilistic methods can be used, the NRC would have to be persuaded to accept the approach of the Competent Authority in West Germany or more formalized methods for probabilistic risk assessments

  12. 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.)

  13. 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

  14. 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

  15. 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.)

  16. SAVIT: a dymanic model to predict vibratory motion within a spent fuel shipping cask; rail car system

    International Nuclear Information System (INIS)

    Fields, S.R.

    1978-03-01

    A dynamic model of a spent fuel shipping cask-rail car system has been developed to provide estimates of the vibratory motion of LWR spent fuel assemblies during transport and to estimate the effects of this motion on the condition of the assemblies when they arrive at receiving and storage facilities. Results of preliminary test computations are presented to illustrate the capabilities of the model

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ju-Chan; Bang, K.S.; Yu, S.H.; Cho, S.S.; Choi, W.S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    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.

  19. 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

  20. Thermal-hydraulic software development for nuclear waste transportation cask design and analysis

    International Nuclear Information System (INIS)

    Brown, N.N.; Burns, S.P.; Gianoulakis, S.E.; Klein, D.E.

    1991-01-01

    This paper describes the development of a state-of-the-art thermal-hydraulic software package intended for spent fuel and high-level nuclear waste transportation cask design and analysis. The objectives of this software development effort are threefold: (1) to take advantage of advancements in computer hardware and software to provide a more efficient user interface, (2) to provide a tool for reducing inefficient conservatism in spent fuel and high-level waste shipping cask design by including convection as well as conduction and radiation heat transfer modeling capabilities, and (3) to provide a thermal-hydraulic analysis package which is developed under a rigorous quality assurance program established at Sandia National Laboratories. 20 refs., 5 figs., 2 tabs

  1. End effects in the criticality analysis of burnup credit casks

    International Nuclear Information System (INIS)

    Brady, M.C.; Parks, C.V.

    1990-01-01

    A study to evaluate the effect of axially dependent burnup on k eff has been performed as part of an effort to qualify procedures to be used in establishing burnup credit in shipping cask design and certification. This study was performed using a generic 31-element modular cast-iron cask (wall thickness 33.1 cm) with a 1-cm-thick borated stainless-steel basket for reactivity control. Fuel isotopics used here are those of the 17 x 17 Westinghouse assemblies from the North Anna Unit 1 reactor. Virginia Power (VP) provided detailed spatial isotopics for the fuel assemblies in-core at beginning-of-cycle 5 (BOC-5) as generated from their PDQ analyses. Twenty-two axial planes were defined in the original VP data. The isotopics used in this study were for a 3.41 initial wt % 235 U and an average burnup of 31.5 GWd/MTU

  2. 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

  3. 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

  4. 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.

  5. Test report for PAS-1 cask certification for shipping payload B

    International Nuclear Information System (INIS)

    MERCADO, J.E.

    1998-01-01

    This test report documents the successful inspection and testing to certify two NuPac PAS-1 casks in accordance with US Department of Energy Certificate of Compliance (CoC) USA/9184/B(U). The primary and secondary containment vessels of each cask met the acceptance criteria defined in the CoC and the test plan

  6. 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

  7. 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

  8. 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

  9. Extension of ship accident analysis to multiple-package shipments

    International Nuclear Information System (INIS)

    Mills, G.S.; Neuhauser, K.S.

    1997-11-01

    Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously. Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence

  10. Extension of ship accident analysis to multiple-package shipments

    International Nuclear Information System (INIS)

    Mills, G.S.; Neuhauser, K.S.

    1998-01-01

    Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously (Spring, 1995). Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis (Spring, 1995) involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well-characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence. (authors)

  11. Radiation resistant and decontaminable coatings for shipping, interim storage and repository storage casks containing radioactive wastes

    International Nuclear Information System (INIS)

    Kunze, S.

    1995-02-01

    All the Corrobesch-DF-Nukelar coatings - black, yellow, blue, red and white - have been excellently decontaminable without and after radiation exposure with 3x10 5 Gy, despite the slightly higher absorbed dose rate applied at KFA Juelich (DIN 55 991 requires ≤1.0 KGy/h). After a further increase to 3x10 6 Gy in the absorbed dose, with an absorbed dose rate up to 1.0 KGy/h conforming to the standard, the coatings black, yellow, blue were still excellent in their decontamination behavior. After exposure to 10 7 Gy all coatings irradiated at Gammaster in their irradiation room (150 m 3 ) with permanent air changes and at absorbed dose rates of 0.9-1.0 KGy/h have been well decontaminable, and the coatings irradiated at KFA Juelich in the 10 l vessel with discontinuous air changes and variable absorbed dose rate (0.22-2.7 KGy/h) have still been fairly well decontaminable only. To be able to evaluate possible changes occurring upon 10 7 Gy radiation exposure, the test specimens were exposed to the action of chemicals according to DIN 55 991 as well as to decontamination cleansing solutions. Different discolorations, very small reductions in brilliancy, and sometimes minor deteriorations in surface hardness occurred. Detrimental visible changes, e.g. bubble and crack formation, swelling, detachment from the base, etc., have not been found for any of the coatings. These results for the test specimens irradiated at Gammaster are identical with the results for the test specimens irradiated at KFA Juelich, except minor deviations. Contrary to expectations, Corrobesch-DF-Nuklear has proved to be a coating material, which, although it consists of organic base material, nevertheless tolerates radiation exposures without visible damage, i.e. conditions under which only electrodeposited nickel coatings have appeared appropriate until now. This means that application of Corrobesch-Nuklear-DF allows the costs of coating of fuel element shipping and storage casks to be reduced

  12. Effect of fission gas leakage on heat transfer within a helium filled spent fuel shipping cask

    International Nuclear Information System (INIS)

    Pope, R.B.; Schimmel, W.P. Jr.

    1978-01-01

    Leakage of Xe from spent fuel elements into a He-filled cask would reduce the thermal conductivity, but it would also increase the nondimensional Grashof and Rayleigh number convection parameters. The thermal performance for various quantities of leaked fission gases was evaluated for a cask containing 9 fuel assemblies, each producing approximately 2 kW, and a He partial pressure of 4 atm. If all pins leaked, the max pin temperature would increase from 761 to 839 0 K. It is concluded that the effect of fission gases is a second order effect

  13. Development of a toroidal shell-type shock absorber for an irradiated fuel shipping cask

    International Nuclear Information System (INIS)

    Sugita, Y.; Mochizuki, S.

    1983-01-01

    This study described the design method of a toroidal shell-type shock absorber and the dynamic responses of the cask body, the internal structure and water when this shock absorber was used. Conclusions are: the calculated results on the basis of the master curves of non-dimensionalized force-deflection relations by static compression tests show a close agreement with the experimental results; the internal structure moves together with the cask body in every position; and the maximum water pressure is larger by a factor of 1.2 than the static pressure multiplied by the maximum deceleration in every direction due to the low-frequency wave propagation

  14. Expansion of the capabilities of the GA-4 legal weight truck spent fuel shipping cask

    International Nuclear Information System (INIS)

    Zimmer, A.; Razvi, J.; Johnson, L.; Welch, B.; Lancaster, D.

    2004-01-01

    General Atomics (GA) has developed the Model GA-4 Legal Weight Truck Spent Fuel Cask, a high capacity cask for the transport of four PWR spent fuel assemblies, and obtained a Certificate of Compliance (CoC No. 9226) in 1998 from the US Nuclear Regulatory Commission (NRC). The currently authorized contents in this CoC however, are much more limiting than the actual capability of the GA-4 cask to transport spent PWR fuel assemblies. The purpose of this paper is to show how the authorized contents can be significantly expanded by additional analyses without any changes to the physical design of the package. Using burnup credit per ISG-8 Rev. 2, the authorized contents can be significantly expanded by increasing the maximum enrichment as the burnup increases. Use of burnup credit eliminates much of the criticality imposed limits on authorized package contents, but shielding still limits the use of the cask for the higher burnup, short cooled fuel. By downloading to two assemblies and using shielding inserts, even the high burnup fuel with reasonable cooling times can be transported

  15. Modal analysis of spent fuel cask for WWER-1000 reactors

    International Nuclear Information System (INIS)

    Azimfar, S. A.; Kazemi, A.

    2011-01-01

    The Spent Fuel Assemblies of WWER-1000 reactors are planned to be transported by special containers which are supposed to be designed in a manner to stand against vibrations and impacts in order to protect the spent fuel from any possible damage. The vibration opposition of these containers shall be far beyond the critical resonance, because the resonances about the natural frequency of the structure will cause the enhancement of its oscillation range and may end with its disintegration. Determination of the amounts of natural frequencies and their mode shape can be achieved by vibration analyzing methods. The amount of the natural frequency of any structure crucially depends on its shape, material and lean points as well as the amount of the loads and the type of these loads. Due to the fact that the Spent Fuel Casks used for transportation in nuclear power plants in Russian Federation are TK-13 type and the pieces of information released are negligible, the scientists in Russia are working on the design and analysis of a new type made up of composite Material. In the presented paper the cask of spent fuel of TK-13 is modeled by ANSYS at 10.0 and ten natural frequency modes have been calculated, followed by the comparison of this result with the composite cask.

  16. Second Annual Maintenance, Inspection, and Test Report for PAS-1 Cask Certification for Shipping Payload B

    International Nuclear Information System (INIS)

    KELLY, D.J.

    2000-01-01

    The Nuclear Packaging, Inc. (NuPac), PAS-1 cask is required to undergo annual maintenance and inspections to retain certification in accordance with U.S. Department of Energy (DOE) Certificate of Compliance USA/9184B(U) (Appendix A). The packaging configuration being tested and maintained is the NuPac PAS-1 cask for Payload B. The intent of the maintenance and inspections is to ensure the packaging remains in unimpaired physical condition. Two casks, serial numbers 2162-026 and 2162-027, were maintained, inspected, and tested at the 306E Development, Fabrication, and Test Laboratory, located at the Hanford Site's 300 Area. Waste Management Federal Services, Inc. (WMFS), a subsidiary of GTS Duratek, was in charge of the maintenance and testing. Cogema Engineering Corporation (Cogema) directed the operations in the test facility. The maintenance, testing, and inspections were conducted successfully with both PAS-1 casks. The work conducted on the overpacks included weighing, gasket replacement, and plastic pipe plug and foam inspections. The work conducted on the secondary containment vessel (SCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, and a helium leak test. The work conducted on the primary containment vessel (PCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, dimensional inspection of the vessel bottom, a helium leak test, and dye penetrant inspection of the welds. The vermiculite material used in the cask rack assembly was replaced

  17. 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)

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

    International Nuclear Information System (INIS)

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

    1992-04-01

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

  19. 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

  20. 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.

  1. The TN-RAM - a new cask for shipping high activity irradiated hardware

    International Nuclear Information System (INIS)

    Neider, T.; Hanson, A.S.

    1993-01-01

    Transnuclear, Inc. has developed a Type B(U) radioactive material transport packaging designed specifically for the transport of dry, irradiated, non-fuel bearing components (NFBC). The TN-RAM is a transport cask in the configuration of a right circular cylinder, fabricated from lead and stainless steel, with wood-filled impact limiters attached at both ends. The lead and stainless steel construction of the lid, walls, and bottom provides a shielding effectiveness of approximately 7.1 inches (18 cm) lead equivalent. (J.P.N.)

  2. Safety analysis of casks under extreme impact conditions

    International Nuclear Information System (INIS)

    Wieser, G.; Qiao, L.; Voelzke, H.; Wolff, D.; Droste, B.

    2004-01-01

    The determination of the inherent safety of casks under extreme impact conditions has been of increasing interest since the terrorist attacks of 11 September 2001. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. One of the most critical scenarios for a cask is the centric impact of a dynamic load onto the lid-seal system. This can be caused, for example, by a direct aircraft crash (or just its engine) as well as by an impact due to the collapse of a building, e.g. a nuclear facility storage hall. In this context BAM is developing methods to calculate the deformation of cask components and-with respect to leak-tightness-relative displacements between the metallic seals and their counterparts. This paper presents reflections on modelling of cask structures for finite-element analyses and discusses calculated results of stresses and deformations. Another important aspect is the behaviour of a cask under a lateral impact by aircraft or fragments of a building. Examples of the kinetic reaction (cask acceleration due to the fragments, subsequent contact with neighbouring structures like the ground, buildings or casks) are shown and discussed in correlation to cask stresses which are to be expected. (author)

  3. Safety analysis of casks under extreme impact conditions

    International Nuclear Information System (INIS)

    Wieser, G.; Qiao Linan; Voelzke, H.; Wolff, D.; Droste, B.

    2004-01-01

    The determination of the inherent safety of casks also under extreme impact conditions has been of increasing interest since the terrorist attacks from 11th September 2001. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. One of the most critical scenarios for a cask is the centric impact of a dynamic load onto the lid seal system. This can be caused e.g. by direct aircraft crash or its engine as well as by an impact due to the collapse of a building e.g. a nuclear facility storage hall. In this context BAM is developing methods to calculate the deformation of cask components and - with respect to leak tightness - relative displacements between the metallic seals and their counterparts. This paper presents reflections on modelling of cask structures for Finite Element analyses and discusses calculated results of stresses and deformations. Another important aspect is the behaviour of a cask under a lateral impact by aircraft and fragments of a building. Examples of the kinetic reaction (cask acceleration due to the fragments, subsequent contact with neighbouring structures like ground, buildings or casks) are shown and discussed in correlation to cask stresses which are to be expected

  4. 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.)

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

    International Nuclear Information System (INIS)

    Noh, Jae Soo; Park, Younwon; Song, Sub Lee; Kim, Hyeun Min

    2016-01-01

    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

  6. 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.

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

    International Nuclear Information System (INIS)

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

    1977-11-01

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

  8. 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

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

    1992-01-01

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

  10. Evaluation of the 252Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1987-01-01

    The 252 Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt LWR fuel submerged in fuel storage pools, fully loaded and as they are being loaded. The motivation for this evaluation was that measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This in turn could lead to more cost-effective cask designs. Evaluation of the method for this application was based on (1) experiments already completed at a critical experiments facility using arrays of PWR fuel pins typical of the size of storage cask configurations, (2) the existence of neutron detectors that can function in shipping cask environments, and (3) the ability to construct ionization chambers containing 252 Cf of adequate intensity for these measurements. These three considerations are discussed

  11. Spreader beam analysis for the CASTOR GSF cask

    International Nuclear Information System (INIS)

    Clements, E.P.

    1997-01-01

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

  12. Experiment and analysis of CASTOR type model cask for verification of radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    Hattori, Seiichi; Ueki, Kohtaro.

    1988-08-01

    The radiation shielding system of CASTOR type cask is composed of the graphite cast iron and the polyethylene lod. The former fomes the cylndrical body of the cask to shield gamma rays and the latter is embeded in the body to shield neutrons. Characteristic of radiation shielding of CASTOR type cask is that zigzag arrangement of the polyethylene lod is adopted to unify the penetrating dose rate. It is necessary to use the three-dimensional analysis code to analyse the shielding performance of the cask with the complicated shielding system precisely. However, it takes too much time as well as too much cost. Therefore, the two-dimensional analysis is usually applied, in which the three-dimensional model is equivalently transformed into the two-dimensional calculation. The reseach study was conducted to verify the application of the two-dimensional analysis, in which the experiment and the analysis using CASTOR type model cask was perfomed. The model cask was manufactured by GNS campany in West Germany and the shielding ability test facilities in CRIEPI were used. It was judged from the study that the two-dimensional analysis is useful means for the practical use.

  13. Total Analysis System for Ship Structural Strength

    OpenAIRE

    Takuya, Yoneya; Hiroyuki, Kobayashi; Abdul M., Rahim; Yoshimichi, Sasaki; Masaki, Irisawa; Technical Investigation and Information Department, Research Center; Technical Investigation and Information Department, Research Center; Singapore Office; Technical Investigation and Information Department, Research Center; Technical Investigation and Information Department, Research Center

    2001-01-01

    This paper outlines a total analysis system for ship hull structures, which integrates a wide variety of analysis functions to realise practical applications of rational methods for assessing ship structural strength. It is based on direct calculation of wave-induced loads as well as three-dimensional structural analysis of an entire-ship or hold structure. Three major analysis functions of the total system are ship motion and wave load analysis, ship structural analysis and statistical analy...

  14. Spent-fuel shipping and cask-handling studies in wet and dry environments. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    McCreery, P.N.

    1982-09-01

    A demonstration cask system has been constructed specifically to be used in examining unconventional techniques in handling spent fuel and fuel-hauling casks. This report demonstrates, through a series of photographs, some of these techniques and discusses others. It includes wet and dry operations, loading and unloading horizontally and vertically, mobile on-site carriers that can eliminate the need for some cranes and, in general, many of the operational options that are open in the design of future fuel handling systems

  15. Study and full-scale test of a high-velocity grade-crossing simulated accident of a locomotive and a nuclear-spent-fuel shipping cask

    International Nuclear Information System (INIS)

    Huerta, M.; Yoshimura, H.R.

    1983-02-01

    This report described structural analyses of a high-speed impact between a locomotive and a tractor-trailer system carrying a nuclear-spent-fuel shipping cask. The analyses included both mathematical and physical scale-modeling of the system. The report then describes the full-scale test conducted as part of the program. The system response is described in detail, and a comparison is made between the analyses and the actual hardware response as observed in the full-scale test. 34 figures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  17. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    International Nuclear Information System (INIS)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu

    2015-01-01

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance

  18. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.

  19. Study of shielding analysis methods for casks of spent fuel and radioactive waste

    International Nuclear Information System (INIS)

    Saito, Ai

    2017-01-01

    Casks are used for storage or transport spent fuels or radioactive waste. Because high shielding performances are required, it is very important to confirm the validity of shielding analysis methods in order to evaluate cask shielding abilities appropriately. For this purpose, following studies were carried out. 1) A series of parameter survey for several codes to evaluated the difference of the results. 2) Calculations using the MCNP code are effective and theoretically have better accuracy. However setting reasonable variance reduction parameters is indispensable. Therefore, effectiveness of the ADVANTG code which produces automatically reasonable variance reduction parameters is carried out by comparison with conventional method. As a result, the validity of shielding analysis methods for casks is confirmed. The results will be taken into consideration in our future shielding analysis. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Goo, Junsung; Shin, Kwangbok [Hanbat Nat' l Univ., Daejeon (Korea, Republic of); Choi, Woosuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-11-15

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

  1. 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.

  2. Structural analysis of a metal spent-fuel storage cask in an aircraft crash for risk assessment

    International Nuclear Information System (INIS)

    Almomani, Belal; Lee, Sanghoon; Kang, Hyun Gook

    2016-01-01

    Highlights: • Several engine-applied loads with different locations of impact on the storage cask body were implemented. • Cask structural responses due to the influence of engine impact loadings were analyzed. • Leakage path areas from lid closure openings were numerically calculated. • Release fractions that depend on the generated seal opening areas and fuel damage ratios were estimated. - Abstract: Evaluations of the impact resistance of a dry storage cask under mechanical impact loadings resulting from a large commercial aircraft crash have become an important issue for designers and evaluators, in order to promote interim dry storage activities and to evaluate design safety margins. This study presents a method to evaluate the structural integrity of a generic metal cask subjected to various mechanical loading conditions, which represent aircraft engine impacts, on different locations of the cask body. Thirty representative impact conditions are analyzed to provide a comprehensive evaluation of cask damage response. The applied engine impact load–time functions were carefully re-derived by utilizing CRIEPI’s proposed curve through Riera’s approach for six impact velocities, and applied to five locations on a freestanding cask: lateral impacts on the lower half, center of gravity, and upper half of the cask body, corner impact on the lid closure, and vertical impact on the center of the lid closure. A nonlinear dynamic finite element analysis is performed to evaluate the dynamic response of the cask lid closure system and to calculate the lid gaps. The release fractions from the cask to the environment for each impact condition are preliminarily estimated by referring to a proposed methodology from literature. It is believed that this paper presents a systematic process to connect the mechanical analysis of a cask response at the moment of aircraft engine impact with its radiological consequence analysis.

  3. Structural analysis of a metal spent-fuel storage cask in an aircraft crash for risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    Almomani, Belal, E-mail: balmomani@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Sanghoon, E-mail: shlee1222@kmu.ac.kr [Department of Mechanical and Automotive Engineering, Keimyung University, Dalgubeol-daero 1095, Dalseo-gu, Daegu (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2016-11-15

    Highlights: • Several engine-applied loads with different locations of impact on the storage cask body were implemented. • Cask structural responses due to the influence of engine impact loadings were analyzed. • Leakage path areas from lid closure openings were numerically calculated. • Release fractions that depend on the generated seal opening areas and fuel damage ratios were estimated. - Abstract: Evaluations of the impact resistance of a dry storage cask under mechanical impact loadings resulting from a large commercial aircraft crash have become an important issue for designers and evaluators, in order to promote interim dry storage activities and to evaluate design safety margins. This study presents a method to evaluate the structural integrity of a generic metal cask subjected to various mechanical loading conditions, which represent aircraft engine impacts, on different locations of the cask body. Thirty representative impact conditions are analyzed to provide a comprehensive evaluation of cask damage response. The applied engine impact load–time functions were carefully re-derived by utilizing CRIEPI’s proposed curve through Riera’s approach for six impact velocities, and applied to five locations on a freestanding cask: lateral impacts on the lower half, center of gravity, and upper half of the cask body, corner impact on the lid closure, and vertical impact on the center of the lid closure. A nonlinear dynamic finite element analysis is performed to evaluate the dynamic response of the cask lid closure system and to calculate the lid gaps. The release fractions from the cask to the environment for each impact condition are preliminarily estimated by referring to a proposed methodology from literature. It is believed that this paper presents a systematic process to connect the mechanical analysis of a cask response at the moment of aircraft engine impact with its radiological consequence analysis.

  4. Standardized analyses of nuclear shipping containers

    International Nuclear Information System (INIS)

    Parks, C.V.; Hermann, O.W.; Petrie, L.M.; Hoffman, T.J.; Tang, J.S.; Landers, N.F.; Turner, W.D.

    1983-01-01

    This paper describes improved capabilities for analyses of nuclear fuel shipping containers within SCALE -- a modular code system for Standardized Computer Analyses for Licensing Evaluation. Criticality analysis improvements include the new KENO V, a code which contains an enhanced geometry package and a new control module which uses KENO V and allows a criticality search on optimum pitch (maximum k-effective) to be performed. The SAS2 sequence is a new shielding analysis module which couples fuel burnup, source term generation, and radial cask shielding. The SAS5 shielding sequence allows a multidimensional Monte Carlo analysis of a shipping cask with code generated biasing of the particle histories. The thermal analysis sequence (HTAS1) provides an easy-to-use tool for evaluating a shipping cask response to the accident capability of the SCALE system to provide the cask designer or evaluator with a computational system that provides the automated procedures and easy-to-understand input that leads to standarization

  5. 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

  6. Safety analysis of dual purpose metal cask subjected to impulsive loads due to aircraft engine crash

    International Nuclear Information System (INIS)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    2009-01-01

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters and seismic tests subjected to strong earthquake motions. Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001. This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine research (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are developed

  7. Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads due to Aircraft Engine Crash

    Science.gov (United States)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters(1) and seismic tests subjected to strong earthquake motions(2). Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001(3)-(6). This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are

  8. Analysis of ductile cast iron for spent fuel cask

    International Nuclear Information System (INIS)

    Sakurai, D.; Minami, M.

    1993-01-01

    143 data from 12 Heavy Section D.C.I. Cast Bodies of 6 manufacturer in Japan were investigated and statistically analyzed about Mechanical Properties, Metallurgical Conditions and Manufacturing Processes. The following results were concluded. (1) The Mechanical Properties of J.I.S. are reasonable, reliable and reasonably achievable. (2) The Mechanical Properties of D.C.I. are reasonably achievable. (3) The Mechanical Properties of D.C.I. are easily controllable through metallurgical method. (4) D.C.I. (JIS G5504-92) is applicable to the material for spent fuel cask. (J.P.N.)

  9. 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

  10. 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

  11. ROCKING. A computer program for seismic response analysis of radioactive materials transport AND/OR storage casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1995-11-01

    The computer program ROCKING has been developed for seismic response analysis, which includes rocking and sliding behavior, of radioactive materials transport and/or storage casks. Main features of ROCKING are as follows; (1) Cask is treated as a rigid body. (2) Rocking and sliding behavior are considered. (3) Impact forces are represented by the spring dashpot model located at impact points. (4) Friction force is calculated at interface between a cask and a floor. (5) Forces of wire ropes against tip-over work only as tensile loads. In the paper, the calculation model, the calculation equations, validity calculations and user's manual are shown. (author)

  12. User's manual of MANYCASK code for calculation of spatial distributions of radiation dose rates in a system composed of many spent-fuel-shipping casks

    International Nuclear Information System (INIS)

    Yamakoshi, Hisao

    1986-01-01

    A calculation code MANYCASK is designed for evaluation of spatial distributions of radiation dose rates in ships loaded with a lot of spent fuel shipping casks. Principle of the calculation method adopted in this code is different from that of ordinary codes, and is advantageous for calculating highly reliable dose rate distributions with a very short calculation time. Basic concept of the principle has been described in other reports in detail. A brief description of the principle will be included in the present report along with a technique named Shadow Technique in this report, in addition to format descriptions of output data as well as input data. Results of sample calculations are compared with measured results in figures so as to show how the calculation method adopted is valid. For the purpose of making this code popular among many people, the author writes the user's manual in the present report in Japanese for domestic users, and in English in another report for people in abroad. (author)

  13. Development of Neutron Shielding Material for Cask and Accelerator

    International Nuclear Information System (INIS)

    Kang, Hee Young; Seo, Ki Seog; Lee, Byung Chul; Park, Chang Jae; Kim, Ho Dong

    2008-01-01

    The neutron shielding materials are used as a neutron shield for spent fuel shipping cask, beam accelerators and neutron generators. At early stage, the neutron attenuations of materials were evaluated with the cross sections. After that, benchmark or mock-up experiments on the multi-layer problem to confirm the shielding characteristics or to evaluate analysis accuracy were reported. Recently, the need to transport spent nuclear fuels is increasing due to the current limited storage capacity. The on-site storage capacity at some of nuclear power plants is expected to be full in near future. With a growing inventory of spent fuels at power plants, these spent fuels need to be transported to other storage facilities. Shipping casks have been developed to safely transport spent fuels that emit high neutrons and gamma-ray radiation. The external radiation level of the shipping cask from the spent fuel must be limited to meet the standards specified by the IAEA radioactive material package regulation, so it is important to develop a proper neutron shielding material for a shipping cask. Neutron shielding experiments and analyses on the shielding effects of materials have been conducted, and some experiments have been performed to examine the shielding effects of selected materials. The shielding experiments consist of evaluating not only the shielding effects of a material alone but also the effects of the material thickness. The experimental results were compared with those obtained by using the MCNP-5c code

  14. FINCRUSH : a computer program for impact analysis of radioactive material transport cask with fins

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1997-05-01

    In drop impact analyses for radioactive material transport cask with cooling fins, relationship between fin plastic deformation and fin energy absorption is used. This relationship was obtained by ORNL experiments and MONSER Co. in Canada. Based on ORNL experiments, a computer program FINCRUSH has been developed for rapid safety analysis of cask drop impact to obtain the maximum impact acceleration and the maximum fin deformation. Main features of FINCRUSH are as follows: (1) annulus fins on a cylindrical shell and plate fins on a disk can be treated, (2) it is capable of graphical representations for calculation results and fin absorption energy data and (3) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of the FINCRUSH. In the paper, brief illustration of calculation method of FINCRUSH is presented. The second section presents comparisons between FINCRUSH and experimental results. The third section provides a use's guide for FINCRUSH. (author)

  15. FINCRUSH : a computer program for impact analysis of radioactive material transport cask with fins

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-05-01

    In drop impact analyses for radioactive material transport cask with cooling fins, relationship between fin plastic deformation and fin energy absorption is used. This relationship was obtained by ORNL experiments and MONSER Co. in Canada. Based on ORNL experiments, a computer program FINCRUSH has been developed for rapid safety analysis of cask drop impact to obtain the maximum impact acceleration and the maximum fin deformation. Main features of FINCRUSH are as follows: (1) annulus fins on a cylindrical shell and plate fins on a disk can be treated, (2) it is capable of graphical representations for calculation results and fin absorption energy data and (3) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of the FINCRUSH. In the paper, brief illustration of calculation method of FINCRUSH is presented. The second section presents comparisons between FINCRUSH and experimental results. The third section provides a use`s guide for FINCRUSH. (author)

  16. 78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System AGENCY: Nuclear...] Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No... Safety Analysis Report for the Standardized NUHOMS[supreg] Horizontal Modular Storage System for...

  17. Thermal analysis of the modified Hallum Nuclear Power Facility cask using experimentally obtained thermal boundary conditions corresponding to an engulfing open pool fire

    International Nuclear Information System (INIS)

    Longenbaugh, R.S.; Sanchez, L.C.; Gregory, J.J.

    1987-08-01

    This report presents the two-dimensional heat transfer analysis of an open pool fire surrounding a modified radioactive materials transport cask. The cask is an older cask that was used by the Hallum Nuclear Power Facility (HNPF). The HNPF cask did not have a neutron shielding region but was modified to include one for testing purposes. Analysis of the thermal effects of an engulfing open pool fire was performed with the use of the heat transfer code Q/TRAN, which had previously been used in thermal benchmarking problems for spent nuclear fuel casks. Boundary condition data for the analysis were derived from experimental open pool fire tests of large-scale calorimeter test articles performed at SNL that produced information about cask surface heat flux versus surface temperature relationships. Data analysis was directed toward a determination of the thermal response of the cask, particularly the extent of lead melt since lead is used within the HNPF cask's gamma-shielding region. Parameters, such as surface emissivity and internal heat generation rate, can affect the results of the thermal analysis which control the amount of lead melt. A parameter sensitivity analysis was performed using a one-dimensional model to describe how surface emissivity and internal heat generation rates affect the temperature distribution within the cask. The information from this analysis was used to determine the range of parameters for the two-dimensional thermal analysis. 13 refs., 57 figs., 8 tabs

  18. 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

  19. Full-scale tests of spent-nuclear-fuel shipping systems

    International Nuclear Information System (INIS)

    Yoshimura, H.R.; Huerta, M.

    1976-01-01

    Sandia Laboratories will be conducting, for the U.S. Energy Research and Development Administration, a series of tests involving spent-nuclear-fuel shipping systems. Large shipping casks in the 20500 to 70000-kg range will be included in the following full-scale tests: (1) Runaway tractor-trailer crash into a solid concrete barrier while carrying a shipping cask. (2) High-speed locomotive grade-crossing impact with a truck carrying a shipping cask. (3) High-speed derailment, collision, and fire involving a special railcar and shipping cask. The hardware and testing procedures for each of the tests are described. The analysis conducted in advance of the tests addresses the modelling technique used and a description of the scale-model tests. Analytical modelling being done before running the full-scale tests is also described. (author)

  20. Source storage and transfer cask: Users Guide

    International Nuclear Information System (INIS)

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

    1985-04-01

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

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

    International Nuclear Information System (INIS)

    Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

    2002-01-01

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

  2. Design lead shielded casks for shipment and spent fuel from power reactors to reprocessing plant at Tarapur

    International Nuclear Information System (INIS)

    Seetharamaiah, P.

    1975-01-01

    Spent fuels from the Tarapur and Rajasthan Atomic Power Stations (TAPS and RAPS) are shipped to Fuel Reprocessing Plant at Tarapur in heavily lead shielded casks weighing about 65 tonnes as they are highly radioactive. The design of the casks has to meet stringemt requirements of safety and the integrity should be ensured to contain activity under credible accidents during handling and transportation. The paper presents the design of two casks for TAPS and RAPS spent fuel transportation particularly with reference to stress analysis considerations. The analysis also includes the handling gadgets and tie down attachments on the rail wagon and road trailer. (author)

  3. Seismic Response Analysis and Test of 1/8 Scale Model for a Spent Fuel Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Park, C. G.; Koo, G. H.; Seo, G. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yeom, S. H. [Chungnam Univ., Daejeon (Korea, Republic of); Choi, B. I.; Cho, Y. D. [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2005-07-15

    The seismic response tests of a spent fuel dry storage cask model of 1/8 scale are performed for an typical 1940 El-centro and Kobe earthquakes. This report firstly focuses on the data generation by seismic response tests of a free standing storage cask model to check the overturing possibility of a storage cask and the slipping displacement on concrete slab bed. The variations in seismic load magnitude and cask/bed interface friction are considered in tests. The test results show that the model gives an overturning response for an extreme condition only. A FEM model is built for the test model of 1/8 scale spent fuel dry storage cask using available 3D contact conditions in ABAQUS/Explicit. Input load for this analysis is El-centro earthquake, and the friction coefficients are obtained from the test result. Penalty and kinematic contact methods of ABAQUS are used for a mechanical contact formulation. The analysis methods was verified with the rocking angle obtained by seismic response tests. The kinematic contact method with an adequate normal contact stiffness showed a good agreement with tests. Based on the established analysis method for 1/8 scale model, the seismic response analyses of a full scale model are performed for design and beyond design seismic loads.

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

    International Nuclear Information System (INIS)

    Almomania, Belal; Kang, Hyun Gook; Lee, Sanghoon

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  6. Analysis of a ship-to-ship collision

    International Nuclear Information System (INIS)

    Porter, V.L.; Ammerman, D.J.

    1996-01-01

    Sandia National Laboratories is involved in a safety assessment for the shipment of radioactive material by sea. One part of this study is investigation of the consequences of ship-to-ship collisions. This paper describes two sets of finite element analyses performed to assess the structural response of a small freighter and the loading imparted to radioactive material (RAM) packages during several postulated collision scenarios with another ship. The first series of analyses was performed to evaluate the amount of penetration of the freighter hull by a striking ship of various masses and initial velocities. Although these analyses included a representation of a single RAM package, the package was not impacted during the collision so forces on the package could not be computed. Therefore, a second series of analyses incorporating a representation of a row of seven packages was performed to ensure direct package impact by the striking ship. Average forces on a package were evaluated for several initial velocities and masses of the striking ship. In addition to. providing insight to ship and package response during a few postulated ship collisions scenarios, these analyses will be used to benchmark simpler ship collision models used in probabilistic risk assessment analyses

  7. Steady natural convection heat transfer experiments in a horizontal annulus for the United States Spent Fuel Shipping Cask Technology Program

    International Nuclear Information System (INIS)

    Boyd, R.D.

    1981-04-01

    This experimental study deals with the measurement of the heat transfer across a horizontal annulus which is formed by an inner hexagonal cylinder and an outer concentric circular cylinder. The geometry simulates, in two dimensions, a liquid metal fast breeder reactor radioactive fuel subassembly inside a shipping container. This geometry is also similar to a radioactive fuel pin inside a horizontal reactor subassembly. The objective of the experiments is to measure the local and mean heat transfer at the surface of the inner hexagonal cylinder

  8. 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 (MPC) 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. (author)

  9. Development of a Computer Program (CASK) for the Analysis of Logistics and Transportation Cost of the Spent Fuels

    International Nuclear Information System (INIS)

    Cha, Jeong-Hun; Choi, Heui-Joo; Cho, Dong-Keun; Kim, Seong-Ki; Lee, Jong-Youl; Choi, Jong-Won

    2008-07-01

    The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The CASK(Cost and logistics Analysis program for Spent fuel transportation in Korea) program is developed to analyze logistics and transportation cost of the spent fuels. And the total amount of PWR spent fuels stored in four nuclear plant sites, a centralized interim storage facility near coast and a permanent disposal facility near the interim storage facility are considered in this program. The CASK program is developed by using Visual Basic language and coupled with an excel sheet. The excel sheet shows a change of logistics and transportation cost. Also transportation unit cost is easily changed in the excel sheet. The scopes of the report are explanation of parameters in the CASK program and a preliminary calculation. We have developed the CASK version 1.0 so far, and will update its parameters in transportation cost and transportation scenario. Also, we will incorporate it into the program which is used for the projection of spent fuels from the nuclear power plants. Finally, it is expected that the CASK program could be a part of the cost estimation tools which are under development at KAERI. And this program will be a very useful tool for the establishment of transportation scenario and transportation cost in Korean situations

  10. Uncertainty Analysis of the Results of Potential Cask Drops in the K West Basin South Loadout Pit

    International Nuclear Information System (INIS)

    LEW, B.S.

    1999-01-01

    If a cask and Multi-Canister Overpack (MCO) were to be dropped, the analysis established possible initial volumetric flow rates. The results of the uncertainty analysis demonstrated that even if a cask and MCO were to be dropped the possible initial volumetric flow rate through the opened floor/wall joint would be small. There is a 95% confidence (probability of 0.95) that the initial volumetric flow rate would be smaller than 411 gpm and a 99% confidence (probability of 0.99) that it would be smaller than 916 gpm

  11. Quantitative analysis method for ship construction quality

    Directory of Open Access Journals (Sweden)

    FU Senzong

    2017-03-01

    Full Text Available The excellent performance of a ship is assured by the accurate evaluation of its construction quality. For a long time, research into the construction quality of ships has mainly focused on qualitative analysis due to a shortage of process data, which results from limited samples, varied process types and non-standardized processes. Aiming at predicting and controlling the influence of the construction process on the construction quality of ships, this article proposes a reliability quantitative analysis flow path for the ship construction process and fuzzy calculation method. Based on the process-quality factor model proposed by the Function-Oriented Quality Control (FOQC method, we combine fuzzy mathematics with the expert grading method to deduce formulations calculating the fuzzy process reliability of the ordinal connection model, series connection model and mixed connection model. The quantitative analysis method is applied in analyzing the process reliability of a ship's shaft gear box installation, which proves the applicability and effectiveness of the method. The analysis results can be a useful reference for setting key quality inspection points and optimizing key processes.

  12. 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

  13. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  14. Double seal door design and analysis for ITER transfer cask

    International Nuclear Information System (INIS)

    Liu, C.L.; Yao, D.M.; Cheng, T.

    2007-01-01

    DSD (Double seal door) design concept was introduced. 3-D model work was performed for DSD in the three typical regions, such as upper port, equatorial port, divertor port. The numerical analysis for some typical components was done based on Finite Element (FE) method by using ANSYS code, especially for the optimization activities. The rescue procedures of the DSD was discussed which could benefit a little for future engineering implementation. The design and analysis work can support and be the important reference for future procurement. (authors)

  15. 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.

  16. 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

  17. Safety analysis of spent fuel transport and storage casks under extreme impact conditions

    International Nuclear Information System (INIS)

    Wolff, D.; Wieser, G.; Ballheimer, V.; Voelzke, H.; Droste, B.

    2005-01-01

    Full text: Worldwide the security of transport and storage of spent fuel with respect to terrorism threats is a matter of concern. In Germany a spent nuclear fuel management program was developed by the government including a new concept of dry on-site interim storage instead of centralized interim storage. In order to minimize transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities, the operators of NPPs have to erect and to use interim storage facilities for spent nuclear fuel on the site or in the vicinity of nuclear power plants. Up to now, 11 on-site interim storage buildings, one storage tunnel and 4 on-site interim storage areas (preliminary cask storage till the on-site interim storage building is completed) have been licensed at 12 nuclear power plant sites. Inside the interim storage buildings the casks are kept in upright position, whereas at the preliminary interim storage areas horizontal storage of the casks on concrete slabs is used and each cask is covered by concrete elements. Storage buildings and concrete elements are designed only for gamma and neutron radiation shielding reasons and as weather protection. Therefore the security of spent fuel inside a dual purpose transport and storage cask depends on the inherent safety of the cask itself. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. Since the terror attacks of 11 September 2001 the determination of casks' inherent safety also under extreme impact conditions due to terrorist attacks has been of our increasing interest. With respect to spent fuel storage one of the most critical scenarios of a terrorist attack for a cask is the centric impact of a dynamic load onto the lid-seal-system caused e.g. by direct aircraft crash or its engine as well as by a

  18. Exergy Analysis of Complex Ship Energy Systems

    Directory of Open Access Journals (Sweden)

    Pierre Marty

    2016-04-01

    Full Text Available With multiple primary and secondary energy converters (diesel engines, steam turbines, waste heat recovery (WHR and oil-fired boilers, etc. and extensive energy networks (steam, cooling water, exhaust gases, etc., ships may be considered as complex energy systems. Understanding and optimizing such systems requires advanced holistic energy modeling. This modeling can be done in two ways: The simpler approach focuses on energy flows, and has already been tested, approved and presented; a new, more complicated approach, focusing on energy quality, i.e., exergy, is presented in this paper. Exergy analysis has rarely been applied to ships, and, as a general rule, the shipping industry is not familiar with this tool. This paper tries to fill this gap. We start by giving a short reminder of what exergy is and describe the principles of exergy modeling to explain what kind of results should be expected from such an analysis. We then apply these principles to the analysis of a large two-stroke diesel engine with its cooling and exhaust systems. Simulation results are then presented along with the exergy analysis. Finally, we propose solutions for energy and exergy saving which could be applied to marine engines and ships in general.

  19. Proceedings of a workshop on the use of burnup credit in spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1989-10-01

    The Department of Energy sponsored a workshop on the use of burnup credit in the criticality design of spent fuel shipping casks on February 21 and 22, 1988. Twenty-five different presentations on many related topics were conducted, including the effects of burnup credit on the design and operation of spent fuel storage pools, casks and modules, and shipping casks; analysis and physics issues related to burnup credit; regulatory issues and criticality safety; economic incentives and risks associated with burnup credit; and methods for verifying spent fuel characteristics. An abbreviated version of the DOE workshop was repeated as a special session at the November 1988 American Nuclear Society Meeting in Washington, DC. Each of the invited speakers prepared detailed papers on his or her respective topic. The individual papers have been cataloged separately

  20. Shipping

    NARCIS (Netherlands)

    Wijnolst, N.; Wergeland, T.

    1996-01-01

    Shipping is a multi-faceted industry which is rather complex to define from an academic point of view. This book attempts to grasp these complexities and provide the reader with an overview of the main topics and terminology in shipping. The book is based on material from our courses in shipping at

  1. Ship Collision and Grounding Analysis

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup

    2010-01-01

    It is the purpose of the paper to present a review of prediction and analysis tools for collision and grounding analyses and to outline a probabilistic procedure whereby these tools can be used by the maritime industry to develop performance based rules to reduce the risk associated with human, e...

  2. COBRA-SFS thermal analysis of a sealed storage cask for the Monitored Retrievable Storage of spent fuel

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.

    1986-01-01

    The COBRA-SFS (Spent Fuel Storage) computer code was used to predict temperature distributions in a concrete Sealed Storage Cask (SSC). This cask was designed for the Department of Energy in the Monitored Retrievable Storage (MRS) program for storage of spent fuel from commercial power operations. Analytical results were obtained for nominal operation of the SSC with spent fuel from 36 PWR fuel assemblies consolidated in 12 cylindrical canisters. Each canister generates 1650 W of thermal power. A parametric study was performed to assess the effects on cask thermal performance of thermal conductivity of the concrete, the fin material, and the amount of radial reinforcing steel bars (rebar). Seven different cases were modeled. The results of the COBRA-SFS analysis of the current cask design predict that the peak fuel cladding temperature in the SSC will not exceed the 37 0 C design limit for the maximum spent fuel load of 19.8 kW and a maximum expected ambient temperature of 37.8 0 C (100 0 F). The results of the parametric analyses illustrate the importance of material selection and design optimization with regard to the SSC thermal performance

  3. Numerical and experimental analysis of the impact of a nuclear spent fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Aquaro, D. [Department of Mechanical, Nuclear and Production Engineering (DIMNP), Pisa University, Via Diotisalvi, Pisa (Italy); Zaccari, N., E-mail: nicola.zaccari@enel.i [Department of Mechanical, Nuclear and Production Engineering (DIMNP), Pisa University, Via Diotisalvi, Pisa (Italy); Di Prinzio, M.; Forasassi, G. [Department of Mechanical, Nuclear and Production Engineering (DIMNP), Pisa University, Via Diotisalvi, Pisa (Italy)

    2010-04-15

    This paper deals with the numerical and experimental analyses of a shell type shock absorber for a nuclear spent fuel cask. Nine-meter free drop tests performed on reduced scale models are described. The results are compared with numerical simulations performed with FEM computer codes, considering reduced scale models as well as the prototype. The paper shows the results of a similitude analysis, with which the data obtained by means of the reduced scale models can be extrapolated to the prototype. Small discrepancies were obtained using large-scale models (1:2 and 1:6), while small-scale models (1:12) did not give reliable results. A 1:9 scale model provided useful information with a less than 20% error.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  5. Initiatives in transport cask licensing

    International Nuclear Information System (INIS)

    Patterson, John

    1998-01-01

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

  6. 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

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taeman, E-mail: tmkim@korad.or.kr [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Dho, Hoseog; Baeg, Chang-Yeal [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Lee, Gang-uk [Korea Nuclear Engineering and Service Co. (KONES), Hyundai Plaza, 341-4 Jangdae-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2014-10-15

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

  9. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  10. Low-cost concepts for dry transfer of spent fuel and waste between storage and transportation casks

    International Nuclear Information System (INIS)

    Schneider, K.L.

    1984-01-01

    The federal government may provide interim storage for spent fuel from commercial nuclear power reactors that have used up their available storage capacity. One of the leading candidate concepts for this interim storage is to place spent fuel in large metal shielding casks. The Federal Interim Storage (FIS) site may not have the capability to transfer spent fuel from transportation casks to storage casks and vice versa. Thus, there may be an incentive to construct a relatively inexpensive but reliable intercask transfer system for use at an FIS site. This report documents the results of a preliminary study of preconceptual design and analysis of four intercask transfer concepts. The four concepts are: a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); a small hot cell containing equipment for transferring fuel betwee shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept) and a large hot cell, shielded by an earthen berm, that houses equipment for handling fuel between casks that enter and leave the cell on a single cart (igloo concept). Information derived for each of the concepts is operating, capital and relocation costs; implementation and relocation time requirements; and overall characteristics

  11. 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

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

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1988-10-01

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

  13. 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.

  14. 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

  15. Shipping

    OpenAIRE

    Wijnolst, N.; Wergeland, T.

    1996-01-01

    Shipping is a multi-faceted industry which is rather complex to define from an academic point of view. This book attempts to grasp these complexities and provide the reader with an overview of the main topics and terminology in shipping. The book is based on material from our courses in shipping at the universities in Delft and Bergen. As with our lectures, we draw upon quite a va ried material, from research studies at a high academic level to lower level student work and purely descriptive ...

  16. A Mathematical Model for Analysis on Ships Collision Avoidance ...

    African Journals Online (AJOL)

    This study develops a mathematical model for analysis on collision avoidance of ships. The obtained model provides information on the quantitative effect of the ship's engine's response and the applied reversing force on separation distance and stopping abilities of the ships. Appropriate evasive maneuvers require the ...

  17. 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

  18. Analysis of ship life cycles: the impact of economic cycles and ship inspection

    NARCIS (Netherlands)

    Bijwaard, G.E.; Knapp, S.

    2009-01-01

    Due to the shipping industry's international legal framework, there are loopholes in the system, which can increase the risk of incidents with high economic costs due to the substandard operation of vessels. This article uses duration analysis and through the creation of ship life cycles provides

  19. Logistical Analysis of the Littoral Combat Ship

    National Research Council Canada - National Science Library

    Rudko, David

    2003-01-01

    The purpose of the Littoral Combat Ship (LCS) is to provide the Navy with an affordable, small, multi-mission ship capable of independent, interdependent, and integrated operations inside the littorals...

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

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    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)

  1. 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)

  2. 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

  3. Activation analysis of dual-purpose metal cask after the end of design lifetime for decommission

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Man; Ku, Ji Young; Dho Ho Seog; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Ko, Jae Hun [Korea Nuclear Engineering and Service Co., Daejeon (Korea, Republic of)

    2016-12-15

    The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5·ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, {sup 60}Co had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as {sup 28}Al and {sup 24}Na had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that

  4. Multi-purpose canister storage unit and transfer cask thermal analysis

    International Nuclear Information System (INIS)

    Montgomery, R.A.; Niemer, K.A.; Lindner, C.N.

    1997-01-01

    Spent Nuclear Fuel (SNF) generated at commercial nuclear power plants throughout the US is a concern because of continued delays in obtaining a safe, permanent disposal facility. Most utilities maintain their SNF in wet storage pools; however, after decades of use, many pools are filled to capacity. Unfortunately, DOE's proposed final repository at Yucca Mountain is at least 10 years from completion, and commercial power utilities have few options for SNF storage in the interim. The Multi-Purpose Canister (MPC) system, sponsored by DOE's Office of Civilian Radioactive Waste Management, is a viable solution to the interim storage problem. The system is designed for interim dry storage, transport, and ultimate disposal of commercial SNF. The MPC system consists of four separate components: an MPC, Transfer Cask, Storage Unit, and Transport Cask. The SNF assemblies are loaded and sealed inside the helium-filled steel MPC. Once sealed, the MPC is not reopened, eliminating the need to re-handle the individual spent fuel assemblies. The MPC is transferred, using the MPC Transfer Cask, into a cylindrical, reinforced-concrete Storage Unit for on-site dry storage. The MPC may be removed from the Storage Unit at any time and transferred into the MPC Transport Cask for transport to the final repository. This paper discusses the analytical approach used to evaluate the heat transfer characteristics of an MPC containing SNF assemblies in the MPC Transfer Cask and Storage Unit

  5. Structural strength analysis of lug component for RTP fuel transfer cask

    International Nuclear Information System (INIS)

    Ahmad Nabil Abdul Rahim; Tonny Lanyau; Mazleha Maskin

    2012-01-01

    A three decade aging PUSPATI TRIGA Reactor will soon require spent fuels to be replaced with fresh fuels. Initial procedure requires these spent fuels be cooled down due to high temperature and high radioactivity resulted from fission products activities and then transferred out of reactor tank into a spent fuel pool. Performing this task will need a fuel transfer cask (FTC) to provide sufficient radiation shielding for the workers during fuels transferring task. Currently the only economically viable route FTC to get into reactor tank is through the top of reactor tank using crane. The huge and high density lead filled FTC will be hanging throughout the task of transferring each fuel element thus FTC is subjected to tensile stress during its operation. Lug component is anticipated to be one of the main weaknesses of FTC design subjected under tensile load due to it diminutive in size but very large amount of stress to be introduced on it. Therefore lugs component requires an adequate stress analysis to prove the design reliability. This paper will discuss the structural strength using conventional engineering calculation based on best estimate approach. (author)

  6. 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

  7. 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

  8. 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

  9. Safety analysis report for packaging: the ORNL tungsten-shielded cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Levine, D.L.; Just, R.A.

    1977-10-01

    The ORNL tungsten-shielded cask was designed and fabricated at the Oak Ridge National Laboratory (ORNL) for the transport of Type B and large quantity nonfissile isotopes. The container was evaluated analytically to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported. Computational methods were employed in a determination of the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and to the standards for the hypothetical accident conditions. The results of these evaluations demonstrate that the container is in compliance with the applicable regulations

  10. Measurement of dose rates and Monte Carlo analysis of neutrons in a spent-fuel shipping vessel

    International Nuclear Information System (INIS)

    Ueki, K.; Namito, Y.; Fuse, T.

    1986-01-01

    On-board experiments were carried out in a spent-fuel shipping vessel, the Pacific Swan, in which 13 casks of TN-12A and Excellox 3 were loaded in five holds, and neutron and gamma-ray dose rates were measured on the hatch covers of the holds. Before shipping those casks, dose rates were also measured on the cask surfaces, one by one, to eliminate radiation from other casks. The Monte Carlo coupling technique was employed successfully to analyze the measured neutron dose rate distributions in the spent-fuel shipping vessel. Through this study, the Monte Carlo coupling code system, MORSE-CG/CASK-VESSEL, on which the MORSE-CG code was based, was established. The agreement between the measured and the calculated neutron dose rates on the TN-12A cask surface was quite satisfactory. The calculated neutron dose rates agreed with the measured values within a factor of 1.5 on the hold 3 hatch cover and within a factor of 2 on the hold 5 hatch cover in which the concrete shield was fixed in the Pacific Swan

  11. Decree no. 2001-1199 of the 10 december 2001 publishing the resolution MSC. 88 (71) notifying adoption of the international compilation of safety rules for the spent nuclear fuels, plutonium and high level radioactive wastes transport in casks on ships (compilation INF) (annexes), adopted at London the 27 may 1999; Decret no. 2001-1199 du 10 decembre 2001 portant publication de la resolution MSC.88 (71) portant adoption du recueil international de regles de securite pour le transport de combustible nucleaire irradie, de plutonium et de dechets hautement radioactifs en colis a bord de navires (recueil INF) (ensemble une annexe), adoptee a Londres le 27 mai 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This legislative text concerns the safety rules of spent nuclear fuels, plutonium and high level radioactive wastes transport, in casks on ships. Rules, fire prevention, temperature control of casks, electric supply, radioprotection, management and emergency plans are detailed. (A.L.B.)

  12. Applying consensus standards to cask development

    International Nuclear Information System (INIS)

    Leatham, J.; Abbott, D.G.; Warrant, M.M.

    1987-01-01

    The Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management is procuring cask systems for transporting commercial spent nuclear fuel and is encouraging development of innovative cask designs and materials to improve system efficiency. New designs and innovative materials require that consensus standards be established so that cask designers and regulators have criteria for determining acceptability. Recent DOE experience in certifying three spent fuel shipping casks, NUPAC-125B, TN-BRP, and TN-REG, is discussed. Certification of the NUPAC-125B was expedited because it was made of conventional American Society for Testing and Materials (ASTM) materials and complied with the American Society of Mechanical Engineers (ASME) Code and Nuclear Regulatory Commission Regulatory Guides. The TN-BRP and TN-REG cask designs are still being reviewed because baskets included in the casks are made of borated stainless steel, which has no ASTM Specification or ASME Code approval. The process of developing and approving consensus standards is discussed, including the role of ANSI and ANSI N14. Specific procedures for ASTM and ASME are described. A draft specification or standard must be prepared and then approved by the appropriate body. For new material applications to the ASME Code, an existing ASTM Specification is needed. These processes may require several years. The status of activities currently in progress to develop consensus standards for spent fuel casks is discussed, including (1) ASME NUPAC, and (2) ASTM Specifications for ductile cast iron and borated stainless steel

  13. 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.

  14. Documentation associated with the shipping of Hot-Cell Waste from WESF 225-B to the 200W (218-W-3AE) burial grounds under shipment number RSR-37338

    International Nuclear Information System (INIS)

    PAWLAK, M.W.

    1998-01-01

    The purpose of this report is to compile the records generated during the Packaging and Shipping of WESF Hot-Cell Waste from the 225-B Facility to 200W (218-W-3AE) burial grounds. A total of six 55-gallon drums were packaged and shipped using the Chem-Nuc Cask in accordance with WHC-SD-TP-SARP-025, Rev.0 ''Safety Analysis Report for Packaging (Onsite) for Type B Material in the CNS-14-215H Cask''

  15. Thermal analysis of HNPF spent fuel shipping container in torch environments

    International Nuclear Information System (INIS)

    Eggers, P.E.; Crawford, H.L.; Trujillo, A.A.; Pope, R.B.; Yoshimura, H.R.

    1980-01-01

    A thermal analysis of a modified HNPF cask has been performed to predict cask temperatures in simulated torch fire environments. A portion of this analysis also provided a basis for the modifications leading to the design of external cooling fins, a pressure relief valve system, a spent fuel torch fire analyses will provide comparison with experimentally measured temperatures of the HNPF cask, in torch fire environments, which will be obtained in the future. Scope of this study was analytical in nature and involved the use of a computer-based heat transfer analysis methods. Product of this study is the prediction of cask temperatures and internal pressures under normal and torch-fire accident conditions. This study also involved an analysis of the torch fire literature in order to develop an empirically based thermal model for the torch environment. 8 figures, 3 tables

  16. Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohammadi, Ashgar; Omidvari, Nima [Iran Radioactive Waste Management Company, Tehran (Iran, Islamic Republic of); Hassanzadeh, Mostafa [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2017-12-15

    Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k{sub eff}) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

  17. Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

    International Nuclear Information System (INIS)

    Mohammadi, Ashgar; Omidvari, Nima; Hassanzadeh, Mostafa

    2017-01-01

    Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k eff ) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

  18. 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

  19. Infrared ship signature analysis and optimisation

    NARCIS (Netherlands)

    Neele, F.P.

    2005-01-01

    The last decade has seen an increase in the awareness of the infrared signature of naval ships. New ship designs show that infrared signature reduction measures are being incorporated, such as exhaust gas cooling systems, relocation of the exhausts and surface cooling systems. Hull and

  20. B cell remote-handled waste shipment cask alternatives study; TOPICAL

    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

  1. Methodology for a thermal analysis of a proposed SFR transport cask with the thermal code SYRTHES

    International Nuclear Information System (INIS)

    Peniguel, C.; Rupp, I.; Schneider, J. P.

    2010-01-01

    Fast reactors with liquid metal coolant have received a renewed interest owing to the need of a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In the framework of the 2006 French law on sustainable management of radioactive materials and waste, an evaluation of the industrial perspectives of minor actinides transmutation advantages and drawbacks in Generation IV fast spectrum reactors system is requested for 2012. The CEA is in charge of studying the global problem, but on some aspects, EDF is interested to do its own exploratory studies. Among other points, transport is seen as important for the nuclear industry, to link points of production and treatment. Nuclear fuel is generally transported in thick walled rail or truck casks. These packages are designed to provide confinement, shielding and criticality protection during normal and severe transport conditions. Heat generated within the fuel (and a contribution of solar heating) makes the package becoming quite hot, but one must demonstrate that the cladding temperature does not exceed a long term temperature limit during normal transport. This paper presents a thermal study done on a package in which 9 SFR assemblies are included. Each of them is of hexagonal shape and contains 271 fuel pins. The approach followed for these calculations is to rely on an explicit representation of all pins. For these calculations a 2D analysis is performed thanks to the thermal code SYRTHES. Conduction is solved thanks to a finite element method, while thermal radiation is handled through a radiosity approach. The main aim of this paper is to present a possible numerical methodology to handle the thermal problem. (authors)

  2. Structural analysis of a ship on global aspect using ANSYS

    Science.gov (United States)

    Rahman, M. Muzibur; Kamol, Rajia Sultana; Islam, Reyana

    2017-12-01

    Ship is a complex geometry which undergoes a combination of loadings such as hydrostatic, hydrodynamic, wind, wave etc. at sea and thus adequate strength in a ship has always been one of the most challenging tasks for the ship designers. International Maritime Organization (IMO) and classification societies are providing the standards to ensure the adequacy of strength for the ship against all demands throughout its service life. Thus, structural analysis is needed to assess the overall strength of hull, and the means in this regard are based on finite element method which may be applied either local or global aspect of the ship. This paper is an attempt to carry out the structural analysis of a ship in global aspect using ANSYS software to locate the most stress concentration and deformed area, which will have ultimate effect on fatigue fracture.

  3. A CASKCOM: A cask life cycle cost model

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    CASKCOM (cask cost model) is a computerized model which calculates the life cycle costs (LCC) associated with specific transportation cask designs and discounts those costs, if the user so chooses, to a net present value. The model has been used to help analyze and compare the life cycle economics of burnup credit and nonburnup credit cask designs being considered as conditions for a new generation of spent fuel transportation casks. CASKCOM is parametric in the sense that its input data can be easily changed in order to analyze and compare the life cycle cost implications arising from alternative assumptions. The input data themselves are organized into two main groupings. The first grouping comprises a set of data which is independent of cask design. This first grouping does not change from the analysis of one cask design to another. The second grouping of data is specific to each individual cask design. This second grouping thus changes each time a new cask design is analyzed

  4. Fuel-assembly behavior under dynamic impact loads due to dry-storage cask mishandling

    International Nuclear Information System (INIS)

    1991-07-01

    Continued operation of nuclear power plants is contingent on the ability to provide adequate storage of spent fuel. Until recently, utilities have been able to maintain interim in-pool spent fuel storage. However, many facilities have reached their capacity and are now faced with shipping their spent fuel in dry casks to alternate storage facilities. The objective of this report is to provide estimates of the structural integrity of irradiated LWR fuel rods subjected to impact loads resulting from postulated cask handling accidents. This is accomplished in five stages: (1) Material properties for irradiated fuel are compiled for use in the structural analyses. (2) Results from parametric analyses of representative assembly designs are used to determine the most limiting case for end and side drop postulated handling accidents. (3) Detailed structural analysis results are presented for these critical designs. The detailed analyses include the coupling of assembly interaction with the cask and cask internals. (4) Criteria for both ultimate stress and brittle fracture failure modes of fuel rod cladding are established. (5) Safe cask handling drop height limits are computed based on items 2 through 4 above. 44 figs., 18 tabs

  5. Ship

    OpenAIRE

    Keuning, J.A.

    2008-01-01

    The invention concerns a ship designed for use at high speed and heavy seas having a single long and slender hull with a narrow beam and a more or less vertical bow, whereby the front half of the hull has more or less vertical sides, minimal flare in the bow sections and towards the bow an increase in draught at its center line combined with a more or less similar increase of freeboard and whereby the aft end of the hull has a flat or slightly V-shaped bottom with one or more propellers and/o...

  6. Nuclear ship accidents, description and analysis

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1993-03-01

    In this report available information on 44 reported nuclear ship events is considered. Of these 6 deals with U.S. ships and 38 with USSR ships. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/ explosions, sea-water leaks into the submarines and sinking of vessels are considered. Comments are made on each of the events, and at the end of the report an attempt is made to point out the weaknesses of the submarine designs which have resulted in the accidents. It is emphasized that some of the information of which this report is based, may be of dubious nature. Consequently some of the results of the assessments made may not be correct. (au)

  7. Safety analysis report for packaging (onsite) L3-181 N basin cask

    International Nuclear Information System (INIS)

    Adkins, H.E. Jr.

    1996-01-01

    Purpose of this Safety Analysis Report (SARP) is to authorize the onsite transfer of a Type B, Fissile Excepted, non-highway route controlled quantity in the L3-181 packaging from the N Basin to a storage/disposal facility within 200 West Area. This SARP provides the evaluation necessary to demonstrate that the L3-181 meets the requirements of the 'Hazardous Material Packaging and Shipping', WHC- CM-2-14, by meeting the applicable performance requirements for normal conditions of transport

  8. Finite element analysis for the impact behaviour of a cask interacting with a rigid pin

    International Nuclear Information System (INIS)

    Altes, J.; Geiser, H.; Voelzer, W.; Frenk, A.; Deeken, G.

    1993-01-01

    Full scale drop tests of casks to be licensed as type B packages according to the IAEA regulations for the safe transport of radioactive materials are expensive. Therefore efforts are being made to use computer codes for calculating the impact behaviour. But these codes have to be verified by experiments. Codes available for these calculations are for example DYNA3D and ABAQUS. In the paper results of both codes are compared. A 11 t ductile cast iron cask (type MOSAIK) without impact limiters was analysed dropping from a height of 1 m with its top onto a cylindrical steel pin. The results of the finite element calculations with both codes show good agreement. The ABAQUS results using the implicit method are in accordance with the explicit method, for which considerably shorter CPU times are noted. (author)

  9. Dynamic analysis to establish normal shock and vibration of radioactive material shipping packages

    International Nuclear Information System (INIS)

    Fields, S.R.

    1980-01-01

    A computer model, CARDS (Cask-Railcar Dynamic Simulator) was developed to provide input data for a broad range of radioactive material package-tiedown structural assessments. CARDS simulates the dynamic behavior of shipping packages and their transporters during normal transport conditions. The model will be used to identify parameters which significantly affect the normal shock and vibration environments which, in turn, provide the basis for determining the forces transmitted to the packages

  10. Crash test of a nuclear spent fuel cask and truck transport system

    International Nuclear Information System (INIS)

    Huerta, M.; Yoshimura, R.H.

    1978-01-01

    Sandia Laboratories has conducted a 96 kph (60 mph) full scale truck impact test for ERDA's Environmental Control Technology Division. Rockets propelled a 20, 500-kg (22-ton) cask mounted on its shipping trailer, coupled to a conventional cab-over tractor, into a massive, heavily reinforced concrete target. This summary report describes and compares the results of the computer analysis, scale model, and full scale tests

  11. Analysis of ship maneuvering data from simulators

    Science.gov (United States)

    Frette, V.; Kleppe, G.; Christensen, K.

    2011-03-01

    We analyze complex manuevering histories of ships obtained from training sessions on bridge simulators. Advanced ships are used in fields like offshore oil exploration: dive support vessels, supply vessels, anchor handling vessels, tugs, cable layers, and multi-purpose vessels. Due to high demands from the operations carried out, these ships need to have very high maneuverability. This is achieved through a propulsion system with several thrusters, water jets, and rudders in addition to standard propellers. For some operations, like subsea maintenance, it is crucial that the ship accurately keeps a fixed position. Therefore, bridge systems usually incorporate equipment for Dynamic Positioning (DP). DP is a method to keep ships and semi submersible rigs in a fixed position using the propulsion systems instead of anchors. It may also be used for sailing a vessel from one position to another along a predefined route. Like an autopilot on an airplane, DP may operate without human involvement. The method relies on accurate determination of position from external reference systems like GPS, as well as a continuously adjusted mathematical model of the ship and external forces from wind, waves and currents. In a specific simulator exercise for offshore crews, a ship is to be taken up to an installation consisting of three nearby oil platforms connected by bridges (Frigg field, North Sea), where a subsea inspection is to be carried out. Due to the many degrees of freedom during maneuvering, including partly or full use of DP, the chosen routes vary significantly. In this poster we report preliminary results on representations of the complex maneuvering histories; representations that allow comparison between crew groups, and, possibly, sorting of the different strategic choices behind.

  12. Cask technology program activities

    International Nuclear Information System (INIS)

    Allen, G.C. Jr.

    1986-01-01

    The civilian waste cask technology program consists of five major activities: (1) technical issue resolution directed toward NRC and DOT concerns, (2) system concept evaluations to determine the benefits of proposals made to DOE for transportation improvements, (3) applied technology and technical data tasks that provide independent information and enhance technology transfer between cask contractors, (4) standards development and code benchmarking that provide a service to DOE and cask contractors, and (5) testing to ensure the adequacy of cask designs. The program addresses broad issues that affect several cask development contractors and areas where independent technical input could enhance the Office of Civilian Radioactive Waste Management goals

  13. Cask technology program activities

    International Nuclear Information System (INIS)

    Allen, G.C. Jr.

    1986-01-01

    The civilian waste cask technology program consists of five major activities: Technical issue resolution directed toward NRC and DOT concerns; system concept evaluations to determine the benefits of proposals made to DOE for transportation improvements; applied technology and technical data tasks that provide independent information and enhance technology transfer between cask contractors; standards development and code benchmarking that provide a service to DOE and cask contractors; and testing to ensure the adequacy of cask designs. This paper addresses broad issues that affect several cask development contractors and areas where independent technical input could enhance OCRWM goals

  14. Functions of the cask maintenance facility: A white paper

    International Nuclear Information System (INIS)

    1987-01-01

    The shipping cask systems are the mobile components of the transportation system, designed to safely transport spent nuclear fuel between different facilities under both normal and accident conditions. The cask system will consist of the heavily shielded cask, the cask transport vehicle (truck trailer or railcar), and any associated ancillary equipment (covers, impact limiters, lifting devices, etc.). The cask and certain parts of the cask system must be operated within the limits imposed by a certificate of compliance (COC) granted by the Nuclear Regulatory Commission (NRC). Each cask system must transport spent fuel safely during the life of the system. To maintain the operational effectiveness and safety of the cask systems, a cask maintenance facility (CMF) will be included as an integral part of the transportation system. The planning activity of the transportation system and the design effort of the CMF require that the functions to be performed by the CMF be explicitly defined. The purpose of this paper is to (1) define the potential transportation system functions to be performed at the CMF; (2) examine the impact of this functional definition on the overall transportation system; (3) identify any unresolved issues concerning the interaction of the CMF with other elements of the transportation system; and (4) make recommendations to resolve any unresolved issues so that decisions can be made early in the transportation system planning process

  15. 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)

  16. 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

  17. 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

  18. 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.)

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

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  1. CRUSH1: a simplified computer program for impact analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

    In drop impact analyses for radioactive transport casks, it has become possible to perform them in detail by using interaction evaluation, computer programs, such as DYNA2D, DYNA3D, PISCES and HONDO. However, the considerable cost and computer time are necessitated to perform analyses by these programs. To meet the above requirements, a simplified computer program CRUSH1 has been developed. The CRUSH1 is a static calculation computer program capable of evaluating the maximum acceleration of cask bodies and the maximum deformation of shock absorbers using an Uniaxial Displacement Method (UDM). The CRUSH1 is a revised version of the CRUSH. Main revisions of the computer program are as follows; (1) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows 3.1 or Windows NT) are available for use of the CRUSH1 and (2) input data set are revised. In the paper, brief illustration of calculation method using UDM is presented. The second section presents comparisons between UDM and the detailed method. The third section provides a use`s guide for CRUSH1. (author)

  2. CRUSH1: a simplified computer program for impact analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1996-07-01

    In drop impact analyses for radioactive transport casks, it has become possible to perform them in detail by using interaction evaluation, computer programs, such as DYNA2D, DYNA3D, PISCES and HONDO. However, the considerable cost and computer time are necessitated to perform analyses by these programs. To meet the above requirements, a simplified computer program CRUSH1 has been developed. The CRUSH1 is a static calculation computer program capable of evaluating the maximum acceleration of cask bodies and the maximum deformation of shock absorbers using an Uniaxial Displacement Method (UDM). The CRUSH1 is a revised version of the CRUSH. Main revisions of the computer program are as follows; (1) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows 3.1 or Windows NT) are available for use of the CRUSH1 and (2) input data set are revised. In the paper, brief illustration of calculation method using UDM is presented. The second section presents comparisons between UDM and the detailed method. The third section provides a use's guide for CRUSH1. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  4. Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

    2000-07-01

    Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

  5. Testing of Metal Cask and Concrete Cask

    International Nuclear Information System (INIS)

    Shirai, K.; Wataru, M.; Takeda, H.; Tani, J.; Arai, T.; Saegusa, T.

    2015-01-01

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

  6. 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)

  7. 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

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

    International Nuclear Information System (INIS)

    CARLSON, A.B.

    1999-01-01

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

  9. Strength Analysis on Ship Ladder Using Finite Element Method

    Science.gov (United States)

    Budianto; Wahyudi, M. T.; Dinata, U.; Ruddianto; Eko P., M. M.

    2018-01-01

    In designing the ship’s structure, it should refer to the rules in accordance with applicable classification standards. In this case, designing Ladder (Staircase) on a Ferry Ship which is set up, it must be reviewed based on the loads during ship operations, either during sailing or at port operations. The classification rules in ship design refer to the calculation of the structure components described in Classification calculation method and can be analysed using the Finite Element Method. Classification Regulations used in the design of Ferry Ships used BKI (Bureau of Classification Indonesia). So the rules for the provision of material composition in the mechanical properties of the material should refer to the classification of the used vessel. The analysis in this structure used program structure packages based on Finite Element Method. By using structural analysis on Ladder (Ladder), it obtained strength and simulation structure that can withstand load 140 kg both in static condition, dynamic, and impact. Therefore, the result of the analysis included values of safety factors in the ship is to keep the structure safe but the strength of the structure is not excessive.

  10. Dynamic analysis of CHASNUPP steam generator structure during shipping

    International Nuclear Information System (INIS)

    Han Liangbi; Xu Jinkang; Zhou Meiwu; He Yinbiao

    1998-07-01

    The dynamic analysis of CHASNUPP steam generator during shipping is described, including the simplified mathematical model, acceleration power spectrum of ocean wave induced random vibration, the dynamic analysis of steam generator structure under random loading, the applied computer code and calculated results

  11. 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.

  12. 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.

  13. Preliminary analysis of the postulated changes needed to achieve rail cask handling capabilities at selected light water reactors

    International Nuclear Information System (INIS)

    Konzek, G.J.

    1986-02-01

    Reactor-specific railroad and crane information for all LWRs in the US was extracted from current sources of information. Based on this information, reactors were separated into two basic groups consisting of reactors with existing, usable rail cask capabilities and those without these capabilities. The latter group is the main focus of this study. The group of reactors without present rail cask handling capabilities was further separated into two subgroups consisting of reactors considered essentially incapable of handling a large rail cask of about 100 tons and reactors where postulated facility changes could result in rail cask handling capabilities. Based on a selected population of 127 reactors, the results of this assessment indicate that usable rail cask capabilities exist at 83 (65%) of the reactors. Twelve (27%) of the remaining 44 reactors are deemed incapable of handling a large rail cask without major changes, and 32 reactors are considered likely candidates for potentially achieving rail cask handling capabilities. In the latter group, facility changes were postulated that would conceptually enable these reactors to handle large rail casks. The estimated cost per plant of required facility changes varied widely from a high of about $35 million to a low of <$0.3 million. Only 11 of the 32 plants would require crane upgrades. Spur track and right-of-way costs would apparently vary widely among sites. These results are based on preliminary analyses using available generic cost data. They represent lower bound values that are useful for developing an initial assessment of the viability of the postulated changes on a system-wide basis, but are not intended to be absolute values for specific reactors or sites

  14. Proposal for the rocking analysis model of the dry cask for spent nuclear fuel attached to the storage pallet subjected to the strong earthquake motions

    International Nuclear Information System (INIS)

    Kondo, Shunsuke; Shirai, Koji; Namba, Kosuke

    2016-01-01

    In Japan, a dry cask for spent nuclear fuel attached to a storage pallet should be transferred and stored in the vertical orientation on the concrete floor in an interim spent nuclear fuel storage facility at or outside reactor site, and a transfer system using air supply will be adopted for such pallet. In case of the hypothetical event, the shutdown of the air supply due to the strong earthquake motions, it is important to evaluate a stability of the metal cask on the concrete floor during seismic motions. A dynamic analysis by the analysis code 'TDAPIII' was executed with a simple lumped mass model by adopting joint elements between a concrete floor and pallet, to reproduce the rocking and sliding behavior. Joint stiffness values were equivalently set to the vibration modes obtained by an eigenvalue analysis. The seismic analysis results were compared with the previous shaking table test results with 2/5 scale model of a real size cask. As a result, although discrepancies of the velocity response of the converted from maximum uplifting potential energy appeared in the range of μ ± 3σ (0.57 ∼ 1.46) among 45 analysis cases comparing with experiment results, it was confirmed that maximum value was about 110kine considerably less than the overturning threshold value 190kine. Moreover, an applicability of the proposed prediction methodology to the real size model was also confirmed. (author)

  15. Safety analysis report for packaging: the ORNL HFIR unirradiated fuel element shipping container

    International Nuclear Information System (INIS)

    Evans, J.H.; Boulet, J.A.M.; Eversole, R.E.

    1977-11-01

    The ORNL HFIR unirradiated fuel element shipping container was designed and fabricated at the Oak Ridge National Laboratory for the transport of HFIR unirradiated fuel elements. The container was evaluated analytically and experimentally to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and the evaluation is the subject of this report. Computational and test procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation demonstrate that the container is in compliance with the applicable regulations

  16. Technical issues affecting the transport of dual purpose casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Ottinger, C.A.; Brimhall, J.L.; Gilbert, E.R.; Jones, R.H.

    1989-01-01

    Spent fuel storage pools at many nuclear reactors in the US have already or will soon be filled to maximum capacity. Approximately 50,000 metric tons of uranium (MTU) spent fuel will be discharged by the projected 2003 start-up date of a federal disposal system. Of this, approximately 6,000 MTU will require storage outside existing or projected pool storage capabilities (DOE, 1988). At-reactor dry storage of spent fuel, including vault, caisson, and cask systems, is being considered as an alternative to accommodate this excess fuel. Two dry storage cask concepts are among those under consideration. One involves placing spent fuel in storage-only casks (SOC) until a monitored retrievable storage (MRS) facility or repository is open when the spent fuel would be transferred to a transport-only cask (TOC) for shipment. The second option, the dual purpose or transportable storage cask (TSC), is a system that would serve for both storage and later transport without requiring the spent fuel to be unloaded. To carry out its purpose, a TSC must be shipped directly from a storage facility to a disposal facility without first being opened to evaluate the cask or the fuel. To assure that both the fuel and the cask are in a transportable condition after 20 to 40 years of storage requires: (1) a definition of expected storage conditions; (2) an assessment of the impact of expected storage conditions on the reliability of the components and functions of the TSC during transport; and (3) the development of an overall TSC system design and operational strategy which ensures that TSC transport reliability meets or exceeds that of a transport-only cask. The later requirement is related to defining what appropriate design features, pre-shipment inspections, and/or alternative fuel and cask monitoring requirements are necessary during long-term storage to ensure the cask will meet transport requirements during later transport

  17. Technical issues affecting the transport of dual purpose casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Ottinger, C.A.; Brimhall, J.L.; Gilbert, E.R.; Jones, R.H.

    1989-01-01

    Approximately 60,000 metric tons of uranium (MTU) spent fuel will be discharged by the projected 2003 startup date of a federal disposal system. Of this, approximately 15,000 MTU will require storage outside existing or projected pool storage capabilities (Orvis et al., 1984). At-reactor dry storage of spent fuel, including vault, caisson, and cask systems, is being considered as an alternative to accommodate this excess fuel. Two dry storage cask concepts are among those under consideration. One involves placing spent fuel in storage-only casks (SOC) until a monitored retrievable storage (MRS) facility or repository is open, when the spent fuel would be transferred to a transport-only cask (TOC) for shipment. The second option, the dual purpose or transportable storage cask (TSC), is a system that would serve for both storage and later transport. To carry out its purpose, a TSC must be shipped directly from a storage facility to a disposal facility without first being opened to evaluate the cask or the fuel. To assure that both the fuel and the cask are in a transportable condition after 20 to 40 years of storage requires: (1) a definition of expected storage conditions; (2) an assessment of the impact of expected storage conditions on the reliability of the components and functions of the TSC during transport; and (3) the development of an overall TSC system design and operational strategy which ensures that TSC transport reliability compares to that of a transport-only cask. The later requirement is related to defining what appropriate design features, pre-shipment inspection, and/or alternative fuel and cask monitoring requirements are necessary during long-term storage to ensure the cask will meet transport performance requirements during later transport. 8 refs., 1 fig., 1 tab

  18. 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

  19. LCA-ship. Design tool for energy efficient ships. A Life Cycle Analysis Program for Ships. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jiven, Karl; Sjoebris, Anders [MariTerm AB, Goeteborg (Sweden); Nilsson, Maria [Lund Univ. (Sweden). Stiftelsen TEM; Ellis, Joanne; Traegaardh, Peter; Nordstroem, Malin [SSPA Sweden AB, Goeteborg (Sweden)

    2004-05-01

    In order to make it easier to include aspects during ship design that will improve environmental performance, general methods for life cycle calculations and a prototype tool for LCA calculations of ships and marine transportation have been developed. The base of the life cycle analyses is a comprehensive set of life cycle data that was collected for the materials and consumables used in ship construction and vessel operations. The computer tool developed makes it possible to quickly and simply specify (and calculate) the use of consumables over the vessel's life time cycle. Special effort has been made to allow the tool to be used for different types of vessels and sea transport. The main result from the project is the computer tool LCA ship, which incorporates collected and developed life cycle data for some of the most important materials and consumables used in ships and their operation. The computer application also contains a module for propulsion power calculations and a module for defining and optimising the energy system onboard the vessel. The tool itself is described in more detail in the Computer application manual. The input to the application should, as much as possible, be the kind of information that is normally found in a shipping company concerning vessel data and vessel movements. It all starts with defining the ship to be analysed and continues with defining how the ship is used over the lifetime. The tool contains compiled and processed background information about specific materials and processes (LCA data) connected to shipping operations. The LCA data is included in the tool in a processed form. LCA data for steel will for example include the environmental load from the steel production, the process to build the steel structure of the ship, the scrapping and the recycling phase. To be able to calculate the environmental load from the use of steel the total amount of steel used over the life cycle of the ship is also needed. The

  20. 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

  1. Analysis framework to calibrate a numerical model to simulate the thermal test of a 1:2 scale dual purpose cask under accident conditions

    International Nuclear Information System (INIS)

    Miranda, Carlos A.J.; Libardi, Rosani M.P.; Marcelino, Sergio; Oliveira, Carlos Alberto de; Mattar Neto, Miguel

    2013-01-01

    This work describes thermal analysis framework including a 3D model and some 2D models to be performed in a 1:2 scale model of a dual-purpose cask to transport and to store spent fuel elements from research reactors to assess the behavior of the cask structure and materials when submitted to heating and drop tests. The analyses should consider all non-linearities involved like the lead phase change and thermal contacts, beside the variation of material properties with the temperature, the air inside it and the heat transfer phenomena (conduction, convection and irradiation) to reproduce the experimental results already obtained in a 1:2 model. A full 3D finite element model takes several hours to run just one analysis. To speed up the analyses to evaluate the significance of some parameters like the emissivity, contact resistance and heat transfer phenomena, among others, two 2D models are planned: one simulating a vertical cut by a diametral plane and another one simulating a horizontal cut by a plane at the cask half height. These 2D models are predicted to run fast enough to allow several analyses in a short period of time and to define options and the best parameters values to match the already obtained experimental results. As this thermal test can not be extrapolated to an 1:1 scale, these parameter values will be used in the final 3D model analysis and also in the full scale model. (author)

  2. Safety analysis report on Model UC-609 shipping package

    International Nuclear Information System (INIS)

    Sandberg, R.R.

    1977-08-01

    This Safety Analysis Report for Packaging demonstrates that model UC-609 shipping package can safely transport tritium in any of its forms. The package and its contents are described. The package when subjected to the transport conditions specified in the Code of Federal Regulations, Title 10, Part 71 is evaluated. Finally, compliance with these regulations is discussed

  3. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  4. Design assessment for transport and storage casks

    International Nuclear Information System (INIS)

    Janberg, K.; Diersch, R.; Spilker, H.; Dreier, G.

    1995-01-01

    The design assessment concerning the mechanical behaviour of transport and storage casks for radioactive material to fulfil nuclear safety criteria has to be based on two essential considerations: (1) Effective analysis of the stress-strain state of the cask components under both normal operational and test conditions including hypothetical accident scenarios with suitable accepted methods. (2) Economic estimation of the required properties and the structural state of the cask components with sufficient exactness. In an overview of the codes which are available at GNS/GNB for cask impact strength analyses (ANSYS, ADINA, VDI Codes), procedures and aspects of benchmarking and validation of calculation codes are described. The results of experimental full size cask drop test programs (CASTOR, POLLUX) and corresponding pre-test calculational analyses show the suitability of the codes used. The influence of dynamic effects on the mechanical properties of material (ductile cast iron, wood) has been investigated experimentally. By consideration of these dynamic values in strength analyses of casks at impact a good agreement between experimental and calculational results has been achieved. (author)

  5. Structural analysis of aircraft impact on a nuclear powered ship

    International Nuclear Information System (INIS)

    Dietrich, R.

    1976-01-01

    As part of a general safety analysis, the reliability against structural damage due to an aircraft crash on a nuclear powered ship is evaluated. This structural analysis is an aid in safety design. It is assumed that a Phantom military jet-fighter hits a nuclear powered ship. The total reaction force due to such an aircraft impact on a rigid barrier is specified in the guidelines of the Reaktor-Sicherheitskommission (German Safety Advisory Committee) for pressurized water reactors. This paper investigates the aircraft impact on the collision barrier at the side of the ship. The aircraft impact on top of the reactor hatchway is investigated by another analysis. It appears that the most unfavorable angle of impact is always normal to the surface of the collision barrier. Consequently, only normal impact will be considered here. For the specific case of an aircraft striking a nuclear powered ship, the following two effects are considered: Local penetration and dynamic response of the structure. (Auth.)

  6. Development of radar cross section analysis system of naval ships

    Directory of Open Access Journals (Sweden)

    Kookhyun Kim

    2012-03-01

    Full Text Available A software system for a complex object scattering analysis, named SYSCOS, has been developed for a systematic radar cross section (RCS analysis and reduction design. The system is based on the high frequency analysis methods of physical optics, geometrical optics, and physical theory of diffraction, which are suitable for RCS analysis of electromagnetically large and complex targets as like naval ships. In addition, a direct scattering center analysis function has been included, which gives relatively simple and intuitive way to discriminate problem areas in design stage when comparing with conventional image-based approaches. In this paper, the theoretical background and the organization of the SYSCOS system are presented. To verify its accuracy and to demonstrate its applicability, numerical analyses for a square plate, a sphere and a cylinder, a weapon system and a virtual naval ship have been carried out, of which results have been compared with analytic solutions and those obtained by the other existing software.

  7. Post test evaluation of a fire tested rail spent fuel cask

    International Nuclear Information System (INIS)

    Rack, H.J.; Yoshimura, H.R.

    1980-01-01

    Postmortem examination of a large rail-transported spent fuel shipping cask which had been exposed to a JP-4 fuel fire revealed the presence of two macrofissures in the outer cask shell. One, a part-through crack located within the seam weld fusion zone of the outer cask shell, is typical of hot cracks found in stainless steel weldments. The other, a through-crack, was apparently initiated during the formation of a copper-stainless steel dissimilar metal joint, with crack propagation through the cask outer shell having occurred during the fire-test. 8 figures

  8. Application of the ASME code in the design of the GA-4 and GA-9 casks

    International Nuclear Information System (INIS)

    Mings, W.J.; Koploy, M.A.

    1992-01-01

    General Atomics (GA) is developing two spent fuel shipping casks for transport by legal weight truck (LWT). The casks are designed to the loading, environmental conditions and safety requirements defined in Title 10 of the Code of Federal Regulations, Part 71 (10CFR71). To ensure that all components of the cask meet the 10CFR71 rules, GA established structural design criteria for each component based on NRC Regulatory Guides and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This paper discusses the criteria used for different cask components, how they were applied and the conservatism and safety margins built into the criteria and assumption

  9. EBRII cask characterization measurements

    International Nuclear Information System (INIS)

    Haggard, D.L.; Brackenbush, L.W.

    1996-01-01

    This report describes the measurements performed to provide the radionuclide content and verify the stated mass of special nuclear material (SNM) in Experimental Breeder Reactor EBRII casks stored in Trench 1, Burial Ground 4C, 218-WAC 200 West Area. this information is needed to characterize the curie content of each cask and the total curies in the storage area. Gamma assay techniques typically employed for nondestructive assay (NDA) were used to determine the gamma-emitting isotopes in each cask, which were fission and activation products from the spent fuel. Passive neutron counting was selected to verify the stated plutonium content because the fission and activation products masked any gamma emissions from plutonium. The fast neutrons emitted by plutonium are highly penetrating and easily detected through several inches of shielding. A slab neutron detector containing five 3 He proportional counters was used to determine the neutron emission rates and estimate the mass of plutonium present. The measurements followed the methods and procedures routinely used for nuclear waste assay and safeguard measurements. The measured neutron yields confirmed the declared plutonium content for the fuel elements, with the exception of several casks that contained recycled plutonium or americium target material. In these casks, the 244 Cm content masked the neutron emissions from any plutonium. For these casks, the plutonium content was estimated by correlation with the 244 Cm neutron emissions

  10. 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.)

  11. 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

  12. Econometric Analysis of Bulk Shipping: implications for investment strategies and financial decision-making

    NARCIS (Netherlands)

    S. Tsolakis

    2005-01-01

    textabstractThis thesis provides an econometric analysis of the bulk shipping markets and the implications for shipping investment and financial decision making. Chapter 1 sets the scene by providing a historic analysis of bulk shipping markets over the last 55 years. From this analysis, four

  13. 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)

  14. Shipment and Storage Containers for Tritium Production Transportation Casks

    International Nuclear Information System (INIS)

    Massey, W.M.

    1998-01-01

    A shipping and storage container for the Tritium production transportation casks may be required but requirements for protection of the irradiated rods and radioactive contamination have not been finalized. This report documents the various possibilities for the container depending on the final requirements

  15. Risk Analysis on Ship Wreck and Container Cargo to Ship Navigation

    Directory of Open Access Journals (Sweden)

    Muhammad Badrus Zaman

    2017-03-01

    Full Text Available Wreck of a ship is an incident that must be avoided. Ship accidents are generally caused by a several cases, such as human error, natural disaster, technical errors, missed communication, poor condition of the ship, and many more. Ship wreckage have huge impact for ship navigation, environment, economics, and others. Those impact have many disadvantages for the shipowners, and also for environment. For examples the fuel spills that pollute the environment, make disturbance to sailing ship because the track for those navigation is blocked by the ship wreck and their cargo especially on shallow location (<50 m. These research will discuss the effect the container when it is floats on the sea and its interference other ships. The main objective of this study is to present a risk assessment on the environmental impact of the wreck and container cargo. Wrecks on the seabed is likely to pose a risk to passing ships. container and its contents as well as the possibility of refloat, and also their environmental risks emanating from the wreck and container cargo, such as fuels, lubricants, and chemical cargo. Variations scenario is a collision between ships that pass by floating containers. The frequency of refloating container, and the consequences of the passing ship depends on several factors, which will be the subject of research. However, because of the frequency of refloating containers is unlikely, then the risk is low and does not pose a danger to navigation. These risk assessment using risk matrix 5x5 which is the combined value of the frequency and consequences of the incident. The results of this study indicate the level of risk, whether the risk is accepted, not accepted or received by considering the costs and benefits (ALARP. To consequence, there are two parameters which energy is absorbed and the penetration occurs. The absorbed energy is divided into two, namely the energy absorbed by ship and the energy absorbed by containers. In this

  16. Development of casting technology and localization for a medical radioisotope transport cask

    International Nuclear Information System (INIS)

    Lee, Y. S.; Kim, H. S.; Jang, S. J.; Seo, K. S.; Kim, C. G.

    2003-01-01

    In order to localize the shielding casks for shipping medical isotopes, this research was carried out. The various casting factors such as the diameter of shielding casting, the temperature of melt and the temperature gradient of a mold were determined with the calculation results of solidification analysis computer code. Through the experiment, the manufacturing method of Ti core was developed to have no defects causing casting failure. As a results of casting experiment, depleted uranium shielding castings were successfully cast without any defect. Also as the results of the radiation shielding capability test, it was good enough to satisfied the standards of transport regulations

  17. Estimation of integrity of cast-iron cask against impact due to free drop test, (1)

    International Nuclear Information System (INIS)

    Itoh, Chihiro

    1988-01-01

    Ductile cast iron is examined to use for shipping and storage cask from a economic point of view. However, ductile cast iron is considered to be a brittle material in general. Therefore, it is very important to estimate the integrity of cast iron cask against brittle failure due to impact load at 9 m drop test and 1 m derop test on to pin. So, the F.E.M. analysis which takes nonlinearity of materials into account and the estimation against brittle failure by the method which is proposed in this report were carried out. From the analysis, it is made clear that critical flaw depth (the minimum depth to initiate the brittle failure) is 21.1 mm and 13.1 mm in the case of 9 m drop test and 1 m drop test on to pin respectively. These flaw depth can be detected by ultrasonic test. Then, the cask is assured against brittle failure due to impact load at 9 m drop test and 1 m drop test on to pin. (author)

  18. Application of the tack weld to cask impact limiter case

    International Nuclear Information System (INIS)

    Ku, J. H.; Choung, W. M.; You, G. S.; Park, S. W.

    2001-01-01

    The objective of this paper is to evaluate the benefit of the application of intermittent tack weld to the cask impact limiter case in the cask impact accident. This paper describes the test results of weldment rupture of foam filled tube type energy absorber and analytical evaluation of the effect of intermittent tack weld to the cask impact limiter case on the cask impact behavior. Prior to the cask impact analysis, the evaluation of weldment joint was carried out for intermittent tack weldment considering the weldment rupture. The intermittent tack welded part is weaker than ordinary weldment so ruptured if the stress exceeds certain limit. The rupture of the impact limiter case causes to lose its constraining effect for the wood blocks, which are filled into the metal incasement between the case and the gussets. The application of intermittent tack weld to the impact limiter case showed great advantage in vertical and horizontal drop impacts

  19. 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

  20. 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

  1. 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.)

  2. Structural analysis of aircraft impact on a nuclear powered ship

    International Nuclear Information System (INIS)

    Dietrich, R.

    1976-01-01

    The paper investigates the aircraft impact on the collision barrier at the side of the ship. The aircraft impact on top of the reactor hatchway is investigated by another analysis. It appears that the most unfavorable angle of impact is always normal to the surface of the collision barrier. Consequently, only normal impact will be considered here. For the specific case of an aircraft striking a nuclear powered ship, the following two effects are considered: Local penetration and dynamic response of the structuure. The local penetration occurs at points where the engines or other rigid objects hit the structure. It is assumed that the aircraft engine is a rigid body projectile and the side wall of the ship is the target. The applied steel penetration formulae for projectiles were empirically derived for military applications, where both the projectile and the target are unlike those of an impact of an aircraft engine. For this reason it is expedient to calculate the upper and the lower limit values of the penetration depths. The results show that the highest penetration depth is less than the sum of all wall thicknesses of the collision barrier. The solution of the dynamic analysis is obtained by using the finite element method. The results are the eigenmodes, the eigenfrequencies, the displacements of the nodes, and the stresses in the applied plane stress elements. It is shown that the maximum stress which only appears in one element is on the same level as the yield stress of the St. 42 steel. The structural analysis shows that the collision barrier is a sufficient safeguard against the perforation of the engine and against the cracking of the structure as a result of the dynamic response to an aircraft impact. (orig./HP) [de

  3. Energy and exergy analysis of a cruise ship

    DEFF Research Database (Denmark)

    Baldi, Francesco; Ahlgren, Fredrik; Nguyen, Tuong-Van

    2015-01-01

    but increasing share of the industry. The complexity of the energy system of a s hip where the energy required by propulsion is no longer the trivial main contributor to the whole energy use thus makes this kind of ship of particular interest for the analysis of how energy is converted from its original form...... destruction is generated in the Diesel engines and in the oil - fired boilers) and in the sea water cooler (5.4%) ; the main exergy losses happen instead in the exhaust gas, mostly from the main engines (6 7% of total losses) and particularly from those not equipped with heat recovery devices . The improved...

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  5. Solutions obtained to international heat transfer benchmarking problems for nuclear fuel casks using Q/TRAN

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-02-01

    In 1985 Sandia National Laboratories participated in the Nuclear Energy Agency Committee on Reactor Physics (NEACRP) Specialists' Meeting on Heat Transfer Assessment of Transportation Packages. The objective of the meeting was to establish a set of model problems for use in comparing the performance of thermal analysis computer codes that may be used in the design of nuclear fuel shipping casks. The selected problems are to be used to compare code results for the thermal phenomena of conduction, convection, and radiation in cask-like problems. Two model problems were used in this study. The first problem required the determination of the steady-state temperatures of a 16 x 16 array of heated and unheated pins (representing fuel and control rod positions) of a simulated PWR fuel assembly. The second problem required the determination of transient temperatures of a finned surface (representing the external surface of a cask) subjected to an internal heat flux and to an external engulfing fire. Solutions to the problems were obtained with the code ''Q/TRAN.'' Solutions and descriptions of the necessary modeling techniques are given in this report

  6. 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)

  7. Finite element analysis of container ship's cargo hold using ANSYS and POSEIDON software

    Science.gov (United States)

    Tanny, Tania Tamiz; Akter, Naznin; Amin, Osman Md.

    2017-12-01

    Nowadays ship structural analysis has become an integral part of the preliminary ship design providing further support for the development and detail design of ship structures. Structural analyses of container ship's cargo holds are carried out for the balancing of their safety and capacity, as those ships are exposed to the high risk of structural damage during voyage. Two different design methodologies have been considered for the structural analysis of a container ship's cargo hold. One is rule-based methodology and the other is a more conventional software based analyses. The rule based analysis is done by DNV-GL's software POSEIDON and the conventional package based analysis is done by ANSYS structural module. Both methods have been applied to analyze some of the mechanical properties of the model such as total deformation, stress-strain distribution, Von Mises stress, Fatigue etc., following different design bases and approaches, to indicate some guidance's for further improvements in ship structural design.

  8. 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.

  9. SMART, Radiation Dose Rates on Cask Surface

    International Nuclear Information System (INIS)

    Yamakoshi, Hisao

    1989-01-01

    1 - Description of program or function: SMART calculates radiation dose rate at the center of each cask surface by using characteristic functions for radiation shielding ability and for radiation current back-scattered from cask wall and cask cavity of each cask, once cask-type is specified. 2 - Method of solution: Matrix Calculation

  10. Econometric analysis of ship life cycles - are safety inspections effective?

    NARCIS (Netherlands)

    G.E. Bijwaard (Govert); S. Knapp (Sabine)

    2008-01-01

    textabstractDue to the shipping industry’s international legal framework and the existence of loopholes in the system, an estimated 5-10 percent of substandard ships exist which are more likely to have incidents with high economic cost. This article uses ship life cycles to provide insight into

  11. 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.)

  12. Model analysis of fresh fuel cask with ANSYS 10.0 software

    International Nuclear Information System (INIS)

    Seyed Aboulfazl Azimfar; Arash Kazemi

    2009-01-01

    The Fresh Fuel for BNPP-1 is due to be transported by special containers which are supposed to be designed in a manner to withstand against stresses and impacts in order to protect the fuel from any possible damage. A static analysis calculates the effects of steady loading conditions on a structure, while ignoring inertia and damping effects, such as those caused by time-varying loads. A static analysis can, however, include steady inertia loads (such as gravity and rotational velocity), and time-varying loads that can be approximated as static equivalent loads. in this paper the computer model of PCS was developed to estimate the safety of the package, in structural static analysis, as well as structural strength of one single or more combined packages to be transported by automobile, rail and air. Safety factor and stresses and strains were calculated by ANSYS software and compared with Russian standards. (Author)

  13. Analysis of Fatigue Crack Growth in Ship Structural Details

    Directory of Open Access Journals (Sweden)

    Leheta Heba W.

    2016-04-01

    Full Text Available Fatigue failure avoidance is a goal that can be achieved only if the fatigue design is an integral part of the original design program. The purpose of fatigue design is to ensure that the structure has adequate fatigue life. Calculated fatigue life can form the basis for meaningful and efficient inspection programs during fabrication and throughout the life of the ship. The main objective of this paper is to develop an add-on program for the analysis of fatigue crack growth in ship structural details. The developed program will be an add-on script in a pre-existing package. A crack propagation in a tanker side connection is analyzed by using the developed program based on linear elastic fracture mechanics (LEFM and finite element method (FEM. The basic idea of the developed application is that a finite element model of this side connection will be first analyzed by using ABAQUS and from the results of this analysis the location of the highest stresses will be revealed. At this location, an initial crack will be introduced to the finite element model and from the results of the new crack model the direction of the crack propagation and the values of the stress intensity factors, will be known. By using the calculated direction of propagation a new segment will be added to the crack and then the model is analyzed again. The last step will be repeated until the calculated stress intensity factors reach the critical value.

  14. Analysis of Ship Groundings on Soft Sea Beds

    DEFF Research Database (Denmark)

    Simonsen, Bo Cerup; Pedersen, Preben Terndrup

    1997-01-01

    it influences the ship heave and pitch motions as well as the friction between the ship and the soil.In this paper a rational calculation model is presented for the sea bed soil reaction forces on the ship bottom. The model is based on the assumption that the penetration of the ship bow generates a flow of pore......The consequences associated with ships running aground depend very much on the soil characteristics of the sea bed and the geometrical shape of the ship bow. The penetration into the sea bed depends on these factors and the penetration is an important factor for the ship motion because...... by the theory of frictional soils in rupture. Frictional stresses on the bow surface are assumed to be related to the normal pressure by a simple Coulumb relation. The total soil reaction as a function of velocity and penetration is found by integration of normal pressure and frictional stresses over...

  15. Investigation of the behaviour of impact limiting devices of transport casks for radioactive materials in the package approval and risk analysis

    International Nuclear Information System (INIS)

    Neumann, Martin

    2009-01-01

    Transport casks for radioactive materials with a Type-B package certificate have to ensure that even under severe accident scenarios the radioactive content remains safely enclosed, in an undercritical arrangement and that ionising radiation is sufficiently shielded. The impact limiter absorbs in an accident scenario the major part of the impact energy and reduces the maximum force applied on the cask body. Therefore the simulation of the behaviour of impact limiting devices of transport casks for nuclear material is of great interest for the design assessment in the package approval as well as for risk analysis in the field of transport of radioactive materials. The behaviour of the impact limiter is influenced by a number of parameters like impact limiter construction, material properties and loading conditions. Uncertainties exist for the application of simplified numerical tools for calculations of impact limiting devices. Uncertainities exist when applying simplified numerical tools. A model describing the compression of wood in axial direction of wood under large deformations for simulation with complex numerical procedures like dynamic Finite Element Methods has not been developed yet. Therefore this thesis concentrates on deriving a physical model for the behaviour of wood and analysing the applicability of different modeling techniques. A model describing the compression of wood in axial direction under large deformations was developed on the basis of an analysis of impact limiter of prototypes of casks for radioactive materials after a 9-m-drop-test and impact tests with wooden specimens. The model describes the softening, which wood under large deformation exhibits, as a function of the lateral strain constraint. The larger the lateral strain restriction, the more energy wood can absorb. The energy absorption capacity of impact limiter depends therefore on the ability of the outer steel sheet structure to prevent wood from evading from the main

  16. Analysis of search and rescue emergency evaluation in ship accidents in Indonesia

    Directory of Open Access Journals (Sweden)

    Arleiny

    2018-01-01

    Full Text Available The objectives og this research is to describe the factors causing ship accident in Indonesia and know the effectiveness of SAR emergency in ship accident in Indonesia. The research method used in this research is qualitative research. Techniques Collection of literature study data and documents. Data validity method using triangulation. Data analysis uses interactive data analysis. The conclusions of this study are Factors that cause the occurrence of ship accidents in Indonesia, among others, the resources of the crew, the eligibility of ships, supporting facilities for shipping, operators, lack of supervision of apparatus, service users and other factors. The high number of ship accidents in Indonesia shows the ineffective implementation of SAR in ship accident in Indonesia.

  17. Materials issues in cask development

    International Nuclear Information System (INIS)

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

    1987-01-01

    The Department of Energy Office of Civilian Radioactive Waste Management (DOE-OCRWM) is chartered by Congress under the Nuclear Waste Policy Act (NWPA) to build a permanent repository for commercial spent nuclear fuel and to provide a supporting transportation system. The OCRWM-sponsored From-Reactor Cask Systems Acquisition Program is developing a family of casks suitable for transporting commercial spent fuel. Phase I of the program is in the process of procuring cask designs for further development and eventual licensing. New materials will probably be proposed for various components of the cask system. This paper identifies potential new materials as a function of their use in the cask (containment, shielding, etc). To the extent that the identified materials are new (not yet qualified for their intended application), this paper identifies probable technical issues and development efforts which may be required to qualify the materials for uses in transportation casks

  18. Alternative cask maintenance facility concepts

    International Nuclear Information System (INIS)

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

    1992-01-01

    In this paper, 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. The third study provides a preliminary concept for adding the capability to repaint the exterior cylindrical portions of casks

  19. 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

  20. 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

  1. Analysis of Real Ship Rolling Dynamics under Wave Excitement Force Composed of Sums of Cosine Functions

    International Nuclear Information System (INIS)

    Zhang, Y. S.; Cai, F.; Xu, W. M.

    2011-01-01

    The ship motion equation with a cosine wave excitement force describes the slip moments in regular waves. A new kind of wave excitement force model, with the form as sums of cosine functions was proposed to describe ship rolling in irregular waves. Ship rolling time series were obtained by solving the ship motion equation with the fourth-order-Runger-Kutta method. These rolling time series were synthetically analyzed with methods of phase-space track, power spectrum, primary component analysis, and the largest Lyapunove exponent. Simulation results show that ship rolling presents some chaotic characteristic when the wave excitement force was applied by sums of cosine functions. The result well explains the course of ship rolling's chaotic mechanism and is useful for ship hydrodynamic study.

  2. Analysis of shield for the nuclear ship MUTSU

    International Nuclear Information System (INIS)

    Fuse, Takayoshi; Takeuchi, Kiyoshi; Yamaji, Akio

    1975-01-01

    On the nuclear ship MUTSU, a higher-than-expected level of radiation was found, with output raised to 1.4 per cent. To investigate the radiation leakage, the analysis of the shielding problem utilized a four-step sequence of PALLAS-2DCY cylindrical r-z calculations with fixed sources distributions in the core. The neutron dose contours show the importance of streaming in the gap between the reactor vessel and the primary shield. Dominant consideration of thermal insulation exclude shielding from this area resulting in an imbalance in the shielding effectiveness. The neutron dose rate at the upper part of the reactor vessel is increased by neutrons incident on the head from cavity scattering. The calculation indicates that the neutron dose rate at the top of the primary shield is 5 rem/hr at 100 per cent output. (auth.)

  3. 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

  4. FACSIM/MRS-1: Cask receiving and consolidation model documentation and user's guide

    International Nuclear Information System (INIS)

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

    1987-06-01

    The Pacific Northwest Laboratory (PNL) has developed a stochastic computer model, FACSIM/MRS, to assist in assessing the operational performance of the Monitored Retrievable Storage (MRS) waste-handling facility. This report provides the documentation and user's guide for the component FACSIM/MRS-1, which is also referred to as the front-end model. The FACSIM/MRS-1 model simulates the MRS cask-receiving and spent-fuel consolidation activities. The results of the assessment of the operational performance of these activities are contained in a second report, FACSIM/MRS-1: Cask Receiving and Consolidation Performance Assessment (Lotz and Shay 1987). The model of MRS canister storage and shipping operations is presented in FACSIM/MRS-2: Storage and Shipping Model Documentation and User's Guide (Huber et al. 1987). The FACSIM/MRS model uses the commercially available FORTRAN-based SIMAN (SIMulation ANalysis language) simulation package (Pegden 1982). SIMAN provides a set of FORTRAN-coded commands, called block operations, which are used to build detailed models of continuous or discrete events that make up the operations of any process, such as the operation of an MRS facility. The FACSIM models were designed to run on either an IBM-PC or a VAX minicomputer. The FACSIM/MRS-1 model is flexible enough to collect statistics concerning almost any aspect of the cask receiving and consolidation operations of an MRS facility. The MRS model presently collects statistics on 51 quantities of interest during the simulation. SIMAN reports the statistics with two forms of output: a SIMAN simulation summary and an optional set of SIMAN output files containing data for use by more detailed post processors and report generators

  5. Analysis of containerized cargo in the ship container terminal

    Energy Technology Data Exchange (ETDEWEB)

    Obhodas, Jasmina, E-mail: jobhodas@irb.h [Ruder Boskovic Institute, Bijenicka c. 54, 10000 Zagreb (Croatia); Sudac, Davorin [Ruder Boskovic Institute, Bijenicka c. 54, 10000 Zagreb (Croatia); Valkovic, Vladivoj [A.C.T.d.o.o., Prilesje 4, 10000 Zagreb (Croatia); Baricevic, Martina [Ruder Boskovic Institute, Bijenicka c. 54, 10000 Zagreb (Croatia); Franulovic, Andrej [Croatian Custom, Alexandera von Humboldta 4, 10000 Zagreb (Croatia); Perot, Bertrand; Carasco, Cedric; Alain, Mariani; Anne-Cecile, Raoux; El Kanawati, Wassila [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France)

    2010-07-21

    The container scanning system based on the elemental analysis of suspect cargo by use of fast 14 MeV neutrons with detection of associated alpha particle has been recently developed under the EU-FP6 'EURITRACK' project. The system is currently set-up in port of Rijeka, Croatia, to be tested and upgraded under the EU-JLS ERITR-C project. Here we present results of analysis performed on 152 screened containers selected by the Croatian Custom and detail descriptive statistical analysis of their shipping manifests. Statistics such as distribution functions of container types, type of packages, type of cargos and average cargo densities were obtained in order to establish the properties of standard cargo traffic. The analyses by using multivariate statistical approach that included principal component analysis (PCA) and between group analysis (BGA) have shown that cargo matrices can be classified as metallic (Fe, Al, Cu, Zn, Ni...), ceramic-glass (Si, O) and organic (C, O, N). In this paper the organic matrices have been analysed in more detail since they are identified as most difficult because of their potential interference with main elements contained in threat materials such as explosives or drugs. Density as a discriminating factor has been included for better recognition of different types of goods.

  6. Validity of scale modeling for large deformations in shipping containers

    International Nuclear Information System (INIS)

    Burian, R.J.; Black, W.E.; Lawrence, A.A.; Balmert, M.E.

    1979-01-01

    The principal overall objective of this phase of the continuing program for DOE/ECT is to evaluate the validity of applying scaling relationships to accurately assess the response of unprotected model shipping containers severe impact conditions -- specifically free fall from heights up to 140 ft onto a hard surface in several orientations considered most likely to produce severe damage to the containers. The objective was achieved by studying the following with three sizes of model casks subjected to the various impact conditions: (1) impact rebound response of the containers; (2) structural damage and deformation modes; (3) effect on the containment; (4) changes in shielding effectiveness; (5) approximate free-fall threshold height for various orientations at which excessive damage occurs; (6) the impact orientation(s) that tend to produce the most severe damage; and (7) vunerable aspects of the casks which should be examined. To meet the objective, the tests were intentionally designed to produce extreme structural damage to the cask models. In addition to the principal objective, this phase of the program had the secondary objectives of establishing a scientific data base for assessing the safety and environmental control provided by DOE nuclear shipping containers under impact conditions, and providing experimental data for verification and correlation with dynamic-structural-analysis computer codes being developed by the Los Alamos Scientific Laboratory for DOE/ECT

  7. Spent fuel transport and storage system for NOK: The TN52L, TN97L, TN24 BHL and TN24 GB casks

    International Nuclear Information System (INIS)

    Wattez, L.; Verdier, A.; Monsigny, P.-A.

    2007-01-01

    NOK nuclear power plants in Switzerland, LEIBSTADT (KKL) BWR nuclear power plant and BEZNAU (KKB) PWR nuclear power plant have opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN52L and TN97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TN24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN52L and TN97L casks by the KKL and ZWILAG operators. In 2004, NOK also ordered three TN24 GB transport and storage casks for PWR fuel types. These casks are presently being manufactured. (author)

  8. Cask and plug handling system design in port cell

    International Nuclear Information System (INIS)

    Martins, Jean-Pierre; Friconneau, Jean-Pierre; Gabellini, Eros; Keller, Delphine; Levesy, Bruno; Selvi, Anna; Tesini, Alessandro; Utin, Yuri; Wagrez, Julien

    2011-01-01

    The ITER maintenance strategy relies partly on the remote transfer of components from vacuum vessel to hot cells. This function will be fulfilled by transfer cask systems. This paper describes the recent design progresses on interfaces in order to increase components handling feasibility by implementing continuous guiding features that avoid cantilevered loads on the in-cask tractor. Also the design has progressed in order to allow generic docking of the casks. When the cask is connected to the port, it becomes part of the machine first confinement boundary, thus it must provide tightness continuity. This high level safety function was one of the main concerns of a finite element analysis study that has been performed to assess the behavior of the whole system. Numerical analysis methodology and results are explained and shown in order to highlight how it has reinforced the knowledge of the system.

  9. Analysis of Methods of Determining the Safe Ship Trajectory

    Directory of Open Access Journals (Sweden)

    Jozef Lisowski

    2016-07-01

    Full Text Available The paper describes six methods of optimal and game theory and artificial neural network for synthesis of safe control in collision situations at sea. The application of optimal and game control algorithms to determine the own ship safe trajectory during the passing of other encountered ships in good and restricted visibility at sea is presented. The comparison of the safe ship control in collision situation: multi-step matrix non-cooperative and cooperative games, multi-stage positional non-cooperative and cooperative games have been introduced. The considerations have been illustrated with examples of computer simulation of the algorithms to determine safe of own ship trajectories in a navigational situation during passing of eight met ships.

  10. Environmental impact analysis of the Nuclear Merchant Ship Program-addendum. Special study

    International Nuclear Information System (INIS)

    1977-06-01

    This study represents a qualitative extension of an earlier Environmental Impact Analysis of the Nuclear Merchant Ship Program (MarAd, 1975) using an Ultra Large Crude Carrier (ULCC) as the reference ship, to other types of ships including a containership, dry-bulk cargo vessel, and an icebreaking oil tanker. This qualitative analysis shows that, for ships operating into coastal ports (containerships and possibly dry-bulk carriers), both ecological and radiological impacts would be somewhat greater than for the reference ULCC docking at an offshore terminal. These impacts can be kept within the guidelines established for land based nuclear power plants by implementation of appropriate operating procedures

  11. 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

  12. Numerical analysis of viscous effect on ship rolling motions based on CFD

    Directory of Open Access Journals (Sweden)

    LUO Tian

    2017-03-01

    Full Text Available During the ship design procedure, the analysis of ship rolling motions is of great significance because the rolling motions have extraordinary effects on the sea-keeping, maneuverability and stability of a ship. It is difficult to simulate rolling motions due to the effect of viscosity, which causes many nonlinear components in computation. As such, the potential theory used for other ship motions cannot be used for rolling motions. This paper simulates the rolling motions of the DTMB 5512 ship model and the ship transverse section of the S60 ship model with a naoe-FOAM-SJTU solver using the Reynolds Averaged Navier Stokes(RANSmethod based on the OpenFOAM. The results of rolling motions are compared with the experimental data, which confirms the reliability of the meshes and results. For the ship transverse section of the S60 ship model, the damping coefficient is divided into three parts with the Euler and RANS methods:friction, vorticity and wave parts. For the DTMB 5512 ship model, the damping coefficient is also respectively analyzed, including the friction, vorticity, wave and bilge keel parts. The results in this paper show that the vorticity part accounts for the greatest proportion, while the friction part accounts for the least, and the bilge keels reduces the damping moment to a certain extent which shows the effect of rolling parameters on rolling motions and moments.

  13. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System AGENCY: Nuclear...] Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No..., Inc. Standardized NUHOMS[supreg] Cask System listing within the ``List of Approved Spent Fuel Storage...

  14. 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

  15. Analysis of International Commodity Shipping Data and the Shipment of NORM to the United States

    Energy Technology Data Exchange (ETDEWEB)

    Baciak, James E.; Ely, James H.; Schweppe, John E.; Sandness, Gerald A.; Robinson, Sean M.

    2011-10-01

    As part of the Spreader Bar Radiation Detector project, PNNL analyzed US import data shipped through US ports collected over the 12 months of 2006 (over 4.5 million containers). Using these data, we extracted a variety of distributions that are of interest to modelers and developers of active and passive detection systems used to 'scan' IMCCs for potential contraband. This report expands on some of the analysis presented in an earlier report from LLNL, by investigation the foreign port distribution of commodities shipped to the US. The majority of containers shipped to the United States are 40 ft containers ({approx}70%); about 25% are 20 ft; and about 3.6% are 45 ft containers. A small fraction (<1%) of containers are of other more specialized sizes, and very few ports actually ship these unique size containers (a full distribution for all foreign ports is shown in Appendix A below). The primary foreign ports that ship the largest fraction of each container are shown in the table below. Given that 45 ft containers comprise 1 of out every 27 containers shipped to the US, and given the foreign ports from which they are shipped, they should not be ignored in screening; further testing and analysis of radiation measurements for national security with this size container is warranted. While a large amount of NORM is shipped in IMCCs, only a few specific commodities are shipped with enough frequency to present potential issues in screening IMCCs at ports. The majority of containers with NORM will contain fertilizers (5,700 containers), granite (59,000 containers), or ceramic (225,000 containers) materials. Fertilizers were generally shipping in either 20- or 40 ft containers with equal frequency. While granite is mostly shipped in 20 ft containers, ceramic materials can be shipped in either 20- or 40 ft containers. The size of container depended on the specific use of the ceramic or porcelain material. General construction ceramics (such as floor and roofing

  16. Analysis of International Commodity Shipping Data and the Shipment of NORM to the United States

    International Nuclear Information System (INIS)

    Baciak, James E.; Ely, James H.; Schweppe, John E.; Sandness, Gerald A.; Robinson, Sean M.

    2011-01-01

    As part of the Spreader Bar Radiation Detector project, PNNL analyzed US import data shipped through US ports collected over the 12 months of 2006 (over 4.5 million containers). Using these data, we extracted a variety of distributions that are of interest to modelers and developers of active and passive detection systems used to 'scan' IMCCs for potential contraband. This report expands on some of the analysis presented in an earlier report from LLNL, by investigation the foreign port distribution of commodities shipped to the US. The majority of containers shipped to the United States are 40 ft containers (∼70%); about 25% are 20 ft; and about 3.6% are 45 ft containers. A small fraction (<1%) of containers are of other more specialized sizes, and very few ports actually ship these unique size containers (a full distribution for all foreign ports is shown in Appendix A below). The primary foreign ports that ship the largest fraction of each container are shown in the table below. Given that 45 ft containers comprise 1 of out every 27 containers shipped to the US, and given the foreign ports from which they are shipped, they should not be ignored in screening; further testing and analysis of radiation measurements for national security with this size container is warranted. While a large amount of NORM is shipped in IMCCs, only a few specific commodities are shipped with enough frequency to present potential issues in screening IMCCs at ports. The majority of containers with NORM will contain fertilizers (5,700 containers), granite (59,000 containers), or ceramic (225,000 containers) materials. Fertilizers were generally shipping in either 20- or 40 ft containers with equal frequency. While granite is mostly shipped in 20 ft containers, ceramic materials can be shipped in either 20- or 40 ft containers. The size of container depended on the specific use of the ceramic or porcelain material. General construction ceramics (such as floor and roofing tiles) tend to

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

    International Nuclear Information System (INIS)

    Michener, T.; Trent, D.S.; Guttmann, J.; Bajwa, C.

    2001-01-01

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

  18. Spent fuel shipping costs for transportation logistics analyses

    International Nuclear Information System (INIS)

    Cole, B.M.; Cross, R.E.; Cashwell, J.W.

    1983-05-01

    Logistics analyses supplied to the nuclear waste management programs of the U.S. Department of Energy through the Transportation Technology Center (TTC) at Sandia National Laboratories are used to predict nuclear waste material logistics, transportation packaging demands, shipping and receiving rates and transportation-related costs for alternative strategies. This study is an in-depth analysis of the problems and contingencies associated with the costs of shipping irradiated reactor fuel. These costs are extremely variable however, and have changed frequently (sometimes monthly) during the past few years due to changes in capital, fuel, and labor costs. All costs and charges reported in this study are based on January 1982 data using existing transport cask systems and should be used as relative indices only. Actual shipping costs would be negotiable for each origin-destination combination

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

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

  20. 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)

  1. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

    Lima, P.; Ribeiro, M.I.; Aparicio, P.

    1998-01-01

    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)

  2. Fluid Distribution Analysis of Kite Sail for Application on Ship

    Directory of Open Access Journals (Sweden)

    . Amiadji

    2017-09-01

    Full Text Available The increasing number of operating ships resulted in high air pollution from the combustion of the ship's engine. Efforts to utilize alternative energy to reduce ship engine work have been done, one of them is using unlimited alternative energy that is wind where one of its application of is the application of new ships sail, kite sail as auxiliary system of ship propulsion . In this final project purposed to find out the value of aerodynamic force of kite sail and power it can generated , with a CFD method that uses 3 kite sail design forms, rectangular, triangular, and elliptical, with an area of 160 m2 this models are simulated at wind speed variations from 13.4 m / s up 15.82 m / s and angel of attack variation of 15.20, and 25. From the variation obtained the total aerodynamic force generated can reach 28.73 kN in rectangular shape, 30.79 kN of Elipsical shape, and 27.55 kN of triangular shape, on variant Angel Of attack 25. From the value of the aerodynamic force, each kite sail capable of generating power, on a rectangular kite sail of up to 263.02 kW, an elipsical 276.75 kW, and a triangular 252.63 kW.

  3. Surface Ship Shock Modeling and Simulation: Two-Dimensional Analysis

    Directory of Open Access Journals (Sweden)

    Young S. Shin

    1998-01-01

    Full Text Available The modeling and simulation of the response of a surface ship system to underwater explosion requires an understanding of many different subject areas. These include the process of underwater explosion events, shock wave propagation, explosion gas bubble behavior and bubble-pulse loading, bulk and local cavitation, free surface effect, fluid-structure interaction, and structural dynamics. This paper investigates the effects of fluid-structure interaction and cavitation on the response of a surface ship using USA-NASTRAN-CFA code. First, the one-dimensional Bleich-Sandler model is used to validate the approach, and second, the underwater shock response of a two-dimensional mid-section model of a surface ship is predicted with a surrounding fluid model using a constitutive equation of a bilinear fluid which does not allow transmission of negative pressures.

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

    International Nuclear Information System (INIS)

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

    1992-04-01

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

  5. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Science.gov (United States)

    2010-01-01

    ... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C... Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Title: Topical Safety Analysis Report for the NAC Storage/Transport Cask for Use at an Independent Spent...

  6. Study on concrete cask for practical use. Heat removal test under normal condition

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Wataru, Masumi; Shirai, Koji; Saegusa, Toshiari

    2005-01-01

    In Japan, it is planed to construct interim storage facilities taking account of dry storage away form reactor in 2010. Recently, a concrete cask is noticed from the economical point of view. But data for its safety analysis have not been sufficient yet. Heat removal tests using to types of full-scale concrete casks were conducted. This paper describes the results under normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced Concrete cask (RC cask) and Concrete Filled Steel cask (CFS cask) were the specimen casks. The levels of decay heat at the initial period of 60 years of storage, the intermediate period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data required for heat removal analyses were obtained. (author)

  7. Intact stability analysis of dead ship conditions using FORM

    DEFF Research Database (Denmark)

    Choi, Ju Hyuck; Jensen, Jørgen Juncher; Kristensen, Hans Otto Holmegaard

    2017-01-01

    The IMO Weather Criterion has proven to be the governing stability criteria regarding minimum GM for e.g. small ferries and large passenger ships. The formulation of the Weather Criterion is based on some empirical relations derived many years ago for vessels not necessarily representative for cu...

  8. Analysis of Basic Transmission Networks for Integrated Ship Control Systems

    DEFF Research Database (Denmark)

    Hansen, T.N.; Granum-Jensen, M.

    1993-01-01

    Description of a computer network for Integrated Ship Control Systems which is going to be developed as part of an EC-project. Today equipment of different make are not able to communicate with each other because most often each supplier of ISC systems has got their own proprietary network.....

  9. Moving the largest capacity PWR dual-purpose cask in the world from Goesgen NPP to the Zwilag interim storage site

    International Nuclear Information System (INIS)

    Delannay, M.; Dudragne, S.

    2002-01-01

    The Swiss Goesgen nuclear power plant (NPP) has decided to use two different methods for the disposal of its spent fuel. (1) To reprocess some of its spent fuel in dedicated facilities. Some of the vitrified waste from the reprocessing will be shipped back to Switzerland using the new COGEMA Logistics, TN81 cask. (2) To ship the other part of its spent fuel to the central interim storage facility of Zwilag (Switzerland) using a COGEMA Logistics dual-purpose TN24G cask. The TN24G is the heaviest and largest dual-purpose cask manufactured so far by COGEMA Logistics in Europe. It is intended for the transport and storage of 37 pressurised water-reactor (PWR) spent fuel assemblies. Four casks were delivered by COGEMA Logistics to Goesgen NPP. Three transports of loaded TN24G casks between Goesgen and Zwilag were successfully performed at the beginning of 2002 with the new COGEMA Logistics Q76 wagon specifically designed to transport heavy casks. This article describes the procedure of operations and shipments for the first TN24G casks up to storage at Zwilag. The fourth transport of loaded TN24G was due to happen in October 2002. The TN24G cask, as part of the TN24 casks family, proved to be a very efficient solution for the KKG spent fuel management. (author)

  10. Ship operation and failure mode analysis using a maneuver simulator

    Science.gov (United States)

    Cabrerizo-Morales, Miguel Angel; Molina, Rafael; de los Santos, Francisco; Camarero, Alberto

    2013-04-01

    In a ship or floating structure operation the agents that contribute to the systems behaviour are not only those derived from fluid-structure interaction, but also the ones linked to mooring-control line set-up evolution and human interaction. Therefore, the analysis of such systems is affected by boundary conditions that change during a complete operation. Frequently, monitoring techniques in laboratory (model) and field (prototype) are based in different instrumental techniques adding difficulty to data comparison and, in some cases, inducing precision and repeatability errors. For this reason, the main aim of this study is to develop the methods and tools to achieve a deep knowledge of those floating systems and obtain capabilities to optimize their operationally thresholds. This abstract presents a methodology and an instrumental system applicable both in field and laboratory: SRECMOCOS Project (Small scale REal-time Caisson MOnitoring and COntrol System). SRECMOCOS compiles three modules. For the monitoring and control of the structure it has been developed a synchronized open and modular microcontroller-based electronic system that comprises sensors, to monitor agents and reactions, and actuators to perform pertinent actions after processing the sensors' data. A secondary objective has been to design and implement a global scaled simulator (1:22), at the 3D basin of The Harbour Research Lab at Technical University of Madrid, in which climatic agents and those derived from the rig/maneuvering setup and the structural design were included. The particular case of Campamento's drydock, in Algeciras Bay (Spain), has been used to apply and validate the methodology. SRECMOCOS Project conjugates control, monitoring and wireless communication systems in a real time basis, offering the possibility to register and simulate all the parameters involved in port operations. This approach offers a step forward into a monitoring strategy to be included in monitoring

  11. Safety analysis report for packaging Lawrence Livermore Laboratories shipping containers

    International Nuclear Information System (INIS)

    Evans, J.H.

    1975-12-01

    The Lawrence Livermore Laboratories shipping containers were designed at Oak Ridge National Laboratory for use in transporting weapons and nuclear components. The design for the containers was evaluated to show compliance with applicable regulations governing packages in which radioactive and fissile materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the containers relative to the standards for the normal conditions of transport. A full-scale container test model was destructively tested to verify compliance with the standards for the accident conditions. The results of the analytical evaluations and the tests demonstrate that the design for the Lawrence Livermore Laboratories shipping containers is in compliance with the applicable regulations

  12. Demonstration of cask transportation and dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Teer, B.R.; Clark, J.

    1984-01-01

    Nuclear Fuel Services, Inc. and the Department of Energy's Idaho Operations Office have signed a cost sharing contract to demonstrate dual purpose shipping and storage casks for spent nuclear fuel. Transnuclear, Inc. has been selected by NFS to design and supply two forged steel casks - one for 40 PWR assemblies from the Ginna reactor, the other for 85 BWR assemblies from the Big Rock Point reactor. The casks will be delivered to West Valley in mid-1985, loaded with the fuel assemblies and shipped by rail to the Idaho National Engineering Laboratory. The shipments will be made under a DOE Certificate of Compliance which will be issued based on reviews by Oak Ridge National Laboratory of Transnuclear's designs

  13. An Economic Analysis of Container Shipping Through Canadian Northwest Passage

    Directory of Open Access Journals (Sweden)

    Dongqin Lu

    2014-12-01

    This paper considers Canada's sovereignty in matters of navigation over the waters at the Arctic Archipelago, in the form of a toll fee for passage usage. We concluded that the NWP has an advantage over the Panama Canal if it is open for free international passage, regardless of ship size. However, if it is not free, its advantages depend on its toll fee. The lower the toll fee is, the more advantages the NWP will boast.

  14. Study of Radiation Shielding Analysis for Low-Intermediate Level Waste Transport Ship

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dohyung; Lee, Unjang; Song, Yangsoo; Kim, Sukhoon; Ko, Jaehoon [Korea Nuclear Engineering and Service Corporation, Seoul (Korea, Republic of)

    2007-07-01

    In Korea, it is planed to transport Low-Intermediate Level Radioactive Waste (LILW) from each nuclear power plant site to Kyongju LILW repository after 2009. Transport through the sea using ship is one of the most prospective ways of LILW transport for current situation in Korea. There are domestic and international regulations for radiation dose limit for radioactive material transport. In this article, radiation shielding analysis for LILW transport ship is performed using 3-D computer simulation code, MCNP. As a result, the thickness and materials for radiation shielding walls next to cargo in the LILW transport ship are determined.

  15. 78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-12-27

    ... Storage Casks: Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System AGENCY: Nuclear Regulatory... storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 11 to Certificate of...

  16. 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

  17. Cask for radioactive material and method for preventing release of neutrons from radioactive material

    International Nuclear Information System (INIS)

    Gaffney, M.F.; Shaffer, P.T.

    1981-01-01

    A cask for radioactive material, such as nuclear reactor fuel or spent nuclear reactor fuel, includes a plurality of associated walled internal compartments for containing such radioactive material, with neutron absorbing material present to absorb neutrons emitted by the radioactive material, and a plurality of thermally conductive members, such as longitudinal copper or aluminum castings, about the compartment and in thermal contact with the compartment walls and with other such thermally conductive members and having thermal contact surfaces between such members extending, preferably radially, from the compartment walls to external surfaces of the thermally conductive members, which surfaces are preferably in the form of a cylinder. The ends of the shipping cask also preferably include a neutron absorber and a conductive metal covering to dissipate heat released by decay of the radioactive material. A preferred neutron absorber utilized is boron carbide, preferably as plasma sprayed with metal powder or as particles in a matrix of phenolic polymer, and the compartment walls are preferably of stainless steel, copper or other corrosion resistant and heat conductive metal or alloy. The invention also relates to shipping casks, storage casks and other containers for radioactive materials in which a plurality of internal compartments for such material, e.g., nuclear reactor fuel rods, are joined together, preferably in modular construction with surrounding heat conductive metal members, and the modules are joined together to form a major part of a finished shipping cask, which is preferably of cylindrical shape. Also within the invention are methods of safely storing radioactive materials which emit neutrons, while dissipating the heat thereof, and of manufacturing the present shipping casks

  18. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    International Nuclear Information System (INIS)

    Audin, L.

    1990-12-01

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs

  19. Numerical prediction of heat-flux to massive calorimeters engulfed in regulatory fires with the cask analysis fire environment (CAFE) model

    International Nuclear Information System (INIS)

    Koski, Jorman A.; Suo-Antitla, Ahti; Kramer M, Alex; Greiner, Miles

    2000-01-01

    Recent observations show that the thermal boundary conditions within large-scale fires are significantly affected by the presence of thermally massive objects. These objects cool the soot and gas near their surfaces, and these effects reduce the incoming radiant heat-flux to values lower than the levels expected from simple σT fire 4 models. They also affect the flow and temperature fields in the fire far from their surfaces. The Cask Analysis Fire Environment (CAFE) code has been developed at Sandia National Laboratories to provide an enhanced fire boundary condition for the design of radioactive material packages. CAFE is a set of computer subroutines that use computational fluid mechanics methods to predict convective heat transfer and mixing. It also includes models for fuel and oxygen transport, chemical reaction, and participating-media radiation heat transfer. This code uses two-dimensional computational models so that it has reasonably short turnaround times on standard workstations and is well suited for design and risk studies. In this paper, CAFE is coupled with a commercial finite-element program to model a large cylindrical calorimeter fully engulfed in a pool fire. The time-dependent heat-flux to the calorimeter and the calorimeter surface temperature are determined for several locations around the calorimeter circumference. The variation of heat-flux with location is determined for calorimeters with different diameters and wall thickness, and the observed effects discussed

  20. Behavior analysis of container ship in maritime accident in order to redefine the operating criteria

    Science.gov (United States)

    Ancuţa, C.; Stanca, C.; Andrei, C.; Acomi, N.

    2017-08-01

    In order to enhance the efficiency of maritime transport, container ships operators proceeded to increase the sizes of ships. The latest generation of ships in operation has 19,000 TEU capacity and the perspective is 21,000 TEU within the next years. The increasing of the sizes of container ships involves risks of maritime accidents occurrences. Nowadays, the general rules on operational security, tend to be adjusted as a result of the evaluation for each vessel. To create the premises for making an informed decision, the captain have to be aware of ships behavior in such situations. Not less important is to assure permanent review of the procedures for operation of ship, including the specific procedures in special areas, confined waters or separation schemes. This paper aims at analysing the behavior of the vessel and the respond of the structure of a container ship in maritime accident, in order to redefine the operating criteria. The method selected by authors for carrying out the research is computer simulations. Computer program provides the responses of the container ship model in various situations. Therefore, the simulations allow acquisition of a large category of data, in the scope of improving the prevention of accidents or mitigation of effects as much as possible. Simulations and assessments of certain situations that the ship might experience will be carried out to redefine the operating criteria. The envisaged scenarios are: introducing of maneuver speed for specific areas with high risk of collision or grounding, introducing of flooding scenarios of some compartments in loading programs, conducting of complex simulations in various situations for each vessel type. The main results of this work are documented proposals for operating criteria, intended to improve the safety in case of marine accidents, collisions and groundings. Introducing of such measures requires complex cost benefit analysis, that should not neglect the extreme economic impact

  1. Ship Power System Analysis Based on Safety Aspects

    Directory of Open Access Journals (Sweden)

    Urbaha Margarita

    2017-08-01

    Full Text Available This article analyses the reasons for the reduction of insulating resistance, processes influencing them and isolation diagnostic methods. It provides a short description of electrical safety situation on ships with isolated neutral electrical power systems. It also covers the methods of protecting personnel from electric shock or preventing ignition or arching damage at the fault location with the help of fault current compensation. Principal fault current compensation circuit diagrams are analysed by using the minimum value and time of transient fault current as criteria.

  2. 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

  3. Stress analysis on passenger deck due to modification from passenger ship to vehicle-carrying ship

    Science.gov (United States)

    Zubaydi, A.; Sujiatanti, S. H.; Hariyanto, T. R.

    2018-03-01

    Stress is a basic concept in learning about material mechanism. The main focus that needs to be brought to attention in analyzing stress is strength, which is the structural capacity to carry or distribute loads. The structural capacity not only measured by comparing the maximum stress with the material’s yield strength but also with the permissible stress required by the Indonesian Classification Bureau (BKI), which certainly makes it much safer. This final project analyzes stress in passenger deck that experiences modification due to load changes, from passenger load to vehicle one, carrying: 6-wheels truck with maximum weight of 14 tons, a passenger car with maximum weight of 3.5 tons, and a motorcycle with maximum weight of 0.4 tons. The deck structure is modelled using finite element software. The boundary conditions given to the structural model are fix and simple constraint. The load that works on this deck is the deck load which comes from the vehicles on deck with three vehicles’ arrangement plans. After that, software modelling is conducted for analysis purpose. Analysis result shows a variation of maximum stress that occurs i.e. 135 N/mm2, 133 N/mm2, and 152 N/mm2. Those maximum stresses will not affect the structure of passenger deck’s because the maximum stress that occurs indicates smaller value compared to the Indonesian Classification Bureau’s permissible stress (175 N/mm2) as well as the material’s yield strength (235 N/mm2). Thus, the structural strength of passenger deck is shown to be capable of carrying the weight of vehicles in accordance with the three vehicles’ arrangement plans.

  4. Development of NESSY (Nuclear ship Engineering Simulation SYstem) and its application to dynamic analysis

    International Nuclear Information System (INIS)

    Kusunoki, T.; Uematsu, H.; Kobayashi, H.

    1992-01-01

    A marine reactor plant sustains incessant load change and the effects of vibration and ship motions due to the maneuvering and dynamic conditions in the marine environment. The change of process variables of the reactor plant is made in accordance with the load change and other effects, and also results in the propeller revolution change and subsequently affects on ship motions. In order to grasp dynamic behavior of the reactor plant in normal operation, including port entry and departure, and also in abnormal conditions such as anticipated transient and accidents, the Nuclear ship Engineering Simulation SYstem (simply ENSSY, hereinafter) carriers out combined analysis in which the behaviors of the ship propulsion, the reactor plant and the secondary systems are simultaneously calculated in each time step. (author)

  5. Reliability analysis of emergency decay heat removal system of nuclear ship under various accident conditions

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi

    1984-01-01

    A reliability analysis is given for the emergency decay heat removal system of the Nuclear Ship ''Mutsu'' and the emergency sea water cooling system of the Nuclear Ship ''Savannah'', under ten typical nuclear ship accident conditions. Basic event probabilities under these accident conditions are estimated from literature survey. These systems of Mutsu and Savannah have almost the same reliability under the normal condition. The dispersive arrangement of a system is useful to prevent the reduction of the system reliability under the condition of an accident restricted in one room. As for the reliability of these two systems under various accident conditions, it is seen that the configuration and the environmental condition of a system are two main factors which determine the reliability of the system. Furthermore, it was found that, for the evaluation of the effectiveness of safety system of a nuclear ship, it is necessary to evaluate its reliability under various accident conditions. (author)

  6. Simulation of the dynamic response of radioactive material shipping package - railcar systems during coupling operations

    International Nuclear Information System (INIS)

    Fields, S.R.

    1981-12-01

    The basic equations of the computer model CARDS (Cask-Railcar Dynamic Simulator), developed for the U.S. Nuclear Regulatory Commission to simulate the dynamic behavior of radioactive material shipping package - railcar systems, are presented. A companion model, CARRS (Casks Railcar Response Spectrum Generator), that generates system response as frequency response spectra is also presented in terms of its basic equations

  7. Simulation of the dynamic response of radioactive material shipping package-railcar systems during coupling operations

    International Nuclear Information System (INIS)

    Fields, S.R.

    1983-10-01

    The basic equations of the computer model CARDS (Cask-Railcar Dynamic Simulator), developed for the US Nuclear Regulatory Commission to simulate the dynamic behavior of radioactive material shipping package - railcar systems, are presented. A companion model, CARRS (Cask Railcar Response Spectrum Generator), that generates system response as frequency response spectra is also presented in terms of its basic equations. 1 reference, 18 figures

  8. 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

  9. 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.

  10. Ship Engine Room Casualty Analysis by Using Decision Tree Method

    Directory of Open Access Journals (Sweden)

    Ömür Yaşar SAATÇİOĞLU

    2017-03-01

    Full Text Available Ships may encounter undesirable conditions during operations. In consequence of a casualty, fire, explosion, flooding, grounding, injury even death may occur. Besides, these results can be avoidable with precautions and preventive operating processes. In maritime transportation, casualties depend on various factors. These were listed as misuse of the engine equipment and tools, defective machinery or equipment, inadequacy of operational procedure and measure of safety and force majeure effects. Casualty reports which were published in Australia, New Zealand, United Kingdom, Canada and United States until 2015 were examined and the probable causes and consequences of casualties were determined with their occurrence percentages. In this study, 89 marine investigation reports regarding engine room casualties were analyzed. Casualty factors were analyzed with their frequency percentages and also their main causes were constructed. This study aims to investigate engine room based casualties, frequency of each casualty type and main causes by using decision tree method.

  11. Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, James E.; Askew, Neal M.; Hensel, Steve J.

    2013-03-21

    Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

  12. 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.

  13. 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.

  14. Simulation-Based Analysis of Ship Motions in Short-Crested Irregular Seas

    Directory of Open Access Journals (Sweden)

    Kıvanç Ali ANIL

    2017-03-01

    Full Text Available Demonstration of the seakeeping calculation results other than polar diagrams and Cartesian plots is important during the initial and detail design stages of naval platforms due to the necessity of numerical simulations (time series data for the design and validation of the systems on board. These time series simulations are called as “real time computer experiments”. Similar simulation algorithms for ship motions and wave elevation are also used by ship-handling simulators for realistic visualization. The goal of this paper is to create a basis for the simulation-based analysis of ship motions and wave elevation for future design and validation studies for both the naval platform itself and the systems on board. The focus of this paper is the clarification of the theoretical background of this process, i.e. all formulations required to create and validate a ship motion and wave surface simulation are given in detail. The results of this study may also be used in ship-handling simulators or helicopter landing on ship simulations.

  15. Construction strength analysis of landing craft tank conversion to passenger ship using finite element method

    Science.gov (United States)

    Nurul Misbah, Mohammad; Setyawan, Dony; Murti Dananjaya, Wisnu

    2018-03-01

    This research aims to determine the longitudinal strength of passenger ship which was converted from Landing Craft Tank with 54 m of length as stated by BKI (Biro Klasifikasi Indonesia / Indonesian Classification Bureau). Verification of strength value is done to 4 (four) loading conditions which are (1) empty load condition during sagging wave, (2) empty load condition during hogging wave, (3) full load condition during sagging wave and (4) full load condition during hogging wave. Analysis is done using Finite Element Analysis (FEA) software by modeling the entire part of passenger ship and its loading condition. The back and upfront part of ship centerline were used as the boundary condition. From that analysis it can be concluded that the maximum stress for load condition (1) is 72,393 MPa, 74,792 MPa for load condition (2), 129,92 MPa for load condition (3), and 132,4 MPa for load condition (4). Longitudinal strength of passenger ship fulfilled the criteria of empty load condition having smaller stress value than allowable stress which is 90 MPa, and during full load condition with smaller stress value than allowable stress which is 150 MPa. Analysis on longitudinal strength comparison with entire ship plate thickness variation of ± 2 mm from initial plate was also done during this research. From this research it can be concluded that plate thickness reduction causes the value of longitudinal strength to decrease, while plate thickness addition causes the value of longitudinal strength to increase.

  16. Estimates of fire environments in ship holds containing radioactive material packages

    International Nuclear Information System (INIS)

    Koski, J.A.; Cole, J.K.; Hohnstreiter, G.F.; Wix, S.D.

    1995-01-01

    Fire environments that occur on cargo ships differ significantly from the fire environments found in land transport. Cargo ships typically carry a large amount of flammable fuel for propulsion and shipboard power, and may transport large quantities of flammable cargo. As a result, sea mode transport accident records contain instances of long lasting and intense fires. Since Irradiated Nuclear Fuel (INF) casks are not carried on tankers with large flammable cargoes, most of these dramatic, long burning fires are not relevant threats, and transport studies must concentrate on those fires that are most likely to occur. By regulation, INF casks must be separated from flammable cargoes by a fire-resistant, liquid-tight partition. This makes a fire in an adjacent ship hold the most likely fire threat. The large size of a cargo ship relative to any spent nuclear fuel casks on board, however, may permit a severe, long lasting fire to occur with little or no thermal impact on the casks. Although some flammable materials such as shipping boxes or container floors may exist in the same hold with the cask, the amount of fuel available may not provide a significant threat to the massive transport casks used for radioactive materials. This shipboard fire situation differs significantly from the regulatory conditions specified in 10 CFR 71 for a fully engulfing pool fire. To learn more about the differences, a series of simple thermal analyses has been completed to estimate cask behavior in likely marine and land thermal accident situations. While the calculations are based on several conservative assumptions, and are only preliminary, they illustrate that casks are likely to heat much more slowly in shipboard hold fires than in an open pool fire. The calculations also reinforce the basic regulatory concept that for radioactive materials, the shipping cask, not the ship, is the primary protection barrier to consider

  17. A Method for Ship Collision Damage and Energy Absorption Analysis and its Validation

    DEFF Research Database (Denmark)

    Zhang, Shengming; Pedersen, Preben Terndrup

    2017-01-01

    For design evaluation, there is a need for a method which is fast, practical and yet accurate enough to deter-mine the absorbed energy and collision damage extent in ship collision analysis. The most well-known sim-plified empirical approach to collision analysis was made probably by Minorsky......, and its limitation is alsowell-recognised. The authors have previously developed simple expressions for the relation between theabsorbed energy and the damaged material volume which take into account the structural arrangements,the material properties and the damage modes. The purpose of the present paper...... is to re-examine thismethod’s validity and accuracy for ship collision damage analysis in ship design assessments by compre-hensive validations with experimental results from the public domain. In total, 20 experimental tests havebeen selected, analysed and compared with the results calculated using...

  18. TMI-2 [Three-Mile Island-Unit 2] rail cask and railcar maintenance

    International Nuclear Information System (INIS)

    Tyacke, M.J.; Ayers, A.L. Jr.; Ball, L.J.; Anselmo, A.A.

    1988-02-01

    This paper describes the NuPac 125-B cask system (i.e., cask and railcar), and the maintenance and inspection requirements for that system. The paper discusses the operations being done to satisfy those requirements and how, in some cases, it has been efficient for the operations to be more rigorous than the requirements. Finally, this paper discusses the experiences gained from those operations and how specific hardware and procedural enhancements have resulted in a reliable and continuous shipping campaign. 2 refs., 2 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  20. Thermoelectric Powered Wireless Sensors for Dry-Cask Storage

    Science.gov (United States)

    Carstens, Thomas Alan

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

  1. Fire resistant nuclear fuel cask

    International Nuclear Information System (INIS)

    Heckman, R.C.; Moss, M.

    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

  2. Uncertainty Analysis of Resistance Tests in Ata Nutku Ship Model Testing Laboratory of Istanbul Technical University

    Directory of Open Access Journals (Sweden)

    Cihad DELEN

    2015-12-01

    Full Text Available In this study, some systematical resistance tests, where were performed in Ata Nutku Ship Model Testing Laboratory of Istanbul Technical University (ITU, have been included in order to determine the uncertainties. Experiments which are conducted in the framework of mathematical and physical rules for the solution of engineering problems, measurements, calculations include uncertainty. To question the reliability of the obtained values, the existing uncertainties should be expressed as quantities. The uncertainty of a measurement system is not known if the results do not carry a universal value. On the other hand, resistance is one of the most important parameters that should be considered in the process of ship design. Ship resistance during the design phase of a ship cannot be determined precisely and reliably due to the uncertainty resources in determining the resistance value that are taken into account. This case may cause negative effects to provide the required specifications in the latter design steps. The uncertainty arising from the resistance test has been estimated and compared for a displacement type ship and high speed marine vehicles according to ITTC 2002 and ITTC 2014 regulations which are related to the uncertainty analysis methods. Also, the advantages and disadvantages of both ITTC uncertainty analysis methods have been discussed.

  3. Reliability analysis of production ships with emphasis on load combination and ultimate strength

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiaozhi

    1995-05-01

    This thesis deals with ultimate strength and reliability analysis of offshore production ships, accounting for stochastic load combinations, using a typical North Sea production ship for reference. A review of methods for structural reliability analysis is presented. Probabilistic methods are established for the still water and vertical wave bending moments. Linear stress analysis of a midships transverse frame is carried out, four different finite element models are assessed. Upon verification of the general finite element code ABAQUS with a typical ship transverse girder example, for which test results are available, ultimate strength analysis of the reference transverse frame is made to obtain the ultimate load factors associated with the specified pressure loads in Det norske Veritas Classification rules for ships and rules for production vessels. Reliability analysis is performed to develop appropriate design criteria for the transverse structure. It is found that the transverse frame failure mode does not seem to contribute to the system collapse. Ultimate strength analysis of the longitudinally stiffened panels is performed, accounting for the combined biaxial and lateral loading. Reliability based design of the longitudinally stiffened bottom and deck panels is accomplished regarding the collapse mode under combined biaxial and lateral loads. 107 refs., 76 refs., 37 tabs.

  4. 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.

  5. 78 FR 73456 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ...-2012-0052] RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment... International HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to... requirements for the HI-STORM 100U part of the HI-STORM 100 Cask System and updates the thermal model and...

  6. Strength and water-tightness of the closure head and valves of a model cask under high external pressure

    International Nuclear Information System (INIS)

    Terada, O.; Kumada, M.; Hayakawa, T.; Mochizuki, S.; Ohrui, K.

    1978-01-01

    This paper describes experimental research on the strength and water-tightness of the closure head and attached valves of a model cask under high external pressure, in simulation of its having been accidentally lost in the deep sea. Both the external pressure tests and the corrosion tests were carried out using scale models of the closure head of an 80-ton spent-fuel shipping cask, and the full size pressure relief valves and drain valves which were to be attached to the actual cask. Based on the results of the above tests, evaluations were made, and new information was obtained on the pressure-proof strength and water-tightness of the closure head of the cask and the valves. Lastly, research which is being carried on in Japan on the pressure equalizer is also introduced

  7. 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

  8. The influence of costs and benefits' analysis on service strategy formulation: Learnings from the shipping industry

    DEFF Research Database (Denmark)

    Pagoropoulos, Aris; Kjær, Louise Laumann; Andersen, Jakob Axel Bejbro

    2017-01-01

    . This paper assesses how the analysis of costs and benefits of Product-Service Systems (PSS) as servitized offerings influences the formulation of service strategies in the shipping industry. The study examines both the manufacturer and customer perspectives using two case studies from the shipping sector....... Life Cycle Costing (LCC) was used as a tool to assess the associated costs and benefits of two proposed PSS. Based on the results of the LCC, the drivers and barriers of the actual transformation processes were explored through workshops and interviews served to map the perspectives of both...

  9. 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.)

  10. The simplified interaction tool for efficient and accurate underwater shock analysis of naval ships

    NARCIS (Netherlands)

    Aanhold, J.E. van; Trouwborst, W.; Vaders, J.A.A.

    2014-01-01

    In order to satisfy the need for good quality UNDEX response estimates of naval ships, TNO developed the Simplified Interaction Tool (SIT) for underwater shock analysis. The SIT is a module of user routines linked to LS-DYNA, which generates the UNDEX loading on the wet hull of a 3D finite element

  11. Evaluation of RAPID for a UNF cask benchmark problem

    Science.gov (United States)

    Mascolino, Valerio; Haghighat, Alireza; Roskoff, Nathan J.

    2017-09-01

    This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for the simulation of a used nuclear fuel (UNF) cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days). A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.

  12. Evaluation of RAPID for a UNF cask benchmark problem

    Directory of Open Access Journals (Sweden)

    Mascolino Valerio

    2017-01-01

    Full Text Available This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection code system for the simulation of a used nuclear fuel (UNF cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days. A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.

  13. Shielding analysis method applied to nuclear ship 'MUTSU' and its evaluation based on experimental analyses

    International Nuclear Information System (INIS)

    Yamaji, Akio; Miyakoshi, Jun-ichi; Iwao, Yoshiaki; Tsubosaka, Akira; Saito, Tetsuo; Fujii, Takayoshi; Okumura, Yoshihiro; Suzuoki, Zenro; Kawakita, Takashi.

    1984-01-01

    Procedures of shielding analysis are described which were used for the shielding modification design of the Nuclear Ship ''MUTSU''. The calculations of the radiation distribution on board were made using Sn codes ANISN and TWOTRAN, a point kernel code QAD and a Monte Carlo code MORSE. The accuracies of these calculations were investigated through the analysis of various shielding experiments: the shield tank experiment of the Nuclear Ship ''Otto Hahn'', the shielding mock-up experiment for ''MUTSU'' performed in JRR-4, the shielding benchmark experiment using the 16 N radiation facility of AERE Harwell and the shielding effect experiment of the ship structure performed in the training ship ''Shintoku-Maru''. The values calculated by the ANISN agree with the data measured at ''Otto Hahn'' within a factor of 2 for fast neutrons and within a factor of 3 for epithermal and thermal neutrons. The γ-ray dose rates calculated by the QAD agree with the measured values within 30% for the analysis of the experiment in JRR-4. The design values for ''MUTSU'' were determined in consequence of these experimental analyses. (author)

  14. 78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

    ... Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY: Nuclear Regulatory... storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the...

  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. 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

  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. 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.

  20. Inspection of NFT-type cask fabrication

    International Nuclear Information System (INIS)

    Takani, M.; Umegaki, O.

    1998-01-01

    NFT-type cask has been developed to transport the high burn-up spent fuel from Japanese nuclear power stations to the reprocessing plant of Japan Nuclear Fuel Limited which is under construction in Rokkasho-mura, Aomori prefecture. NFT placed orders of 53 casks to 5 fabricators in Japan and overseas, and these casks have been fabricated since 1994. There are two types of NFT-type casks for PWR spent fuel and four types of NFT-type cask for BWR spent fuel. These are designed in consideration of the number of spent fuels accommodated into each type of casks and the handling conditions at domestic nuclear power stations. According to Japanese notification, it is required to be confirmed by competent authority that casks are manufactured in accordance with approved designs. Furthermore, additional tests are performed such as through-gauge test for basket and pressure test on the shielding material space to ensure the performance of cask by NFT other than items inspected by the competent authority. In order to enhance maintainability of casks, replacement parts such as bolts and valves are shared as much as possible. (authors)

  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. 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

  3. 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...

  4. Noise source analysis of nuclear ship Mutsu plant using multivariate autoregressive model

    International Nuclear Information System (INIS)

    Hayashi, K.; Shimazaki, J.; Shinohara, Y.

    1996-01-01

    The present study is concerned with the noise sources in N.S. Mutsu reactor plant. The noise experiments on the Mutsu plant were performed in order to investigate the plant dynamics and the effect of sea condition and and ship motion on the plant. The reactor noise signals as well as the ship motion signals were analyzed by a multivariable autoregressive (MAR) modeling method to clarify the noise sources in the reactor plant. It was confirmed from the analysis results that most of the plant variables were affected mainly by a horizontal component of the ship motion, that is the sway, through vibrations of the plant structures. Furthermore, the effect of ship motion on the reactor power was evaluated through the analysis of wave components extracted by a geometrical transform method. It was concluded that the amplitude of the reactor power oscillation was about 0.15% in normal sea condition, which was small enough for safe operation of the reactor plant. (authors)

  5. RISK ANALYSIS FOR SHIP CONVERTING PROJECT ACCOMPLISHMENT (Case study of KRI KP Converting Project

    Directory of Open Access Journals (Sweden)

    Dimas Endro W

    2013-10-01

    Full Text Available Ship converting has become as prospective activity in ship building area. Operational and economical aspect are the most dominant rationale. Baseon a new fuction of converted ship, a task list which contain several jobs that must be done is listed. This accomplishment schedule not only contain a task list, but also duration for certain job title. In practical apllication job duration is maintained based on experience of project manager.  Further more, total accomplish duration is setted as time accomplishment for the project. This setted time has become reference for the project bid. Occasionaly, if accomplishment time which offered is strict, than schedule slip become as potencial nightmare. For this situation, project manager has had a cristal clearconsideration to select a proper decision wheter he will take the tender offer or not. practically, project mananger has layed on his experience to handle previous project and face  penalty if the project delayed. This paper focussed on how to measure tender offer based on risk analysis, specially for converted ship tender which has a strike time accomplishment.A new application method to analysis proposed tender based on time and penalty parameter has become a topic of this paper.

  6. Sensitivity analysis of parameters important to nuclear criticality safety of Castor X/28F spent nuclear fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Leotlela, Mosebetsi J. [Witwatersrand Univ., Johannesburg (South Africa). School of Physics; Koeberg Operating Unit, Johannesburg (South Africa). Regulations and Licensing; Malgas, Isaac [Koeberg Nuclear Power Station, Duinefontein (South Africa). Nuclear Engineering Analysis; Taviv, Eugene [ASARA consultants (PTY) LTD, Johannesburg (South Africa)

    2015-11-15

    In nuclear criticality safety analysis it is essential to ascertain how various components of the nuclear system will perform under certain conditions they may be subjected to, particularly if the components of the system are likely to be affected by environmental factors such as temperature, radiation or material composition. It is therefore prudent that a sensitivity analysis is performed to determine and quantify the response of the output to variation in any of the input parameters. In a fissile system, the output parameter of importance is the k{sub eff}. Therefore, in attempting to prevent reactivity-induced accidents, it is important for the criticality safety analyst to have a quantified degree of response for the neutron multiplication factor to perturbation in a given input parameter. This article will present the results of the perturbation of the parameters that are important to nuclear criticality safety analysis and their respective correlation equations for deriving the sensitivity coefficients.

  7. Structural analysis for diagnosis with application to ship propulsion problem

    DEFF Research Database (Denmark)

    Izadi-Zamanabadi, Roozbeh; Blanke, Mogens

    2002-01-01

    Aiming at design of algorithms for fault diagnosis, structural analysis of systems offers concise yet easy overall analysis. Graph-based matching, which is the essential tech-nique to obtain redundant information for diagnosis, is reconsidered in this paper. Matching is reformulated as a problem...... of relating faults to known parameters and measurements of a system. Using explicit fault modelling, minimal overdetermined subsystems are shown to provide necessary redundancy relations from the matching. Details of the method are presented and a realistic example used to clearly describe individual steps....

  8. Interim and final storage casks

    International Nuclear Information System (INIS)

    Stumpfrock, L.; Kockelmann, H.

    2012-01-01

    The disposal of radioactive waste is a huge social challenge in Germany and all over the world. As is well known the search for a site for a final repository for high-level waste in Germany is not complete. Therefore, interim storage facilities for radioactive waste were built at plant sites in Germany. The waste is stored in these storage facilities in appropriate storage and transport casks until the transport in a final repository can be carried out. Licensing of the storage and transport casks aimed for use in the public space is done according to the traffic laws and for handling in the storage facility according to nuclear law. Taking into account the activity of the waste to be stored, different containers are in use, so that experience is available from the licensing and operation in interim storage facilities. The large volume of radioactive waste to be disposed of after the shut-down of power generation in nuclear power stations makes it necessary for large quantities of licensed storage and transport casks to be provided soon.

  9. Investigation of the behaviour of impact limiting devices of transport casks for radioactive materials in the package approval and risk analysis; Untersuchung des Verhaltens stossdaempfender Bauteile von Transportbehaeltern fuer radioaktive Stoffe in Bauartpruefung und Risikoanalyse

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Martin

    2009-08-20

    Transport casks for radioactive materials with a Type-B package certificate have to ensure that even under severe accident scenarios the radioactive content remains safely enclosed, in an undercritical arrangement and that ionising radiation is sufficiently shielded. The impact limiter absorbs in an accident scenario the major part of the impact energy and reduces the maximum force applied on the cask body. Therefore the simulation of the behaviour of impact limiting devices of transport casks for nuclear material is of great interest for the design assessment in the package approval as well as for risk analysis in the field of transport of radioactive materials. The behaviour of the impact limiter is influenced by a number of parameters like impact limiter construction, material properties and loading conditions. Uncertainties exist for the application of simplified numerical tools for calculations of impact limiting devices. Uncertainities exist when applying simplified numerical tools. A model describing the compression of wood in axial direction of wood under large deformations for simulation with complex numerical procedures like dynamic Finite Element Methods has not been developed yet. Therefore this thesis concentrates on deriving a physical model for the behaviour of wood and analysing the applicability of different modeling techniques. A model describing the compression of wood in axial direction under large deformations was developed on the basis of an analysis of impact limiter of prototypes of casks for radioactive materials after a 9-m-drop-test and impact tests with wooden specimens. The model describes the softening, which wood under large deformation exhibits, as a function of the lateral strain constraint. The larger the lateral strain restriction, the more energy wood can absorb. The energy absorption capacity of impact limiter depends therefore on the ability of the outer steel sheet structure to prevent wood from evading from the main

  10. Review and Application of Ship Collision and Grounding Analysis Procedures

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup

    2010-01-01

    It is the purpose of the paper to present a review of prediction and analysis tools for collision and grounding analyses and to outline a probabilistic procedure for which these tools can be used by the maritime industry to develop performance based rules to reduce the risk associated with human,......, environmental and economic costs of collision and grounding events. The main goal of collision and grounding research should be to identify the most economic risk control options associated with prevention and mitigation of collision and grounding events....

  11. 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

  12. Use of commercial robotics in radioactive waste shipping and receiving

    International Nuclear Information System (INIS)

    Berger, J.D.

    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. Unacceptable personnel exposure to ionizing radiation is projected to occur if current limits for radiation levels at the cask surface and current handling methods are used. To reduce these exposure levels, alternate handling methods are being developed and demonstrated. The production nature of cask receiving operations suggests commercial robotics be incorporated into a remote handling system to reduce predicted worker exposure to acceptable levels, while maintaining or increasing throughput. The first phase of cask handling system development culminated in a proof-of-principle test demonstrating the feasibility of performing cask receiving and unloading operations in a remote and partially automated manner. 6 refs., 12 figs

  13. 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

  14. Method for handling nuclear fuel casks

    International Nuclear Information System (INIS)

    Weems, S.J.

    1976-01-01

    A heavy shielded nuclear fuel cask is lowered into and removed from a water filled spent fuel pool by providing a vertical guide tube in the pool, affixing to the bottom of the cask a base plate that approximates the transverse dimension of the guide tube, and lowering and elevating the cask and base plate assembly into and out of the pool by causing it to traverse within the guide tube. The guide tube and base plate coact to function as a dashpot, thereby cushioning and controlling the fall of the cask in the pool should it break loose while being lowered into or raised out of the pool. a specified approach path to the guide tube insures that the cask assembly will not fall into the pool, should it break loose on its approach to the guide tube

  15. 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

  16. 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

  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. 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.

  19. 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

  20. Assessment of hygiene standards and Hazard Analysis Critical Control Points implementation on passenger ships.

    Science.gov (United States)

    Mouchtouri, Varavara; Malissiova, Eleni; Zisis, Panagiotis; Paparizou, Evina; Hadjichristodoulou, Christos

    2013-01-01

    The level of hygiene on ferries can have impact on travellers' health. The aim of this study was to assess the hygiene standards of ferries in Greece and to investigate whether Hazard Analysis Critical Control Points (HACCP) implementation contributes to the hygiene status and particularly food safety aboard passenger ships. Hygiene inspections on 17 ferries in Greece were performed using a standardized inspection form, with a 135-point scale. Thirty-four water and 17 food samples were collected and analysed. About 65% (11/17) of ferries were scored with >100 points. Ferries with HACCP received higher scores during inspection compared to those without HACCP (p value food samples, only one was found positive for Salmonella spp. Implementation of management systems including HACCP principles can help to raise the level of hygiene aboard passenger ships.

  1. 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.

  2. Feasibility Study and Cost Benefit Analysis of Thin-Client Computer System Implementation Onboard United States Navy Ships

    National Research Council Canada - National Science Library

    Arbulu, Timothy D; Vosberg, Brian J

    2007-01-01

    The purpose of this MBA project was to conduct a feasibility study and a cost benefit analysis of using thin-client computer systems instead of traditional networks onboard United States Navy ships...

  3. Combined Thermal Management and Power Generation Concept for the Spent Fuel Dry Storage Cask

    International Nuclear Information System (INIS)

    Kim, In Guk; Bang, In Cheol

    2017-01-01

    The management of the spent nuclear fuel generated by nuclear power plants is a major issue in Korea due to insufficient capacity of the wet storage pools. Therefore, it is considered that dry storage system is the one possible solution for storing spent fuel. A dual-purpose metal cask (transportation and storage) is currently developing in Korea. This cask has 21 of fuel assemblies and 16.8 kW of maximum decay heat. To evaluate the critical safety in normal/off normal and accident conditions, critical stabilities were conducted by using CSAS 6.0. The experimental investigation of heat removal of a concrete storage cask was also conducted under normal, off normal and accident conditions. The results of the evaluation showed a good safety of the dry storage cask. The results showed the enhanced thermal performance according to modification of flow rate. To verify combined thermal management and power generation concept, a new type of test facility for dry storage cask was designed in 1/10 scale of concrete dry storage cask. The experimental study involved the cooling methods that are an integrated system on the top of the dry cask and air flow path on the canister wall. The results showed the temperature distribution of the wall and inside of the dry cask at the normal condition. The influence of the change of the heat load and cooling system were investigated. The heat removal by the integrated system is approximately 20% of the total heat removal of the dry cask with reduced wall temperature. In these tests, economic analysis is conducted by applying the concept of the cost and efficiency. Under different decay power cases, the energy efficiency of the heat pipe and Stirling engine are determined and compared based on experimental results. The average efficiencies of the Stirling engine were the range of 2.375 to 3.247% under the power range of 35– 65W. These results showed that advanced dry storage concept had a better cooling performance in comparison with

  4. Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

    International Nuclear Information System (INIS)

    Zhang Yanzhao; Zhang Fan; Zhao Xinwen; Zheng Yingfeng

    2013-01-01

    Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

  5. 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

  6. Cask manufacturing methods and quality assurance problems

    International Nuclear Information System (INIS)

    Riddle, J.H.; Cross, H.D.; Trujillo, A.A.; Pope, R.B.; Rack, H.J.

    1980-01-01

    This paper presents the results of a study performed to identify and evaluate cask manufacturing methods and materials which are presently or could be readily available in the United States. The study includes a comparison of the economic and technical advantages to be gained from manufacturing casks specifically for application to long-cooled fuel (5 years or greater) and waste materials. The conclusions that have been drawn from this study, which is still in progress are as follows: if spent fuel payloads are in fact cooled for five or more years prior to transportation there is a significant advantage to be gained from cask designs which are tailored to this longer cooling time; the most economic cask design is that which utilizes a lead gamma shield; the all-steel, single wall casks show substantial potential benefits resulting from ready availability and the cost and inspection advantages that result from the design simplicity; the future use of depleted uranium shielded casks in the United States is expected to be low as a result of the high cost and lack of manufacturing facilities; of the three types of neutron absorber systems evluated, the borated cast-in-place silicone rubber shows significant technical and cost advantages; and design and fabrication code requirements and regulatory requirements yet to be developed in the United States are likely to have a profound effect on the next generation of transportation casks

  7. Burnup credit for storage and transportation casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1988-01-01

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

  8. 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

  9. Conceptual design report for a remotely operated cask handling system. Revision 1

    International Nuclear Information System (INIS)

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

    1984-09-01

    Recent advances in remote handling utilizing commercial robotics are conceptually applied to lowering operator cumulative radiation exposure and increasing throughput during cask handling operations in nuclear shipping and receiving facilities. Revision 1 incorporates functional criteria for facility equipment, equipment technical outline specifications, and interface control drawings to assist Architect Engineers in the application of remote handling to waste shipping and receiving facilities. The document has also been updated to show some of the equipment used in proof-of-principle testing during fiscal year 1984. 10 references, 50 figures, 1 table

  10. Finite element analysis of propellant of solid rocket motor during ship motion

    Directory of Open Access Journals (Sweden)

    Kai Qu

    2013-03-01

    Full Text Available In order to simulate the stress and strain of solid rocket motors (SRMs, a finite element analysis model was established. The stress spectra of the SRM elements with respect to time in the case that the vessel cruises under a certain shipping condition were obtained by simulation. According to the analysis of the simulation results, a critical zone was confirmed, and the Mises stress amplitudes of the different critical zones were acquired. The results show that the maximum stress and strain of SRM are less than the maximum tensile strength and elongation, respectively, of the propellant. The cumulative damage of the motor must also be evaluated by random fatigue loading.

  11. A solution for the integration of finite element analysis in a ship design environment

    International Nuclear Information System (INIS)

    Finifter, D.; Wyniecki, P.; Castel, J. de

    1982-01-01

    The DEMAIN system for the pre- and post-processing of finite element analyses of ship structures is presented. It is shown that this new modelling concept, although being self-contained and specialized, has features which relate it to computer-aided design applications of a more general nature. Thus, compared to other finite element pre/post-processors, it allows a more natural occurrence of the structural analysis task in the design flow and can be considered a major step towards an integrated design and analysis system. (orig.)

  12. Application of a fracture toughness analysis for ferritic steel components of transport/storage casks using an adapted EUROCODE 3 approach

    International Nuclear Information System (INIS)

    Sedlacek, G.; Dahl, W.; Langenberg, P.; Stranghoner, N.; Dreier, G.; Diersch, R.

    1998-01-01

    The choice of a structural steel material for a component of a transport and storage cask for RAM is strongly determined by the demand that the cask has to withstand a free fall from 9 m height without losing its integrity and leak tightness. In terms of fracture mechanics this means that instable crack growth must not occur even under the conditions of high amplitude dynamic loading at temperatures of -40 deg. C. In the course of harmonization of European design guidelines, the Eurocode 3 has been developed which contains a fracture mechanic based concept for the steel selection to avoid brittle fracture, called Annex C. This method combines fracture mechanics tools like the failure assessment diagram (CEGB-R6-procedure) with fracture mechanics life time assessment procedure for fatigue loaded structures. The required toughness in terms of the stress intensity factor K I is related to the T 27J Charpy transition temperature by means of a master curve and by a correlation between the fracture mechanics transition temperature T K100 and the Charpy transition temperature T 27J . Both relations have been proved to be valid for structural steels in the range of 235 to 960 MPa yield strength. Besides that a semi-probabilistic safety approach that takes account of the model inaccuracies by calibration of large scale tests has been applied to derive a safety element for a risk of failure of p f 10 -5 . The fracture mechanic concept of Eurocode 3 has been adopted to calculate critical failure lengths for lids made from ferritic steels of transport and storage casks. The safety requirements of Appendix VI of the IAEA Advisory Material have been taken into account. It has been shown that the adopted Eurocode 3, Annex C, method allows an economical calculation of critical failure length on a high level of safety. A failure probability of p f = 8 . 10 -7 is reached by applying lower bound estimates of fracture toughness and an additional additive safety factor ΔT a of 20 deg

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  14. Damping layout optimization for ship's cabin noise reduction based on statistical energy analysis

    Directory of Open Access Journals (Sweden)

    WU Weiguo

    2017-08-01

    Full Text Available An optimization analysis study concerning the damping control of ship's cabin noise was carried out in order to improve the effect and reduce the weight of damping. Based on the Statistical Energy Analysis (SEA method, a theoretical deduction and numerical analysis of the first-order sensitivity analysis of the A-weighted sound pressure level concerning the damping loss factor of the subsystem were carried out. On this basis, a mathematical optimization model was proposed and an optimization program developed. Next, the secondary development of VA One software was implemented through the use of MATLAB, while the cabin noise damping control layout optimization system was established. Finally, the optimization model of the ship was constructed and numerical experiments of damping control optimization conducted. The damping installation region was divided into five parts with different damping thicknesses. The total weight of damping was set as an objective function and the A-weighted sound pressure level of the target cabin was set as a constraint condition. The best damping thickness was obtained through the optimization program, and the total damping weight was reduced by 60.4%. The results show that the damping noise reduction effect of unit weight is significantly improved through the optimization method. This research successfully solves the installation position and thickness selection problems in the acoustic design of damping control, providing a reliable analysis method and guidance for the design.

  15. Crash testing of spent-nuclear-fuel shipping systems

    International Nuclear Information System (INIS)

    Yoshimura, H.R.

    1978-01-01

    Full scale testing to date has verified that current analytical tools and the use of scale model testing are both accurate methods for predicting shipping cask response to severe accident conditions. The containers tested are capable of surviving severe transportation accidents

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

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1993-01-01

    The use of burnup credit in the criticality safety analysis of the GA-4 Cask increases the cask's capacity from three spent fuel assemblies to four, resulting in reduced public and occupational risk and reduced life cycle costs. GA's criticality calculations for burnup credit, including the associated uncertainties and analytical bias, establish the minimum burnup required as a function of initial enrichment to maintain K eff ≤ 0.95 under any conceivable condition. The minimum burnup requirement as a function of initial enrichment has been determined to be 15,000 MWd/MTU for 3.5 wt% U-235 fuel, 20,000 MWd/MTU for 4.0 wt% U-235 fuel and 25,000 MWd/MTU for 4.5 wt% U-235 fuel. The minimum burnup requirement as a function of enrichment is well below the typical burnup levels seen in the current and projected spent fuel inventory. (J.P.N.)

  17. 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

  18. A revision of the Cask Designers Guide for the '90s

    International Nuclear Information System (INIS)

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

    1992-01-01

    The report A Guide for the Design Fabrication, and Operation of Shipping Casks for Nuclear Applications, ORNL-NSIC-68, commonly called the Cask Designers Guide, is being revised at the request of the Transportation and Packaging Safety Division of the Department of Energy (DOE). The new document will be called the Packaging Handbook. The Cask Designers Guide was published in 1970 during the period when many radioactive materials packagings were being developed and many technical studies applicable to these packagings were being performed. Since that period, many improvements in packaging design have appeared, designers have improved their calculational techniques, and much effort has gone into applying Quality Assurance (QA) principles to cask development Materials, and their limitations, have surfaced as a very important consideration in the licensing process. While the Packaging Handbook considers all Type B packages, most of the authors' experience lies in the technical areas found in the licensing of spent nuclear fuel (SNF) packagings and this is reflected in the document

  19. Outer Dynamics of Ship Collisions

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup

    1996-01-01

    The purpose is to present analysis procedures for the motion of ships during ship-ship collisions and for ship collisions with offshore structures. The aim is to estimate that part of the lost kinetic energy which will have to be absorbed by rupture and plastic damage of the colliding structures....

  20. Outer Dynamics of Ship Collisions

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup

    1996-01-01

    The purpose of these notes is to present analysis procedures for the motion of ships during ship-ship collisions and for ship collisons with offshore structures. The aim is to estimate that part of the lost kinetic energy which will have to be absorbed by rupture and plastic damage of the colliding...

  1. Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

    International Nuclear Information System (INIS)

    1991-01-01

    The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173

  2. A Cost Estimation Analysis of U.S. Navy Ship Fuel-Savings Techniques and Technologies

    Science.gov (United States)

    2009-09-01

    Horngren , C. T., Datar, S . M., & Foster, G. (2006). Cost Accounting : A Managerial Emphasis. 12th ed. Saddle River, NJ: Pearson...COVERED Master’s Thesis 4. TITLE AND SUBTITLE A Cost Estimation Analysis of U.S. Navy Ship Fuel-Savings Techniques and Technologies 6. AUTHOR( S ...FY12 FY13 FY14 FY15 FY16 FY17 FY18 N P V   C u m   S a v i n g s   ( $ / y r / S D   s h i p s ) Time

  3. 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

  4. 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.)

  5. Use of remote systems and automation in radioactive waste shipping and receiving

    International Nuclear Information System (INIS)

    Berger, J.D.; Gneiting, B.C.; Sanders, T.L.

    1986-01-01

    A cask handling technology development program is being pursued at the Hanford Engineering Development Laboratory in Richland, Washington and at Sandia National Laboratories in Albuquerque, New Mexico. This is part of the United States Department of Energy's program to enhance nuclear waste transportation technology. The goal of the effort is two-fold; first, to develop concepts that will reduce occupational exposures at shipping cask handling facilities within As Low As Reasonably Achievable (ALARA) objectives as directed by US DOE order (1), and second, to identify any design requirements for future cask development activities that comply with 'design for automation' goals

  6. Review and Analysis of Selected Items Management (SIM) Inventory Program Aboard US Surface Ships

    National Research Council Canada - National Science Library

    Axinto, Mark I; Giles, Michael S

    2005-01-01

    .... While SIM items constitute only a small portion of the ship's entire COSAL spare parts inventory, they are critical to the ship's material condition and to its preventive and corrective maintenance programs...

  7. Failure assessment techniques to ensure shipping container integrity

    International Nuclear Information System (INIS)

    McConnell, P.

    1986-02-01

    This report discusses several methodologies which may be used to ensure the structural integrity of containment systems to be used for the transport and storage of high-level radioactive substances. For economic reasons, shipping containers constructed of ferritic materials are being considered for manufacture by vendors in the US and Europe. Ferritic show an inherent transition from a ductile, high energy failure mode to a brittle, low energy fracture mode with decreasing temperature. Therefore, formal consideration of means by which to avoid unstable brittle fracture is necessary prior to the licensing of ferritic casks. It is suggested that failure of a shipping container wall be defined as occurring when a flaw extends through the outer wall of the containment system. Crack initiation which may lead to unstable brittle crack growth should therefore be prevented. It is suggested that a fundamental linear elastic fracture mechanics (lefm) approach be adopted on a case-by-case basis, applied perhaps by means of appropriate modifications to ASMA Section III or Section XI. A lefm analysis requires information concerning service temperatures, loading rates, flaw sizes, and applied stresses. Tentative judgments regarding these parameters for typical shipping containers have been made

  8. 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

  9. Radiation shielding and criticality safety assessment for KN-12 spent nuclear fuel transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Kyung; Shin, Chang Ho; Kim, Gi Hwan [Hanyang Univ., Seoul (Korea, Republic of)

    2001-08-15

    Because SNFs involve TRU (Transuranium), fission products, and fissile materials, they are highly radioactive and also have a possibility to be critical. Therefore, radiation shielding and criticality safety for transport casks containing the SNFs should be guaranteed through reliable valuation procedure. IAEA safety standard series No ST-1 recommends regulation for safe transportation of the SNFs by transport casks, and United States is carrying out it according to the regulation guide, 10 CFR parts 71 and 72. Present research objective is to evaluate the KN-12 spent nuclear fuel transport cask that is designed for transportation of up to 12 assemblies and is standby status for being licensed in accordance with Korea Atomic Energy Act. Both radiation shielding and criticality analysis using the accurate Monte Carlo transport code, MCNP-4B are carried out for the KN-12 SNF cask as a benchmark calculation. Source terms for radiation shielding calculation are obtained using ORIGEN-S computer code. In this work, for normal transport conditions, the results from MCNP-4B shows the maximum dose rate of 0.557 mSv/hr at the side surface. And the maximum dose rate of 0.0871 mSv/hr was resulted at the 2 m distance from the cask. The level of calculated dose rate is 27.9% of the limit at the cask surface, 87.1% at 2 m from the cask surface for normal transport condition. For hypothetical accident conditions, the maximum rate of 2.5144 mSv/hr was resulted at the 1 m distance from the cask and this level is 25.1% of the limit for hypothetical accident conditions. In criticality calculations using MCNP-4B, the k{sub eff} values yielded for 5.0 w/o U-235 enriched fresh fuel are 0.92098 {+-} 0.00065. This result confirms subcritical condition of the KN-12 SNF cask and gives 96.95% of recommendations for criticality safety evaluation by US NRC these results will be useful as a basis for approval for the KN-12 SNF cask.

  10. 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

  11. Dynamic fracture toughness data for CASTOR registered casks

    International Nuclear Information System (INIS)

    Winkler, H.P.; Trubitz, P.; Pusch, G.; Warnke, E.P.; Beute, K.; Novotny, V.

    2004-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    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.

  14. 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

  15. 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

  16. An Analysis of Ship-Source Marine Pollution in Nigeria Seaports

    Directory of Open Access Journals (Sweden)

    D. E. Onwuegbuchunam

    2017-08-01

    Full Text Available Existing studies indicate that marine pollution control in the ports of developing economies is marred by a lack of administrative control and inadequate provision of waste reception facilities. In Nigeria ports, ship generated waste control services and provision of waste reception facilities are outsourced to private companies with no requirement for an activity audit. Apart from the port authority, other government agencies are also involved in pollution monitoring and control. Hence, functions are duplicated and effective regulation is arguably weakened by conflicts of interest. A scientific based integrated model is therefore proposed to address the managerial problem posed in the control of marine pollution in Nigerian ports. In this paper, we conduct a physico-chemical and microbiological analysis of samples of ships’ wastewater to determine the status of marine pollution in the port environment. The samples were collected from randomly selected ships at berths in seaport locations. The outputs from the analysis are then integrated as inputs into an administrative framework model. The integrated model developed is proposed as an alternative administrative tool for monitoring and controlling pollution in seaports. The policy implications of the developed model are discussed.

  17. A Method for Ship Collision Damage and Energy Absorption Analysis and its Validation

    DEFF Research Database (Denmark)

    Zhang, Shengming; Pedersen, Preben Terndrup

    2016-01-01

    -examine this method’s validity andaccuracy for ship collision damage analysis in shipdesign assessments by comprehensive validations withthe experimental results from the public domain. Twentyexperimental tests have been selected, analysed andcompared with the results calculated using the proposedmethod. It can......For design evaluation there is a need for a method whichis fast, practical and yet accurate enough to determine theabsorbed energy and collision damage extent in shipcollision analysis. The most well-known simplifiedempirical approach to collision analysis was madeprobably by Minorsky and its...... limitation is also wellrecognized.The authors have previously developedsimple expressions for the relation between the absorbedenergy and the damaged material volume which take intoaccount the structural arrangements, the materialproperties and the damage modes. The purpose of thepresent paper is to re...

  18. 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

  19. Fuel exchanging machine for a nuclear ship

    International Nuclear Information System (INIS)

    Hayashi, Tetsuji.

    1984-01-01

    Purpose: To prevent atmospheric contaminations upon fuel exchange thereby keep the environmental circumstance clean in the periphery of the nuclear ship. Constitution: A nuclear reactor container is disposed to the inside of a containing vessel in the ship body and a shutter is mounted to the upper opening of the ship body. Further, a landing container having a bottom opening equipped with shutter for alingning the upper opening equipped with shuuter of the ship is elevatably suspended to the trolley of a crane by way of a wire rope and a winch, and a fuel exchange cask is elevatably disposed to the inside of the landing container. Further, airs in the inside of the container is adapted to be discharged externally through a filter by means of a blower and the inside is kept at a negative pressure. Thus, since the containing vessel is covered with the landing container upon fuel exchanging operation, atmospheric contamination can be prevented sufficiently. (Sekiya, K.)

  20. PM10 source apportionment applying PMF and chemical tracer analysis to ship-borne measurements in the Western Mediterranean

    Science.gov (United States)

    Bove, M. C.; Brotto, P.; Calzolai, G.; Cassola, F.; Cavalli, F.; Fermo, P.; Hjorth, J.; Massabò, D.; Nava, S.; Piazzalunga, A.; Schembari, C.; Prati, P.

    2016-01-01

    A PM10 sampling campaign was carried out on board the cruise ship Costa Concordia during three weeks in summer 2011. The ship route was Civitavecchia-Savona-Barcelona-Palma de Mallorca-Malta (Valletta)-Palermo-Civitavecchia. The PM10 composition was measured and utilized to identify and characterize the main PM10 sources along the ship route through receptor modelling, making use of the Positive Matrix Factorization (PMF) algorithm. A particular attention was given to the emissions related to heavy fuel oil combustion by ships, which is known to be also an important source of secondary sulphate aerosol. Five aerosol sources were resolved by the PMF analysis. The primary contribution of ship emissions to PM10 turned out to be (12 ± 4)%, while secondary ammonium sulphate contributed by (35 ± 5)%. Approximately, 60% of the total sulphate was identified as secondary aerosol while about 20% was attributed to heavy oil combustion in ship engines. The measured concentrations of methanesulphonic acid (MSA) indicated a relevant contribution to the observed sulphate loading by biogenic sulphate, formed by the atmospheric oxidation of dimethyl sulphide (DMS) emitted by marine phytoplankton.

  1. Technical Analysis of Organic Rankine Cycle System Using Low-Temperature Source to Generate Electricity in Ship

    Directory of Open Access Journals (Sweden)

    Akram Faisal

    2017-01-01

    Full Text Available Nowadays, the shipping sector has growth rapidly as followed by the increasing of world population and the demands for public transportation via sea. This issue entails the large attention on emission, energy efficiency and fuel consumption on the ship. Waste Heat Recovery (WHR is one of the solution to overcome the mentioned issue and one of the WHR method is by installing Organic Rankine Cycle (ORC system in ship. ORC demonstrate to recover and exploit the low temperature waste heat rejected by the ship power generation plant. The main source of heat to be utilized is obtained from container ship (7900 kW BHP, DWT 10969 mt ship jacket water cooling system and use R-134a as a refrigerant. The main equipment consists of evaporator, condenser, pump and steam turbine to generate the electricity. The main objective is to quantifying the estimation of electrical power which can be generated at typical loads of the main engine. As the final result of analysis, the ORC system is able to generate the electricity power ranged from 77,5% - 100% of main engine load producing power averagely 57,69 kW.

  2. The analysis of thermal-hydraulic performances of nuclear ship reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Shinshichi; Hamada, Masao

    1975-01-01

    Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

  3. Analysis Of Inland Waterway Transport For Container Shipping: Cikarang To Port Of TanjungPriok

    Science.gov (United States)

    Achmadi, T.; Nur, H. I.; Rahmadhon, L. R.

    2018-03-01

    Industry's development which is in the center of Cikarang's industrial estate causes a considerable increase from 7% to 13% of container's flow from and to Port of Tanjung Priok per year. Therefore, those obstacles rise the number of traffic congestion and transport cost. This research aims to analyze the potential alternative of transportation in order to transport containers at the route of Tanjung Priok to Cikarang utilizing Inland Waterways Transport through Cikarang Bekasi Laut (CBL) river. This research will be conducted by comparing component of total logistic cost that emerging caused by container trucks and vessels. Self Propelled Container Barge (SPCB) is a pointed alternative transportation in which it is used to transport containers through the waterways. The result of analysis obtained that the capacity of Cikarang Bekasi Laut river is 18,558 roundtrip per year. Furthermore, the collaboration of 3 SPCB operations, as well payload 32 TEUS can decrease the amount of road traffic congestion/density of Cikarang-Port of Tanjung Priok as much as 18.6%. The cost of containers shipping per unit transported by truck is IDR 2.2 Million per TEUs, whereas containers shipping transported by Inland Waterways cost only IDR 1.8 Million per TEUs.

  4. Nonlinear analysis for a ship with a general roll-damping model

    Energy Technology Data Exchange (ETDEWEB)

    El-Bassiouny, A F [Mathematics Department, Faculty of Science, Benha University, Benha 13518 (Egypt)

    2007-05-15

    The qualitative behaviour of the response of a ship rolling in longitudinal waves whose amplitude or frequency (parameters) are slowly varied is presented. An analytical and numerical technique is used to predict the qualitative change taking place in the stable solutions of a ship model as one of the parameters is slowly changed. The analysis took into consideration linear, cubic and quantic terms in the polynomial expansion of the relative roll angle. The damping moment consists of the linear term associated with radiation and viscous damping and a cubic term due to frictional resistance and eddies behind bilge keels and hard bilge corners. Two methods (the averaging and the multiple time scales) are used to investigate a first-order approximate analytical solution. The modulation equations of the amplitudes and phases are obtained. These equations are used to determine steady state solutions. Numerical calculations are presented which illustrate the behaviour of the steady state response amplitude as a function of the detuning parameter. The stability of the proposed solution is determined applying Liapunov's first method. The effects of different parameters on the system behaviour are investigated numerically. The results obtained by the two methods are in excellent agreement.

  5. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G{sub eff}, is defined by: G{sub eff} - {Sigma}{sub M} (F{sub M} x G{sub M}) F{sub M}-fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials.

  6. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    International Nuclear Information System (INIS)

    1994-10-01

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G eff , is defined by: G eff - Σ M (F M x G M ) F M -fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials

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

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

    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

  8. 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.

  9. Fuel shipment experience, fuel movements from the BMI-1 transport cask

    International Nuclear Information System (INIS)

    Bauer, Thomas L.; Krause, Michael G.

    1986-01-01

    The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including several factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask

  10. Thick nickel plating of spent fuel transport and storage casks CASTOR and POLLUX

    International Nuclear Information System (INIS)

    Wilbuer, K.

    1991-01-01

    Spent fuel elements have to be safely handled in containers for transport and storage. These large casks (100-120 t) are made by various firms according to the specifications given by the nuclear plant operator. For shielding and protection of the hazardous material, the casks' inner surface is coated with a nickel plating about 3000 μm thick. The product and the production process are subject to very stringent requirements, due to the hazardous potential of the material to be shipped or stored. Therefore, both the extremely high quality standards to be met by the nickel plating and the dimensions and capability of the plating plant required for the process are problems that cannot be solved by a usual commercial plating plant. The new concept and process that had to be established are explained in the paper. (orig./MM) [de

  11. Storage/transport cask design and challenges

    International Nuclear Information System (INIS)

    Houston, J.V.; Viebrock, J.M.

    1989-01-01

    The concept of spent-fuel casks that could be used for both storage and for transport has been around for some years, but was only seriously evaluated when utilities started becoming concerned about adequate fuel storage. In the early 1980s, the U.S. Department of Energy proposed to solve the problem with their away-from-reactor storage facility concept. This was superceded by passage of the Nuclear Waste Policy Act of 1982, which directed the development of one or more waste repositories, the first of which was to be in operation by 1998. Delays in this program now indicate an opening data of 2003 or later. This, together with the lack of significant progress on a monitored retrievable storage facility, leaves the utility companies to solve their storage problems individually. One alternative is to use dual-purpose casks. The use of such a cask should eliminate the need to move the cask and fuel back into the spent-fuel pool for transfer to a transport cask. However, a dual-purpose cask must be licensed for use under both 10CFR71 and 10CFR72 of the U.S. Code of Federal Regulations. The purpose of this paper is to examine the differences between the requirements of 10CFR71 and 10CFR72, to note the changes over the past several years in the NRC's interpretation of 10CFR71 requirements, and to review the design modifications that the Nuclear Assurance Corporation (NAC) believes are required to make a licensed storage cask acceptable for transport under 10CFR71

  12. 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.

  13. 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

  14. Storage and transport casks combine to bring benefits

    International Nuclear Information System (INIS)

    Thorup, C.

    1988-01-01

    The Nuclear Assurance Corporation is currently preparing a safety report on its new spent fuel storage/transport casks. The report is due to be submitted to the NRC in 1989, together with an application for a licence. The aim of the combined casks is to simplify the process of dealing with spent fuel, whilst keeping costs down. The design of the casks is described, together with questions relating to the licensing of the casks. (author)

  15. CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2010-07-11

    Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and

  16. Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Highlights: • Hybrid heat pipe was presented as a passive cooling device for dry storage cask of SNF. • A method to utilize waste heat from spent fuel was suggested using hybrid heat pipe. • CFD analysis was performed to evaluate the thermal performance of hybrid heat pipe. • Hybrid heat pipe can increase safety margin and storage capacity of the dry storage cask. - Abstract: Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry

  17. 78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

    ...-2012-0052] RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment... document proposed to amend the NRC's spent fuel storage regulations by revising the Holtec International HI...

  18. Criticality studies for dry storage cask

    International Nuclear Information System (INIS)

    Krishnani, P.D.; Srinivasan, K.R.

    1993-01-01

    Spent nuclear fuel from Tarapur Atomic Power Station (TAPS) is stored in a storage pool located inside the reactor building. The capacity of this pool was initially to meet storage requirements of 528 bundles which was later augmented from time to time. Since the enhanced capacity was also getting exhausted, setting up of a storage pool away from reactor was envisaged. As an interim measure, the dry storage casks were designed to store the spent fuel already cooled for a few years in the storage pools. If water enters the cask, the cask interior may be covered with steam water or air-water mixture. This paper gives the results of criticality calculations for storage cask under various conditions of steam water mixture, using the computer code LWRBOX. In these calculations, it has been assumed that the cask contains the most reactive fuel assemblies of reload-1 at zero burnup. It also gives the comparison of some of the results with General Electric (GE) calculations. (author). 3 refs., 1 fig., 2 tabs

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Buckling of Ship Structures

    CERN Document Server

    Shama, Mohamed

    2013-01-01

    Buckling of Ship Structures presents a comprehensive analysis of the buckling problem of ship structural members. A full analysis of the various types of loadings and stresses imposed on ship plating and primary and secondary structural members is given. The main causes and consequences of the buckling mode of failure of ship structure and the methods commonly used to control buckling failure are clarified. This book contains the main equations required to determine the critical buckling stresses for both ship plating and the primary and secondary stiffening structural members. The critical buckling stresses are given for ship plating subjected to the induced various types of loadings and having the most common boundary conditions encountered in ship structures.  The text bridges the gap existing in most books covering the subject of buckling of ship structures in the classical analytical format, by putting the emphasis on the practical methods required to ensure safety against buckling of ship structur...