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

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)

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

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

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

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

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.)

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

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

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

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)

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

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

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

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

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.

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

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

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

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

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

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

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

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

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

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

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

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

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

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)

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

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

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

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.

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

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.

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)

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

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

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

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

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

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

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

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)

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

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%

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

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)

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

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

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)

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

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

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.)

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

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

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

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.)

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

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

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%

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

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

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

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

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.

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

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

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

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

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

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

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

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

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.)

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)

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)

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

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

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

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

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

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

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.

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

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

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

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

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

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

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

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)

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

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.

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

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.

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

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

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)

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

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.)

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

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

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

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

An assessment of the transportation costs of shipping non-fuel assembly hardware to FWMS facilities

International Nuclear Information System (INIS)

Shappert, L.B.; Joy, D.S.; Johnson, P.E.; Danese, F.L.; Best, R.E.

1991-01-01

This study examines the cost of using Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) Initiative I casks for transporting 62,700 MTU of spent fuel plus associated non-fuel assembly hardware (NFAH) between reactor sites and either a monitored retrievable storage (MRS) or a repository facility. The study further considers the benefits of increasing the cell size of the Initiative I BWR cask baskets to accommodate the fuel plus channels (which also would decrease the capacity of the cask to carry BWR fuel without channels) and the use of a commercial, non-spent-fuel cask to carry compacted NFAH that could not be shipped integrally. Costs that are developed involve transportation charges, capital costs for casks, and canning costs of NFAH that have been separated from the fuel. The results indicate that significant cost savings are possible if NFAH is accepted into the Federal Waste Management System (FWMS) that is either integral with the spent fuel, or consolidated if it has been separated. Shipment of unconsolidated NFAH is very expensive. Transportation costs for shipping to a western repository are approximately 50 to 75% higher than the costs for shipping to an eastern MRS

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

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

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

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

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.

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)

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

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

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

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

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.

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.

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

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

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

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

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

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.

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)

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

Crash testing of nuclear fuel shipping containers

International Nuclear Information System (INIS)

Jefferson, R.M.; Yoshimura, H.R.

1977-08-01

In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. The program encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

Crash testing of nuclear fuel shipping containers

International Nuclear Information System (INIS)

Jefferson, R.M.; Yoshimura, H.R.

1977-12-01

In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. This program, which began in 1975, encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography, the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

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

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

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

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)

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

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

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

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

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

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

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.

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

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

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.)

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)

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

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

Full scale simulations of accidents on spent-nuclear-fuel shipping systems

International Nuclear Information System (INIS)

Yoshimura, H.R.

1978-01-01

In 1977 and 1978, five first-of-a-kind full scale tests of spent-nuclear-fuel shipping systems were conducted at Sandia Laboratories. The objectives of this broad test program were (1) to assess and demonstrate the validity of current analytical and scale modeling techniques for predicting damage in accident conditions by comparing predicted results with actual test results, and (2) to gain quantitative knowledge of extreme accident environments by assessing the response of full scale hardware under actual test conditions. The tests were not intended to validate the present regulatory standards. The spent fuel cask tests fell into the following configurations: crashes of a truck-transport system into a massive concrete barrier (100 and 130 km/h); a grade crossing impact test (130 km/h) involving a locomotive and a stalled tractor-trailer; and a railcar shipping system impact into a massive concrete barrier (130 km/h) followed by fire. In addition to collecting much data on the response of cask transport systems, the program has demonstrated thus far that current analytical and scale modeling techniques are valid approaches for predicting vehicular and cask damage in accident environments. The tests have also shown that the spent casks tested are extremely rugged devices capable of retaining their radioactive contents in very severe accidents

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

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

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

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)

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.

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)

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

LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE

Energy Technology Data Exchange (ETDEWEB)

Dunn, K. [Savannah River National Laboratory; Bellamy, S. [Savannah River National Laboratory; Daugherty, W. [Savannah River National Laboratory; Sindelar, R. [Savannah River National Laboratory; Skidmore, E. [Savannah River National Laboratory

2013-08-18

Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintain integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.

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

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

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

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

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

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)

Environmental impact of accidents involving radioactive material shipping systems

International Nuclear Information System (INIS)

Yoshimura, H.R.; Pope, R.B.; Huerta, M.; Nilson, R.H.

1978-01-01

Four full-scale spent fuel cask crash tests have been performed, including two head-on truck-barrier impacts (100 and 135 km/h), one railcar-barrier impact (130 km/h), and one locomotive grade crossing impact (130 km/h). Releases to the environment were limited to seepage of about 100 cc of cavity liquid from the cask head in the 135 km/h truck impact test and a slight head seal air leak in the 130 km/h locomotive grade crossing test. These releases were well within the limits specified by the NRC regulations, would have been easily cleaned up, and would have caused little effect on the environment and virtually no risk to the public. To further evaluate cask capability, the crashed spent-fuel rail cask system was fire tested. The cask withstood 90 minutes of a fully engulfing hydrocarbon pool fire while maintaining its structural integrity. At approximately 100 minutes into the fire test, the outer shell of the cask cracked resulting in the partial loss of lead radiation shielding. The failure of the shell was attributed to poor quality control during the original fabrication of the cask in the early 1960's. Present regulatory standards would prevent such occurrences in casks built and licensed today. In addition, the test was much more severe than the qualification criteria specified by present licensing requirements. 4 tables, 13 figures

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)

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

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

Dry cask storage: a Vepco/DOE/EPRI cooperative demonstration program

International Nuclear Information System (INIS)

Smith, M.L.

1984-01-01

In response to a Department of Energy (DOE) Solicitation for Cooperative Agreement Proposal, Virginia Electric and Power Company (Vepco) proposed to participate in a spent fuel storage demonstration program utilizing the dry cask storage technology. This proposed program includes dry cask storage at Vepco's Surry Nuclear Power Station and research and development activities at a DOE site in support of the licensed program at Surry. Phase I of Vepco's two-phase program involves a demonstration of the licensed dry cask storage of spent fuel in an inert atmosphere at the Surry Power Station site. Phase II of Vepco's proposed program will involve the demonstration of storing unconsolidated and consolidated spent fuel in dry casks filled only with air. This phase of the program will involve DOE site testing similar to Phase I and is expected to require an additional (fourth) cask to demonstrate storage of unconsolidated spent fuel in air-filled casks

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

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

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

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

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

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

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

Operations experience with the NAC-1 legal weight truck cask

International Nuclear Information System (INIS)

Viebrock, J.M.; Hoffman, C.C.

1978-01-01

The first three years of operation of Nuclear Assurance Corporation's (NAC) four (4) NAC-1 Casks have demonstrated that shipments of spent fuel, fuel rods and other highly irradiated reactor components can be moved routinely by legal weight truck transport. Shipments of these materials have involved some 800,000 miles of highway travel and cask handling at some fifteen different nuclear facilities. This paper presents details on NAC's operations experience with these casks including cask description, cask handling (loading and unloading), pre-shipment testing, facility turnaround and transit times, operator exposure, transport vehicles and shipper/carrier/cask owner responsibilities, actual experience with regard to facility interfacing requirements and operational procedures. Cask and equipment utilization is discussed together with the methods used to control operation costs and to improve the economics of truck transport

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)

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.)

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

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

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

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)

Protecting against failure by brittle fracture in ferritic steel shipping containers

International Nuclear Information System (INIS)

Schwartz, M.W.; Langland, R.T.

1983-01-01

The possible use of ferritic steels for the containment structure of shipping casks has motivated the development of criteria for assuring the integrity of these casks under both normal and hypothetical accident conditions specified in Part 71 of the Code of Federal Regulations. The US Nuclear Regulatory Commission Regulation Guide 7.6 provides design criteria for preventing ductile failure steel shipping containers. The research described in this paper deals with criteria for preventing brittle fracture of ferritic steel shipping containers. Initially guidelines were developed for ferritic steel up to four inches thick (I). This was followed by an investigation of various criteria that might be used for monolithic thick walled casks greater than four inches thick (2). Three categories of safety are identified in the design of shipping containers. Category I, the highest level of safety, is appropriate for containment systems for spent nuclear fuel and high level waste transport packaging. In Category I, containers are designed to the highest level of safety and brittle fracture is essentially not possible. Categories II and III represent levels of safety commensurate with the consequences of release of lower levels of radioactivity. In these latter categories, consideration of factors contributing to brittle fracture, good engineering practice, and careful selection of material make brittle fracture unlikely under environmental conditions encountered during shipping. This paper will deal primarily with Category I containers. The guidelines for Category II and III containers are fully described elsewhere. 5 references, 10 figures, 3 tables

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

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

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)

Description of a materials/coolant laboratory for support of the Breeder Reactor Technology Shipping Program

International Nuclear Information System (INIS)

Rack, H.J.; Rohde, R.W.

1979-04-01

A description of a facility devoted to evaluating the environmental compatibility and mechanical response of materials suitable for a breeder reactor spent-fuel shipping cask is given. The facility presently consists of a closed-loop servo-controlled hydraulic, horizontal test system coupled to an environmental chamber, generalized mechanical test equipment and high-rate mechanical behavior apparatus. Future plans include the procurement of real-time computer control equipment which will be used to assess the effects of complex load-time histories on spent-fuel shipping cask materials

Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study

International Nuclear Information System (INIS)

Tyacke, M.

1993-08-01

This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood

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.

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

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

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

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.

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

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.

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

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

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

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

International Nuclear Information System (INIS)

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

1990-12-01

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

CASK inhibits ECV304 cell growth and interacts with Id1

International Nuclear Information System (INIS)

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

2005-01-01

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

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

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

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

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

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...

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

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

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

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

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

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

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

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

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

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

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

Loading 076 assemblies in two IV-04 transport casks for transport to the U.S. Savannah River Site (SC); Trasferimento di 72 elementi irraggiati MTR dalla piscina dell`impianto EUREX a due contenitori IU-04 per il trasporto al Savannah River Site-Department of Energy (USA)

Energy Technology Data Exchange (ETDEWEB)

Gili, Michele [ENEA, Centro Ricerche Saluggia, Vercelli (Italy). Dipt. Energia

1997-09-01

The National Agency for New Technologies and the Environments has signed with the US Department of Energy a contract for the transfer of 150 irradiated MTR fuel assemblies stored in the EUREX plant pool at The National Agency for New Technologies and the Environments Research Centre of Saluggia. The first scheduled transport has been made in february 1997 and has involved the successful loading of 76 assemblies in two IU-04 (Pegase) transport casks. The loaded casks have been shipped to the U.S. Savannah River Site (SC).

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

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...

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...

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

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

Probabilistic risk assessment on maritime spent nuclear fuel transportation (Part II: Ship collision probability)

International Nuclear Information System (INIS)

Christian, Robby; Kang, Hyun Gook

2017-01-01

This paper proposes a methodology to assess and reduce risks of maritime spent nuclear fuel transportation with a probabilistic approach. Event trees detailing the progression of collisions leading to transport casks’ damage were constructed. Parallel and crossing collision probabilities were formulated based on the Poisson distribution. Automatic Identification System (AIS) data were processed with the Hough Transform algorithm to estimate possible intersections between the shipment route and the marine traffic. Monte Carlo simulations were done to compute collision probabilities and impact energies at each intersection. Possible safety improvement measures through a proper selection of operational transport parameters were investigated. These parameters include shipment routes, ship's cruise velocity, number of transport casks carried in a shipment, the casks’ stowage configuration and loading order on board the ship. A shipment case study is presented. Waters with high collision probabilities were identified. Effective range of cruising velocity to reduce collision risks were discovered. The number of casks in a shipment and their stowage method which gave low cask damage frequencies were obtained. The proposed methodology was successful in quantifying ship collision and cask damage frequency. It was effective in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • Proposes a probabilistic framework on the safety of spent nuclear fuel transportation by sea. • Developed a marine traffic simulation model using Generalized Hough Transform (GHT) algorithm. • A transportation case study on South Korean waters is presented. • Single-vessel risk reduction method is outlined by optimizing transport parameters.

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

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.)

Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

Energy Technology Data Exchange (ETDEWEB)

Tyacke, M.

1993-08-01

This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Ventura, Rodrigo, E-mail: rodrigo.ventura@isr.ist.utl.pt [Laboratório de Robótica e Sistemas em Engenharia e Ciência – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Filip, Iulian, E-mail: ifilip@gmail.com [Faculty of Mechanical Engineering – Technical University Gheorghe Asachi of Iasi, 61 Dimitrie Mangeron Bldv., Iasi 700050 (Romania); Vale, Alberto, E-mail: avale@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

2013-10-15

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Basic planning of a newly built exclusive ship for spent fuel transport

International Nuclear Information System (INIS)

Obara, I.; Sasao, T.; Akiyama, H.; Kybota, T.

1998-01-01

A commercial reprocessing plant is under construction at the Fuel Cycle Facilities in Rokkasho-mura, Aomori Prefecture. To prepare for the transport of spent nuclear fuels (SF) from all Japanese nuclear power stations to this reprocessing plant, the need for an exclusive transport ship was recognized. Nuclear Fuel Transport Co. Ltd. (NFT), in cooperation with electric power utilities planned the construction of such a ship over a period of several years. During this period NFT developed new types of cask to transport high burn-up spent fuels to the reprocessing plant. Six kinds of casks were developed and 40 units are now under fabrication. The ship was designed to carry a maximum of 20 units. Based on the Irradiated Nuclear Fuel (INF) Code adopted by the International Maritime Organization (IMO), the Japanese Ministry of Transport (MoT) issued new domestic regulations in September, 1995 which covered design criteria for ships carrying Irradiated Nuclear Fuels. The new SF transport ship is the first one to which this new regulation was applied. Although the ship will only ply the coastal routes of Japan, she has been designed to conform with all the international requirements for the Class-3 of the INF Code. In May 1995, Nuclear Fuel Shipping Co. Ltd (NFS), a wholly-owned subsidiary of NFT, concluded a contract with Mitsui Engineering and Shipbuilding Co., Ltd. for the construction of the exclusive transport ship. The keel was laid in November 1995. The ship was launched in april 1996 and named 'Rokuei-Maru'. At the end of September, she was completed and delivered to the ship owner, NFS. (authors)

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...

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

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)

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

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.

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

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

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

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

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

Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

International Nuclear Information System (INIS)

Mobasheran, A.S.; Boshoven, J.; Lake, B.

1995-01-01

The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit

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)

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

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

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

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

Predicted occurrence rate of severe transportation accidents involving large casks

International Nuclear Information System (INIS)

Dennis, A.W.

1978-01-01

A summary of the results of an investigation of the severities of highway and railroad accidents as they relate to the shipment of large radioactive materials casks is discussed. The accident environments considered are fire, impact, crash, immersion, and puncture. For each of these environments, the accident severities and their predicted frequencies of occurrence are presented. These accident environments are presented in tabular and graphic form to allow the reader to evaluate the probabilities of occurrence of the accident parameter severities he selects

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''

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

Confirmation tests of PWR surveillance capsule shipping container

International Nuclear Information System (INIS)

Tomita, N.; Ue, K.; Ohashi, M.; Asada, K.; Yoneda, Y.

1980-01-01

Mitsubishi Heavy Industries, Ltd. carried out the confirmation tests to confirm the reliability of the PWR surveillance capsule shipping container and to collect cask design data using a 10-ton weight full scale model at Kobe Shipyard and Engine Works. This report presents the outline of these tests. The B Type container was a cylinder 3289 mm long, 1080 mm in diameter and designed in accordance with the new modified Japanese regulations similar to IAEA regulation. These tests consist of four 9 m drop tests, two 1 m puncture tests, a fire test and an immersion test. In conclusion, safetyness of this container has been proved and various technical data for cask design were also collected through these tests. (author)

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...

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

Thermal Shielding of the Shock Absorber to a Seal of a Hot-cell Cask

Energy Technology Data Exchange (ETDEWEB)

Bang, K. S.; Lee, J. C.; Kim, K. Y.; Seo, C. S.; Seo, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-05-15

In order to safely transport the radioactive waste arising from the hot test of ACP(Advanced Spent Fuel Conditioning Process) a shipping package is required. Therefore KAERI is developing a shipping package to transport the radioactive waste arising in the ACPF during a hot test. Regulatory requirements for a Type B package are specified in the Korea MOST Act 2008-69, IAEA Safety Standard Series No. TS-R-1, and US 10 CFR Part. These regulatory guidelines classify the hot cell cask as a Type B package, and state that the Type B package for transporting radioactive materials should be able to withstand a test sequence consisting of a 9 m drop onto an unyielding surface, a 1 m drop onto a puncture bar, and a 30 minute fully engulfing fire. Greiner et al. performed a research on the thermal protection provided by shock absorbers by using CAFE computer code. This paper discusses the experimental approach used to simulate the response of the hot cell cask to fire in a furnace with chamber dimensions of 300 cm(W) x 400 cm(L) x 200 cm(H) by using a 1/2 scale model which was damaged by both a 9 m drop test and a 1 m puncture test

Impact of an exploding LPG rail tank car onto a CASTOR spent fuel cask

Energy Technology Data Exchange (ETDEWEB)

Droste, B.; Probst, U.; Heller, W

1999-07-01

On 27 April 1999 a fire test was performed with a 45 m{sup 3} rail tank car partially filled with 10 m3 pressurised liquid propane. A CASTOR THTR/AVR spent fuel transport cask was positioned beside the propane tank as to suffer maximum damage from any explosion. About 17 min after fire ignition the propane tank ruptured. This resulted in a BLEVE with an expanding fireball, heat radiation, explosion overpressure, and tank fragments projected towards the cask. This imposed severe mechanical and thermal impacts directly onto the CASTOR cask, moving it 17 m from its original position. This involved rotation of the cask with the lid end travelling 10 m before it crashed into the ground. Post-test investigations of the CASTOR cask demonstrated that no loss of leaktightness or containment and shielding integrity occurred. (author)

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.

Interfacing the existing cask fleet with the MRS

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Impacts of SNF burnup credit on the shipment capability of the GA-4 cask

International Nuclear Information System (INIS)

Mobasheran, A.S.; Lake, W.; Richardson, J.

1996-01-01

Scoping analyses were performed to determine the impacts of two different levels of burnup credit and two different spent fuel pickup rates on the shipment capability and the minimum fleet size of the GA-4 cask. The analyses involved developing loading curves for the GA-4 cask based on the actinide-only and principal-isotope burnup credit considerations. The analyses also involved examination of the spent nuclear fuel assembly population at nine reactor sites and categorization of the assemblies in accordance with the loading restrictions imposed. The results revealed that for the nine sites considered, depending on the level of burnup credit and the pickup rate assumed, the total savings in shipment and cask fleet costs (1994 dollars) can range from $55 million to $74 million

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.)

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

Structural dimensioning of dual purpose cask prototype

International Nuclear Information System (INIS)

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

2005-01-01

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

Criteria for cesium capsules to be shipped as special form radioactive material

International Nuclear Information System (INIS)

Lundeen, J.E.

1994-01-01

The purpose of this report is to compile all the documentation which defines the criteria for Waste Encapsulation and Storage Facility (WESF) cesium capsules at the IOTECH facility and Applied Radiant Energy Corporation (ARECO) to be shipped as special form radioactive material in the Beneficial Uses Shipping System (BUSS) Cask. The capsules were originally approved as special form in 1975, but in 1988 the integrity of the capsules came into question. WHC developed the Pre-shipment Acceptance Test Criteria for capsules to meet in order to be shipped as special form material. The Department of Energy approved the criteria and directed WHC to ship the capsules at IOTECH and ARECO meeting this criteria to WHC as special form material

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

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

Seismic stability of unanchored spent nuclear fuel storage casks

International Nuclear Information System (INIS)

Ofoegbu, G. I.; Gute, G. D.; Chowdhury, A. H.

2003-01-01

Dynamic soil-structure interaction analyses were performed to examine the effects of a potential earthquake on the stability of unanchored cylindrical spent nuclear fuel casks for an above-ground storage installation. The casks would be placed on a cluster of reinforced concrete pads founded on a deep sequence of clays and silts underlain by sandstones. The analyses focused on evaluating the geometric stability of the casks during an earthquake with respect to a design concept that a cask should not tip over, slide off the storage pad, or collide with another cask. The analyses were performed using LS-DYNA with a three-dimensional explicit finite element model representing the site soil and a fully loaded storage pad. Three statistically independent acceleration time histories were applied simultaneously at the base of the model to generate a free-field ground motion time history representing the design-basis earthquake. Sensitivity studies were performed to examine the effects of the interface conditions between the storage pad and the surrounding soil, and between the base of the storage casks and the top surface of the pad. The results indicate that ground motion from the design-basis earthquake would not cause any cask to tip over, slide off the pad, or collide with another cask. The contact conditions at the cask-to-pad and pad-to-soil interfaces have a strong effect on potential cask motions during an earthquake. If the cask-base friction coefficient is small, the casks may slide, but would not experience any significant rocking. If the cask-base friction is large enough to permit a significant transfer of earthquake lateral motions across the cask-to-pad interface, a design with bonded pad-to-soil interfaces would produce larger cask motions than a design with frictional pad-to-soil interfaces. Furthermore, a cask strage design in which the cask motions are essentially isolated from the motions of the pad-soil system, which can be accomplished if the cask

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

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

International Nuclear Information System (INIS)

Hopf, J.E.; Lloyd, T.

2004-01-01

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

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

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

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)

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

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

Science.gov (United States)

2013-12-06

... Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY: Nuclear Regulatory Commission. ACTION... storage regulations by revising the Holtec International HI- STORM 100 Cask System listing within the...C) No. 1014. Amendment No. 9 broadens the subgrade requirements for the HI-STORM 100U part of the HI...

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.)

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

International Nuclear Information System (INIS)

Shimizu, Masashi; Hayashi, Makoto; Kashiwakura, Jun

2003-01-01

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

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

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

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)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

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

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

Truck and rail charges for shipping spent fuel and nuclear waste

International Nuclear Information System (INIS)

McNair, G.W.; Cole, B.M.; Cross, R.E.; Votaw, E.F.

1986-06-01

The Pacific Northwest Laboratory developed techniques for calculating estimates of nuclear-waste shipping costs and compiled a listing of representative data that facilitate incorporation of reference shipping costs into varius logistics analyses. The formulas that were developed can be used to estimate costs that will be incurred for shipping spent fuel or nuclear waste by either legal-weight truck or general-freight rail. The basic data for this study were obtained from tariffs of a truck carrier licensed to serve the 48 contiguous states and from various rail freight tariff guides. Also, current transportation regulations as issued by the US Department of Transportation and the Nuclear Regulatory Commission were investigated. The costs that will be incurred for shipping spent fuel and/or nuclear waste, as addressed by the tariff guides, are based on a complex set of conditions involving the shipment origin, route, destination, weight, size, and volume and the frequency of shipments, existing competition, and the length of contracts. While the complexity of these conditions is an important factor in arriving at a ''correct'' cost, deregulation of the transportation industry means that costs are much more subject to negotiation and, thus, the actual fee that will be charged will not be determined until a shipping contract is actually signed. This study is designed to provide the baseline data necessary for making comparisons of the estimated costs of shipping spent fuel and/or nuclear wastes by truck and rail transportation modes. The scope of the work presented in this document is limited to the costs incurred for shipping, and does not include packaging, cask purchase/lease costs, or local fees placed on shipments of radioactive materials

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)

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.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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)

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

Development of metal cask for nuclear spent fuel

International Nuclear Information System (INIS)

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

2001-01-01

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

Operational and safety aspects of vitrified waste casks

International Nuclear Information System (INIS)

Kirchner, B.

1993-01-01

For the time being two technical solutions have been developed for the interim storage: 1) one is based on forced air cooled pits set out in a concrete structure, as presently provided close to the Vitrification Facilities on reprocessing sites; 2) the other one is based on transportable storage casks standing vertically onto a storage pad, following principles similar to those already experienced with spent fuel storage casks. Considering these two solutions for interim storage, TRANSNUCLEAIRE has developed two main types of transportable casks for vitrified HAW; one is a routine transport cask; the other one is a transportable storage cask. Both are covered by the generic name TN28V and have already been described in previous papers. This paper deals with the safety and operation aspects of the casks under both transport and storage conditions. (J.P.N.)

Development of integrated cask body and base plate

International Nuclear Information System (INIS)

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

2015-01-01

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

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)

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

Transportation cask decontamination and maintenance at the potential Yucca Mountain repository

International Nuclear Information System (INIS)

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

1992-04-01

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

Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

International Nuclear Information System (INIS)

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

1992-01-01

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

A Cask Processing Enclosure for the TRU Waste Processing Center - 13408

Energy Technology Data Exchange (ETDEWEB)

Newman, John T.; Mendez, Nicholas [IP Systems, Inc., 2685 Industrial Lane, Broomfield, Colorado 80020 (United States)

2013-07-01

This paper will discuss the key elements considered in the design, construction, and use of an enclosure system built for the TRU Waste Processing Center (TWPC). The TWPC system is used for the repackaging and volume reduction of items contaminated with radioactive material, hazardous waste and mixed waste. The modular structural steel frame and stainless steel skin was designed for rapid field erection by the use of interchangeable self-framing panel sections to allow assembly of a sectioned containment building and for ease of field mobility. The structure was installed on a concrete floor inside of an outer containment building. The major sections included an Outer Cask Airlock, Inner Cask Airlock, Cask Process Area, and Personnel Airlocks. Casks in overpacks containing transuranic waste are brought in via an inter-site transporter. The overpack lid is removed and the cask/overpack is transferred into the Outer Cask Airlock. A contamination cover is installed on the overpack body and the Outer Cask Airlock is closed. The cask/overpack is transferred into the Inner Cask Airlock on a cask bogie and the Inner Cask Airlock is closed. The cask lid is removed and the cask is transferred into the Cask Process Area where it is placed on a cask tilting station. Once the Cask Processing Area is closed, the cask tilt station is activated and wastes are removed, size reduced, then sorted and re-packaged into drums and standard waste boxes through bag ports. The modular system was designed and built as a 'Fast Track' project at IP Systems in Broomfield Colorado and then installed and is currently in use at the DOE TWPC located near Oak Ridge, Tennessee. (authors)

Probabilistic Risk Assessment of Cask Drop Accident during On-site Spent Nuclear Fuel Transportation

Energy Technology Data Exchange (ETDEWEB)

Ham, Jae Hyun; Christian, Robby; Momani, Belal Al; Kang, Hyun Gook [KAERI, Daejeon (Korea, Republic of)

2016-05-15

There are two ways to transfer the SNF from a site to other site, one is land transportation and the other is maritime transportation. Maritime transportation might be used because this way uses more safe route which is far from populated area. The whole transportation process can be divided in two parts: transferring the SNF between SNP and wharf in-Nuclear Power Plant (NPP) site by truck, and transferring the SNF from the wharf to the other wharf by ship. In this research, on-site SNF transportation between SNP and wharf was considered. Two kinds of single accident can occur during this type of SNF transportation, impact and fire, caused by internal events and external events. In this research, PRA of cask drop accident during onsite SNF transportation was done, risk to a person (mSv/person) from a case with specific conditions was calculated. In every 11 FEM simulation drop cases, FDR is 1 even the fuel assemblies are located inside of the cask. It is a quite larger value for all cases than the results with similar drop condition from the reports which covers the PRA on cask storage system. Because different from previous reports, subsequent impact was considered. Like in figure 8, accelerations which are used to calculate the FDR has extremely higher values in subsequent impact than the first impact for all SNF assemblies.

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

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

Rail tiedown tests with heavy casks for radioactive shipments

International Nuclear Information System (INIS)

Petry, S.F.

1980-08-01

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

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)

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.

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

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

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

BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

International Nuclear Information System (INIS)

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

2006-01-01

The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN TM 24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN TM 9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN TM 9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN TM 24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN TM 9/4 round trips are performed, and one TN TM 24BH is loaded. 5 additional TN TM 24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN TM 24BH high capacity dual purpose cask and the TN TM 9/4 transport cask characteristics and capabilities will then be detailed. (authors)

Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport

International Nuclear Information System (INIS)

Boshoven, J.K.

1992-01-01

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. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. 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 initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the ''fresh fuel'' assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions

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

Standard casks for the transport of LWR spent fuel. Storage/transport casks for long cooled spent fuel

International Nuclear Information System (INIS)

Blum, P.; Sert, G.; Gagnon, R.

1983-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks are presently used for European and intercontinental transports and manufactured under TRANSNUCLEAIRE supervision in different countries. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardization faciliting fabrication, operation and spare part supply. Recently, TRANSNUCLEAIRE also developed a new generation of casks for the dry storage and occasional transport of LWR spent fuel which has been cooled for 5 years or 7 years in case of consolidated fuel rods. These casks have an optimum payload which takes into account the shielding requirements and the weight limitations at most sites. This paper deals more particularly with the TN 24 model which exists in 4 versions among which one for 24 PWR 900 fuel assemblies and another one for the consolidated fuel rods from 48 of same fuel assemblies

Recommendations for cask features for robotic handling from the Advanced Handling Technology Project

International Nuclear Information System (INIS)

Drotning, W.

1991-02-01

This report describes the current status and recent progress in the Advanced Handling Technology Project (AHTP) initiated to explore the use of advanced robotic systems and handling technologies to perform automated cask handling operations at radioactive waste handling facilities, and to provide guidance to cask designers on the impact of robotic handling on cask design. Current AHTP tasks have developed system mock-ups to investigate robotic manipulation of impact limiters and cask tiedowns. In addition, cask uprighting and transport, using computer control of a bridge crane and robot, were performed to demonstrate the high speed cask transport operation possible under computer control. All of the current AHTP tasks involving manipulation of impact limiters and tiedowns require robotic operations using a torque wrench. To perform these operations, a pneumatic torque wrench and control system were integrated into the tool suite and control architecture of the gantry robot. The use of captured fasteners is briefly discussed as an area where alternative cask design preferences have resulted from the influence of guidance for robotic handling vs traditional operations experience. Specific robotic handling experiences with these system mock-ups highlight a number of continually recurring design principles: (1) robotic handling feasibility is improved by mechanical designs which emphasize operation with limited dexterity in constrained workspaces; (2) clearances, tolerances, and chamfers must allow for operations under actual conditions with consideration for misalignment and imprecise fixturing; (3) successful robotic handling is enhanced by including design detail in representations for model-based control; (4) robotic handling and overall quality assurance are improved by designs which eliminate the use of loose, disassembled parts. 8 refs., 15 figs

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

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

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.

Castor-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed vacuum, nitrogen, and helium backfill environments with the cask in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium environments to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen environments with the cask in a vertical orientation and with helium with the cask in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen test runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design goal of less than 200 mrem/hr. Cask surface dose rates of <75 mrem/hr can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

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.

Final Technical Report: Imaging a Dry Storage Cask with Cosmic Ray Muons

Energy Technology Data Exchange (ETDEWEB)

Yang, Haori; Hayward, Jason; Can, Liao; Liu, Zhengzhi

2018-03-31

The goal of this project is to build a scaled prototype system for monitoring used nuclear fuel (UNF) dry storage casks (DSCs) through cosmic ray muon imaging. Such a system will have the capability of verifying the content inside a DSC without opening it. Because of the growth of the nuclear power industry in the U.S. and the policy decision to ban reprocessing of commercial UNF, the used fuel inventory at commercial reactor sites has been increasing. Currently, UNF needs to be moved to independent spent fuel storage installations (ISFSIs), as its inventory approaches the limit on capacity of on-site wet storage. Thereafter, the fuel will be placed in shipping containers to be transferred to a final disposal site. The ISFSIs were initially licensed as temporary facilities for ~20-yr periods. Given the cancellation of the Yucca mountain project and no clear path forward, extended dry-cask storage (~100 yr.) at ISFSIs is very likely. From the point of view of nuclear material protection, accountability and control technologies (MPACT) campaign, it is important to ensure that special nuclear material (SNM) in UNF is not stolen or diverted from civilian facilities for other use during the extended storage.

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.)

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...

FACSIM/MRS [Monitored Retrievable Storage]-2: Storage and shipping model documentation and user's guide

International Nuclear Information System (INIS)

Huber, H.D.; Chockie, A.D.; Hostick, C.J.; Otis, P.T.; Sovers, R.A.

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 FACSIM/MRS-2, which is also referred to as the back-end model. The FACSIM/MRS-2 model simulates the MRS storage and shipping operations, which include handling canistered spent fuel and secondary waste in the shielded canyon cells, in onsite yard storage, and in repository shipping cask loading areas

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

Impact of more conservative cask designs of the CRWMS transportation system

International Nuclear Information System (INIS)

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

1993-01-01

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

Conceptual Design Report Cask Loadout Sys and Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform at 105 K West

Energy Technology Data Exchange (ETDEWEB)

LANGEVIN, A.S.

1999-07-12

This conceptual design report documents the redesign of the IPSS and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5/A.6, Canister Transfer Facility Modifications. Project A.5/A.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The function of the IPSS is to suspend, guide, and position the immersion pail. The immersion pail protects the cask-MCO from contamination by basin water and acts as a lifting device for the cask-MCO. The OIP provides operator access to the south loadout pit. Previous analyses studied the effects of a cask-MCO drop on the south loadout pit concrete structure and on the IPSS. The most recent analysis considered the resulting loads at the pit slab/wall joint (Kanjilal, 1999). This area had not been modeled previously, and the analysis results indicate that the demand capacity exceeds the allowable at the slab/wall joint. The energy induced on the south loadout pit must be limited such that the safety class function of the basin is maintained. The solution presented in this CDR redesigns the IPSS and the OIP to include impact-absorbing features that will reduce the induced energy. The impact absorbing features of the new design include: Impact-absorbing material at the IPSS base and at the upper portion of the IPSS legs. A sleeve which provides a hydraulic means of absorbing energy. Designing the OIP to act as an impact absorber. The existing IPSS structure in 105 KW will be removed. This conceptual design considers only loads resulting from drops directly over the IPSS and south loadout pit area. Drops in other areas of the basin are not considered, and will be covered as part of a future revision to this CDR.

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

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

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.)

Shielding designs and tests of a new exclusive ship for transporting spent nuclear fuels

International Nuclear Information System (INIS)

Ono, M.; Ito, D.; Kitano, T.; Ueki, K.; Akiyama, H.; Obara, I.; Sanui, T.

2000-01-01

The Rokuei-Maru, a ship built specially for the transport of spent nuclear fuels in casks, was launched April in 1996. She is the first ship to comply with special Japanese regulations, KAISA 520, based on the INF code. DOT3.5 and MCNP-4A were used for the evaluation of dose equivalent rates of her shielding structures. On-board gamma-ray shielding tests were executed to confirm the effectiveness of the ship's shielding performance. The tests confirmed that effective shielding has been achieved and the dose equivalent rate in the accommodation and other inhabited spaces is sufficiently lower than the regulated limitations. This was achieved by employing the appropriate calculation methods and shielding materials. (author)

Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

International Nuclear Information System (INIS)

Dreesen, K.; Goetze, H.G.; Holst, L.; Gerstenberg, H.; Schreckenbach, K.

2000-01-01

According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was -5 Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

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)

Operation and maintenance of the T-3 cask system

International Nuclear Information System (INIS)

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

1983-01-01

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

Past experience and future needs for the use of burnup credit in LWR fuel storage

International Nuclear Information System (INIS)

Boyd, W.A.; Wrights, G.N.

1987-01-01

To achieve improved fuel economics and reduce the amount of fuel discharged annually, utilities are engaging in fuel management strategies that will achieve higher discharge burnups for their fuel assemblies. Although burnup credit methodologies have been developed and spent-fuel racks have been licensed, burnup credit fuel storage racks are not the answer for all utilities. Off-site and out-of-pool spent-fuel storage may be more appropriate. This is leading to the development of dry spent-fuel storage and shipping casks. Cask designs with spent-fuel storage capability between 20 and 32 assemblies are being developed by several vendors. The US Dept. of Energy is also funding work by VEPCO. Westinghouse is currently licensing its dry storage cask, developing a shipping cask for the domestic market, and is involved in a joint venture to develop a cask for the international market. Although methods of taking credit for fuel burnup in spent-fuel storage racks have been developed and licensed, use of these methods on dry spent-fuel storage and shipping casks can lead to new issues. These issues arise because the excess reactivity margin that is inherent in a burnup credit spent-fuel storage rack criticality analysis will not be available in a dry cask analysis

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

International Nuclear Information System (INIS)

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

1986-02-01

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

Application of advanced handling techniques to transportation cask design

International Nuclear Information System (INIS)

Bennett, P.C.

1992-01-01

Sandia National Laboratories supports the US Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) applying technology to the safe transport of nuclear waste. Part of that development effort includes investigation of advanced handling technologies for automation of cask operations at nuclear waste receiving facilities. Although low radiation levels are expected near transport cask surfaces, cumulative occupational exposure at a receiving facility can be significant. Remote automated cask handling has the potential to reduce both the occupational exposure and the time necessary to process a cask. Thus, automated handling is consistent with DOE efforts to reduce the lifecycle costs of the waste disposal system and to maintain public and occupational radiological risks as low as reasonably achievable. This paper describes the development of advanced handling laboratory mock-ups and demonstrations for spent fuel casks. Utilizing the control enhancements described below, demonstrations have been carried out including cask location and identification, contact and non-contact surveys, impact limiter removal, tiedown release, uprighting, swing-free movement, gas sampling, and lid removal operations. Manually controlled movement around a cask under off-normal conditions has also been demonstrated

Design considerations for on-site spent-fuel transfer systems

International Nuclear Information System (INIS)

Jones, R.H.; Jones, C.R.

1989-06-01

Studies on spent fuel shipping logistics and operation make it clear that the use of large casks, i.e., 100--125 tons, is superior to smaller casks of similar construction. This superiority manifests itself in both transportation and/or shipping economics and safety as well as reduced personnel exposure in the processing of the casks. An on-site system for the transfer of spent fuel from the storage pool to a large shipping or storage cask, as well as the transfer of spent fuel directly from a storage cask to a shipping cask, could bring the large cask benefits to those restricted reactors. Sensing the need to look more closely at this opportunity, EPRI contracted with S. Levy, Incorporated of Campbell, CA to develop a set of design considerations for such transfer systems. Rather then embark on another design study, EPRI decided to first identify the system considerations that must be factored into any design. The format for this effort presents both the Consideration and the Rationale for the consideration. The resulting work identified thirty-six General Considerations and two Special Considerations. The Considerations are in the form of mandatory requirements and desirable but nonmandatory requirements. Additionally, a brief economic study was performed to get a feel for the cost considerations of on-site transfers. The study results suggest a relatively narrow set of scenarios where on-site transfers are economically superior to alternatives. These scenarios generally involve the use of concrete casks as on-site storage devices

Virginia Power and Department of Energy spent fuel transportation experience

International Nuclear Information System (INIS)

Ruska, M.D.; Schoonen, D.H.

1986-12-01

Spent fuel assemblies for the Spent Fuel Storage Cask Testing Program conducted by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) were transported to the INEL. A total of 69 spent fuel assemblies (23 shipments) were shipped from Virginia Power's nuclear power plant at Surry, Virginia, to the INEL between July 1985 and June 1986 to fill and test three spent fuel storage casks. The shipments were made over the highway system in Transnuclear, Inc., TN-8L shipping casks on specially constructed trailers. The shipments were moved by diesel tractors owned and operated by Tri-State Motor Transit Company of Joplin, Missouri. The gross vehicle weight for each shipment was 112,000 lb, which was a major consideration when selecting routes for the shipments. Cooperative negotiations with officials for the 17 states involved obtained authorization to transport through their states. The shipping campaign was successfully completed through close communication and cooperation and careful planning and operation by all organizations involved

Modification of SKYSHINE-III to include cask array shadowing

Energy Technology Data Exchange (ETDEWEB)

Hertel, N.E. [Georgia Institute of Technology, Atlanta, GA (United States); Pfeifer, H.J. [NAC International, Norcross, GA (United States); Napolitano, D.G. [NISYS Corporation, Duluth, GA (United States)

2000-03-01

The NAC International version of SKYSHINE-III has been expanded to represent the radiation emissions from ISFSI (Interim Spent Fuel Storage Installations) dry storage casks using surface source descriptions. In addition, this modification includes a shadow shielding algorithm of the casks in the array. The resultant code is a flexible design tool which can be used to rapidly assess the impact of various cask loadings and arrangements. An example of its use in calculating dose rates for a 10x8 cask array is presented. (author)

QA in the design and fabrication of the TMI-2 rail cask

International Nuclear Information System (INIS)

Hayes, G.R.

1988-01-01

EGandG Idaho, Inc., acting on behalf of the US Department of Energy, is responsible for transporting core debris from Three Mile Island-Unit 2 to the Idaho National Engineering Laboratory. Transportation of the debris is being accomplished using an NRC licensed container, called the NuPac 125-B. This paper describes the NuPac 125-B Rail Cask and the quality assurance (QA) requirements for that system. Also discussed are the QA roles of the various organizations involved in designing, building, inspecting and testing the NuPac 125-B. The paper presents QA/QC systems implemented during the design, procurement, and fabrication of the cask to assure compliance with all applicable technical codes, standards and regulations. It also goes beyond the requirements aspect and describes unique QA/QC measures employed to assure that the cask was built with minimum QA problems. Finally, the lessons learned from the NuPac 125-B project is discussed. 4 refs., 4 figs

Transport package maintenance requirements and operations

International Nuclear Information System (INIS)

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

1988-01-01

The NuPac 125-B rail cask, which transports the damaged core debris from Three Mile Island Unit 2 (TMI-2) to the Idaho National Engineering Laboratory (INEL), is the only new spent-fuel rail shipping cask to be licensed in the United States within the last decade. EG ampersand G Idaho, Inc. (EG ampersand G), acting on behalf of the US Department of Energy, is responsible for ensuring that those new casks and rail cars are properly maintained per regulatory requirements. Both the casks and rail cars have comprehensive in-service inspection and preventive maintenance programs, which are more involved than implied by the requirements. The TMI-2 shipping campaign is the most ambitious spent-fuel transport activity being conducted in the nuclear industry. The experience gained in this campaign, as it relates to maintenance of a transport system, should be of interest and have direct applicability to similar shipping activities planned in the years ahead

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

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)

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)

Sewage Solids Irradiator Transportation System (SSITS) cask: preliminary design description

International Nuclear Information System (INIS)

Eakes, R.G.; Kempka, S.N.; Lamoreaux, G.H.; Sutherland, S.H.

1983-02-01

The preliminary design of the Sewage Solids Irradiator Transportation System (SSITS) Cask is presented in this document. The SSITS cask is to be used for the transport of radioactive cesium chloride and strontium fluoride capsules which are of use in irradiators or as heat sources. The SSITS cask is approximately 1.4 m in diameter, 1.3 m high, weighs roughly 9 t, provides 33 cm of steel shielding, and can dissipate up to 5.2 kW of decay heat. The cask design criteria are identified and a description of the cask design and operation is provided. Detailed analyses of the design were performed to demonstrate licensability of the cask by the Nuclear Regulatory Commission (NRC). Results of the analyses indicate that the preliminary design is in compliance with the pertinent regulatory requirements for licensing of a radioactive material transportation container

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

FACSIM/MRS (Monitored Retrievable Storage)-2: Storage and shipping model documentation and user's guide

Energy Technology Data Exchange (ETDEWEB)

Huber, H.D.; Chockie, A.D.; Hostick, C.J.; Otis, P.T.; Sovers, R.A.

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 FACSIM/MRS-2, which is also referred to as the back-end model. The FACSIM/MRS-2 model simulates the MRS storage and shipping operations, which include handling canistered spent fuel and secondary waste in the shielded canyon cells, in onsite yard storage, and in repository shipping cask loading areas.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

Fiarman, S.

1987-01-01

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

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

International Nuclear Information System (INIS)

Chen, T.F.; Chen, J.C.; Witte, M.C.; Fischer, L.E.

1993-01-01

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

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.)

Fabrication and operational experience with the interim storage cask

International Nuclear Information System (INIS)

Scott, P.L.

1998-01-01

This paper discusses the fabrication and operational experience of the Interim Storage Cask (ISC). The ISC is a dry storage cask which is used to safely store a Core Component Container (CCC) containing up to seven Fast Flux Test Facility (FFTF) spent fuel assemblies at the US Department of Energy's Hanford Site. Under contract to B and W Hanford Company (BWHC), General Atomics (GA) designed and fabricated thirty ISC casks which BWHC is remotely loading at the FFTF facility. BWHC designed and fabricated the CCCS. As of December 1997, thirty ISCs have been fabricated, of which eighteen have been loaded and moved to a storage site adjacent to the FFTF facility. Fabrication consisted of three sets of casks. The first unit was completed and acceptance tested before any other units were fabricated. After the first unit passed all acceptance tests, nine more units were fabricated in the first production run. Before those nine units were completed, GA began a production run of twenty more units. The paper provides an overview of the cask design and discusses the problems encountered in fabrication, their resolution, and changes made in the fabrication processes to improve the quality of the casks. The paper also discusses the loading process and operational experiences with loading and handling of the casks. Information on loading times, worker dose exposure, and total dose for loading are presented

CASTOR-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed with vacuum, nitrogen, and helium backfills in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium backfills to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen backfills in a vertical orientation and with helium in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design expectation of less than 200 mrem/h. Cask surface dose rates of <75 mrem/h can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

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

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

Packaging and shipment of U.S. breeder reactor experiments

International Nuclear Information System (INIS)

Berger, J.D.

1980-01-01

Irradiation testing of fuels and materials in the Fast Test Reactor (FTR) required development of a shipping cask (designated T-3) and associated hardware for loading and shipping of these experiments to postirradiation examination facilities. The T-3 shipping-cask program included design, fabrication, and testing of internal cask packages to protect the experiments during loading, shipping, and unloading. The cask was designed for loading in both the vertical and horizontal attitudes

A compilation of structural property data for computer impact calculation (5/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 5 involve structural property data of wood. (author)

78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Science.gov (United States)

2013-03-18

... Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule... (MAGNASTOR[supreg]) System listing within the ``List of Approved Spent Fuel Storage Casks'' to include... for the MAGNASTOR[supreg] System cask design within the list of approved spent fuel storage casks that...

The use of burnup credit for spent fuel cask design

International Nuclear Information System (INIS)

Lake, W.H.

1993-01-01

A new generation of high capacity spent fuel transport casks is being developed by the U.S. Department of Energy (DOE) as part of the Federal Waste Management System (FWMS). Burnup credit, which recognizes the reduced reactivity of spent fuel is being used for these casks. Two cask designs being developed for DOE by Babcock and Wilcox and General Atomics use burnup credit. The cask designs must be certified by the U.S. Nuclear Regulatory Commission (NRC) if they are to be used in the FWMS. Certification of these casks by the NRC would not require any change in the NRC's transport regulations, and would be consistent with past practices. Furthermore, use of burnup credit casks appears to be consistent with current International Atomic Energy Agency (IAEA) rules and regulations. To support NRC certification, DOE has identified the technical issues related to burnup credit, and embarked on a development program to resolve them. (J.P.N.)

Studies and research concerning BNFP: cask handling equipment standardization

International Nuclear Information System (INIS)

McCreery, P.N.

1980-10-01

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

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

Consequences of postulated losses of LWR spent fuel and plutonium shipping packages at sea

International Nuclear Information System (INIS)

Heaberlin, S.W.; Baker, D.A.; Beyer, C.E.; Friley, J.R.; Mandel, S.; Peterson, P.L.; Sominen, F.A.

1977-10-01

The potential consequences of the loss of a large spent fuel cask and of a single 6M plutonium shipping package into the sea for two specific accident cases are estimated. The radiation dose to man through the marine food chain following the loss of undamaged and fire-damaged packages to the continental shelf and in the deep ocean are conservatively estimated. Two failure mechanisms that could lead to release of radioactive material after loss of packages into the ocean have been considered: corrosion and hydrostatic pressure. A third possible mechanism is thermal overpressurization following burial in marine sediments. It was determined that the seals or pressure relief devices on an undamaged spent fuel cask might fail from hydrostatic forces for losses on the continental shelf although some cask designs would retain their integrity at this depth. The population dose to man through the marine food chain following these scenarios has been estimated. The dose estimates are made relating the radioactive material released and the seafood productivity in the region of the release. Doses are based on a one-year consumption of contaminated seafood. The loss of a single plutonium package on the continental shelf is estimated to produce a population dose commitment of less than 250 man-rem for recycle plutonium. The dose commitment to the average individual is less than one millirem. Doses for losses of undamaged casks to the continental shelf and deep ocean and for loss of a fire-damaged cask to the deep ocean were determined to be several orders of magnitude smaller. 22 tables, 10 figures

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

International Nuclear Information System (INIS)

Bentele, W.; Kinzelmann, T.

1999-01-01

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

Method to mount defect fuel elements i transport casks

International Nuclear Information System (INIS)

Borgers, H.; Deleryd, R.

1996-01-01

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

Development of cask body integrated with bottom plate

International Nuclear Information System (INIS)

Yoshida, Takuji; Sasaki, Tomoharu; Koyama, Yoichi; Kumagai, Yasuyuki; Watanabe, Yuichi; Takasa, Seiju

2017-01-01

The main parts of a metal cask for storage and transport of spent nuclear fuel consists of main body, neutron shield material and external cylinder. The forged main body has been manufactured as a cup shape by welding of 'forged body' and 'forged bottom plate' which are independently forged. JSW has developed the manufacturing technology of 'cask body integrated with bottom plate' which has no weld line with the goal of cost reduction, manufacturing period shortening and further reliability improvement. Manufacturing for the prototype of 'cask body integrated with bottom plate' has completed to verify mechanical properties and uniformity of the product which satisfy the specified values stipulated in JSME Code S FA1 2007 edition. Here, we report the manufacturing technology and obtained properties of 'cask body integrated with bottom plate'. (author)

Method for decreasing radiation hazard in transporting radioactive material

International Nuclear Information System (INIS)

Wodrich, D.D.

1975-01-01

At the end of their useful life, fuel rods are removed from a nuclear reactor and transferred underwater into a shipping cask. The water in the pool of the nuclear reactor system (or fuels reprocessing plant) may contain troublesome amounts of radioactive isotopes, creating biological hazards for any shipping cask unless adequately cleaned after contacting pool water. Potential contamination of substantially all of the entire exterior of the shipping cask is avoided because such shipping cask is at least predominantly immersed in fresh water within a vertically shiftable container which can be, for example, shifted between the bottom and the surface of the pool. Fresh water is supplied to the interior of the shiftable container whereby substantially all of the exterior of the shipping cask is immersed in fresh water, maintained at a pressure and/or flow velocity preventing the pool water from contacting the exterior of the shipping cask

Development of concrete cask storage technology for spent nuclear fuel

International Nuclear Information System (INIS)

Saegusa, Toshiari; Shirai, Koji; Takeda, Hirofumi

2010-01-01

Need of spent fuel storage in Japan is estimated as 10,000 to 25,000 t by 2050 depending on reprocessing. Concrete cask storage is expected due to its economy and risk hedge for procurement. The CRIEPI executed verification tests using full-scale concrete casks. Heat removal performances in normal and accidental conditions were verified and analytical method for the normal condition was established. Shielding performance focus on radiation streaming through the air outlet was tested and confirmed to meet the design requirements. Structural integrity was verified in terms of fracture toughness of stainless steel canister for the cask of accidental drop tests. Cracking of cylindrical concrete container due to thermal stress was confirmed to maintain its integrity. Seismic tests of concrete cask without tie-down using scale and full-scale model casks were carried out to confirm that the casks do not tip-over and the spent fuel assembly keeps its integrity under severe earthquake conditions. Long-term integrity of concrete cask for 40 to 60 years is required. It was confirmed using a real concrete cask storing real spent fuel for 15 years. Stress corrosion cracking is serious issue for concrete cask storage in the salty air environment. The material factor was improved by using highly corrosion resistant stainless steel. The environmental factor was mitigated by the development of salt reduction technology. Estimate of surface salt concentration as a function of time became possible. Monitoring technology to detect accidental loss of containment of the canister by the stress corrosion cracking was developed. Spent fuel integrity during storage was evaluated in terms of hydrogen movement using spent fuel claddings stored for 20 years. The effect of hydrogen on the integrity of the cladding was found negligible. With these results, information necessary for real service of concrete cask was almost prepared. Remaining subject is to develop more economical and rational

What drives Greek consumer preferences for cask wine?

DEFF Research Database (Denmark)

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

2012-01-01

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

Dry reloading and packaging of spent fuel at TRIGA MARK I reactor of Medical University Hanover (MHH), Germany

International Nuclear Information System (INIS)

Haferkamp, D.

2008-01-01

Between 1994 and 1998 the equipment for dry reloading of a research reactor was developed by Noell, which was funded by the German Federal Government and State of Saxonia. The task of this development programme was the design and delivery of an equipment able to load the spent fuel into the shipping casks in a dry mode for research reactors, where wet loading inside the storage pool is impossible. ALARA and infrastructure conditions had to be taken into consideration. Most of the research reactors of TRIGA MARK I type or WWR-SM have operating modes for handling of spent fuel inside the pond or for transfer of spent fuel from pond to dry/wet storage pools. On the other hand, most of them cannot handle heavy weighted shipping casks inside the reactor building because of the crane capacity, or inside water pool because of dimensions and weight of shipping casks. A typical licensed normal operating procedure for spent fuel in research reactors (TRIGA MARK I) is shown. Dry unloading procedure is described. Additionally to the normal operating procedures at the MHH research reactor the following steps were necessary: - dry packaging of spent fuel elements into the loading units (six packs) in order to minimise the transfer and loading steps between the pool and shipping cask; - transfer of spent fuel loading units from dry storage pool to the shipping cask (outside the reactor building) in a shielded transfer cask; - dry reloading of loading units, into the shipping casks outside the reactor building. The Dry Reloading Equipment implies the following 5 items: 1. loading units (six packs), which includes: - capacity up to six spent fuel elements; - criticality safe placement of spent fuel elements; - handling of several spent fuel elements in an aluminium loading unit. 2. Special Transfer Cask, which includes: - shielded housing with locks; - gripper inside housing; - hoist outside housing; - computer aided operation mode for loading and unloading. 3. Transfer Vehicle

NUHOMS registered - MP197 transport cask

International Nuclear Information System (INIS)

Shih, P.; Sicard, D.; Michels, L.

2004-01-01

The NUHOMS registered -MP197 cask is an optimized transport design which can be loaded in the spent fuel pool (wet loading) or loaded the canister from the NUHOMS concrete modules at the ISFSI site. With impact limiters attached, the package can be transported within the states or world-wide. The NUHOMS registered -MP197 packaging can be used to transport either BWR or PWR canisters. The NUHOMS registered -MP197 cask is designed to the ASME B and PV Code and meets the requirements of Section III, Division 3 for Transport Packaging. The cask with impact limiters has undergone drop testing to verify the calculated g loadings during the 9m drops. The test showed good correlation with analytical results and demonstrate that the impact limiters stay in place and protect the package and fuel during the hypothetical accidents

Neutron and Gamma Shielding Evaluation for KN-12 Spent Nuclear Fuel Transport Cask

Energy Technology Data Exchange (ETDEWEB)

Cho, I. J.; Min, D. K.; Lee, J. C.; You, G. S.; Yoon, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, G. H.; Jeong, Y. C.; Ko, Y. W. [Korea Hydro and Nuclear Power Co., LTD., Kori (Korea, Republic of)

2007-07-01

The CASTOR KN-12 is designed to transport 12 intact PWR spent fuel assemblies for dry and wet transportation conditions. The overall cask length is 480.1 cm with a wall thickness 37.5 cm. Shield for the KN-12 is maintained by the thick walled cask body and the lid. For neutron shielding, polyethylene rods (PE) are arranged in longitudinal boreholes in the vessel wall and PE-plates are inserted between the cask lid and lid side shock absorber and between the cask bottom and bottom steel plate. The shielding evaluation of the cask has been performed with MCNP to confirm the shielding integrity of cask for pre-service inspection of transport cask.

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

International Nuclear Information System (INIS)

Powell, F.P.

1995-04-01

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

A compilation of structural property data for computer impact calculation (1/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi; Nagata, Norio.

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 1 involve structural property data and data processing computer program. (author)

A compilation of structural property data for computer impact calculation (3/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 3 involve structural property data of stainless steel. (author)

A compilation of structural property data for computer impact calculation (2/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 2 involve structural property data of mild steel. (author)

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.

Standard casks for the transport of LWR spent fuel

International Nuclear Information System (INIS)

Blum, P.

1986-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufactured in different countries and are presently used for european and intercontinental transports. The main advantages of these casks are: large payload, moderate cost, reliability, standardisation facilitating fabrication, operation and spare part supply [fr

Transport casks help solve spent fuel interim storage problems

International Nuclear Information System (INIS)

Dierkes, P.; Janberg, K.; Baatz, H.; Weinhold, G.

1980-01-01

Transport casks can be used as storage modules, combining the inherent safety of passive cooling with the absence of secondary radioactive waste and the flexibility to build up storage capacity according to actual requirements. In the Federal Republic of Germany, transport casks are being developed as a solution to its interim storage problems. Criteria for their design and licensing are outlined. Details are given of the casks and the storage facility. Tests are illustrated. (U.K.)

Estimated risk contribution for dry spent fuel storage cask

International Nuclear Information System (INIS)

Santos, C.; Kirk, M.T.; Abramson, L.; Guttmann, J.; Hackett, E.; Simonen, F.A.

2001-01-01

The U.S. Nuclear Regulatory Commission (NRC) is pursuing means to risk-inform its regulations and programs for dry storage of spent nuclear fuel. In pursuit of this objective, the NRC will develop safety goals and probabilistic risk assessments for implementing risk-informed programs. This paper provides one example method for calculating the risk of a dry spent fuel storage cask under normal and accident conditions. The example is on the HI-STORM 100 cask at a proposed site containing four thousand such casks. The paper evaluates the risk to the public by determining the likelihood a welded stainless steel container will leak. In addition, the study addresses the risk at a site where 4,000 casks may be stored until the U.S. Department of Energy accepts the casks for placement in a repository. The methods used employ the PRODIGAL computer code to assess the probability of a faulty weld on a stainless steel-welded canister. These analyses are only the initial stages of a comprehensive risk study that the NRC is performing in support of its regulatory initiatives. (author)

Optimization of cask for transport of radioactive material under impact loading

Energy Technology Data Exchange (ETDEWEB)

Sharma, Kuldeep, E-mail: kuldeep.brit@gmail.com [Indian Institute of Technology Bombay (India); Pawaskar, D.N.; Guha, Anirban [Indian Institute of Technology Bombay (India); Singh, R.K. [Bhabha Atomic Research Center (India)

2014-07-01

Highlights: • Cost and weight are important criteria for fabrication and transportation of cask used for transportation of radioactive material. • Reduction of cask cost by modifying few cask geometry parameters using complex search method. • Maximum von Mises stress generated and deformation after impact as design constraints. • Up to 6.9% reduction in cost and 4.6% reduction in weight observed in the examples used. - Abstract: Casks used for transporting radioactive material need to be certified fit by subjecting them to a specific set of tests (IAEA, 2012). The high cost of these casks gives rise to the need for optimizing them. Conducting actual experiments for the process of design iterations is very costly. This work outlines a procedure for optimizing Type B(U) casks through simulations of the 9 m drop test conducted in ABAQUS{sup ®}. Standard designs and material properties were chosen, thus making the process as realistic as reasonable even at the cost of reducing the options (design variables) available for optimization. The results, repeated for different source cavity sizes, show a scope for 6.9% reduction in cost and 4.6% reduction in weight over currently used casks.

Research on localization and alignment technology for transfer cask

Energy Technology Data Exchange (ETDEWEB)

Wang, Jingchuan, E-mail: jchwang@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, Shanghai (China); Yang, Ming; Chen, Weidong [Department of Automation, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, Shanghai (China)

2015-10-15

Highlights: • A method for the alignment between TB and HCB based on localizability is proposed. • A localization method based on the localizability estimation is proposed to realize the cask's localization accurately and ensures the transfer cask's accurate docking in the front of the window of Tokmak Building. • The experimental results show that the proposed algorithm works well in the indoor simulation environment. This system will be test in EAST of China. - Abstract: According to the long length characteristics of transfer cask compared to the environment space between Tokmak Building (TB) and HCB (Hot Cell Building), this paper proposes an autonomous localization and alignment method for the internal components transportation and replacement. A localization method based on the localizability estimation is used to realize the cask's localization and navigation accurately. Once the cask arrives at the front of the TB window, the position and attitude measurement system is used to detect the relative alignment error between the seal door of pallet and the window of TB real-time. The alignment between seal door and TB window could be realized based on this offset. The simulation experiment based on the real model is designed according to the real TB situation. The experiment results show that the proposed localization and alignment method can be used for transfer cask.

Concrete storage cask for interim storage of spent nuclear fuel

International Nuclear Information System (INIS)

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

2004-01-01

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

CERCA 01: a new safe multi-design MTR transport cask

Energy Technology Data Exchange (ETDEWEB)

Faure-Geors, B.S. [Framatome ANP Nuclear Fuel, CERCA, F-26104 Romans (France); Doucet, M.E. [Framatome ANP Nuclear Fuel, F-69006 Lyon (France)

2001-07-01

CERCA, a subsidiary company of FRAMATOME ANP, manufactures fuel for research reactors all over the world. To comply with customer requirements, fabrication of material testing reactors elements is a mixed of various parameters. Worldwide transportation of elements requires a flexible cask, which accommodates different designs and meets international transportation regulations. To be able to deliver most of fuel elements, and to cope with non-validation of casks used previously, CERCA decided to design its own cask. All regulatory tests were successfully performed. They completely validated and qualified the safety of this new cask concept. No matter the accidental conditions are, a 5 % {delta}K subcriticality margin is always met.

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)

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

Signatures of Extended Storage of Used Nuclear Fuel in Casks

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-28

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

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

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

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

NAC international dry spent fuel transfer technology

International Nuclear Information System (INIS)

Shelton, Thomas A.; Malone, James P.; Patterson, John R.

1996-01-01

Full text: For more than ten years NAC International (NAC) has designed, fabricated, tested and operated a variety of Dry Transfer Systems (DTS's) to transfer spent nuclear fuel from facilities with limited crane capabilities, limited accesses or limiting features to IAEA and USNRC licensed spent fuel transport casks or vice-versa. These DTS's have been operated in diverse environments in the United States and throughout the world and have proven to be a significant enhancement in transferring fuel between spent fuel pools, dry storage and hot cell facilities and spent fuel transport casks. Over the years, NAC has successfully and safely transferred more than two thousand fuel assemblies in DTS's. Our latest generation DTS incorporates years of extensive design and operating experience. It consists of a transfer cask with integrated fuel canister grapple, fuel canisters, and facility and cask adapters as well as a complement of related tools and equipment. The transfer cask is used to move irradiated HEU and LEU MTR fuel onsite in those instances where direct loading or unloading of the shipping cask is not possible due to dimensional, weight or other restrictions. The transfer cask is used to move canisters of fuel from the fuel storage location to the shipping cask. Adapters are employed to ensure proper interfacing of the transfer cask with fuel storage locations and shipping casks (NAC-LWT and NLI-1/2). Our existing fuel storage location adapter is designed for use with a storage pool; however, site or equipment specific adapters can easily be developed to allow interfacing with virtually any storage facility. Prior to movement of the first fuel canister in the transfer cask, the shipping cask is prepared for loading by proper set up of the base plate, shipping cask and shipping cask adapter. The fuel canisters are loaded with fuel and then retracted into the transfer cask via the fuel storage location adapter. The transfer cask is then moved to the shipping

Size and transportation capabilities of the existing US cask fleet

International Nuclear Information System (INIS)

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

1990-01-01

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

GA-4 half-scale cask model fabrication

International Nuclear Information System (INIS)

Meyer, R.J.

1995-01-01

Unique fabrication experience was gained during the construction of a half-scale model of the GA-4 Legal Weight Truck Cask. Techniques were developed for forming, welding, and machining XM-19 stainless steel. Noncircular 'rings' of depleted uranium were cast and machined to close tolerances. The noncircular cask body, gamma shield, and cavity liner were produced using a nonconventional approach in which components were first machined to final size and then welded together using a low-distortion electron beam process. Special processes were developed for fabricating the bonded aluminum honeycomb impact limiters. The innovative design of the cask internals required precision deep hole drilling, low-distortion welding, and close tolerance machining. Valuable lessons learned were documented for use in future manufacturing of full-scale prototype and production units

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

The dry spent RBMK fuel cask storage site at the Ignalina NPP in Lithuania

International Nuclear Information System (INIS)

Penkov, V.V.; Diersch, R.

1999-01-01

At present, there are about 15,000 spent RBMK fuel assemblies stored in the water pools near the reactors at the Ignalina Nuclear Power Plant (INPP). Part of them are cut in two bundles and stored in standardized baskets in the pools. Each basket is loaded with 102 bundles. For long-term interim storage of this fuel, it was decided to use dry storage in casks. For this reason, the total activity to be stored is split into individual units (casks). Each cask represents a closed and independent safety system, fulfilling all safety-relevant requirements for both normal operational and hypothetical accidental conditions. The main safety relevant features of the storage cask system are: (1) Inherent safety system; (2) Double barrier system; (3) Passive cooling by natural convection; (4) Safety against accidents. The cask dry storage system is a cost effective and multi-functional system for storage, transport after the operation time and final disposal under consideration of additional protective elements. From an economical point of view, cask storage has a number of advantages. Two cask types have been intended for the INPP storage site: (1) The CASTOR RBMK cask made of ductile cast iron; (2) The CONSTOR RBMK sandwich cask made of an inner and outer steel shell and reinforced heavy concrete. The CASTOR RBMK and the CONSTOR RBMK casks are designed to withstand severe storage site accidents and with help of impact limiters - to fulfil the IAEA test criteria for type B(U)F packages. The INPP spent RBMK fuel storage site is designed as an open air storage for an operational time of 50 years. The casks are arranged on the concrete storage pad. The site is equipped with a crane for cask handling and technological buildings and security systems. The safety analyses for fuel and cask handling and for cask handling and for cask technology at the site have been made and accepted by the Lithuanian Competent Authority. (author)

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

Drop test of reinforced concrete slab onto storage cask

International Nuclear Information System (INIS)

Kato, Y.; Hattori, S.; Ito, C.; Sirai, K.; Ozaki, S.; Kato, O.

1993-01-01

In this research, drop tests onto full-scale casks considering the specifications of a falling object (weight, construction, drop height, etc.) demonstrate and evaluate the integrity of casks in case a heavy object drops into the storage facilities. (J.P.N.)

Radiation Templates of Spent Fuel in Casks

Energy Technology Data Exchange (ETDEWEB)

Vanier, Peter

2018-05-07

BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, but baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.

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.

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

International Nuclear Information System (INIS)

Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N.

2004-01-01

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

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

International Nuclear Information System (INIS)

Weiss, M.

2001-01-01

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

Standard casks for the transport of LWR spent fuel

International Nuclear Information System (INIS)

Blum, P.

1985-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufacturer under TRANSNUCLEAIRE supervision in different countries and are presently used for European and intercontinental transports. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardisation facilitating fabrication, operation and spare part supply [fr

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

Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

International Nuclear Information System (INIS)

Diggs, J.M.

1985-01-01

This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Spent fuel transport cask thermal evaluation under normal and accident conditions

Energy Technology Data Exchange (ETDEWEB)

Pugliese, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Lo Frano, R., E-mail: rosa.lofrano@ing.unipi.i [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Forasassi, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy)

2010-06-15

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the '80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are

Developing new transportable storage casks for interim dry storage

International Nuclear Information System (INIS)

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

2004-01-01

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

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.)

Dry storage systems using casks for long term storage in an AFR and repository

International Nuclear Information System (INIS)

Einfeld, K.; Popp, F.W.

1986-01-01

In conclusion it can be stated that two basic routes with respect to spent fuel storage casks are feasible. One is the Multiple Transport Cask, which with certain modifications can be upgraded to meet the criteria for intermediate storage. Its status is characterized by the licensing of several types of Castor Casks for an intermediate storage period of 30 years in the AFR Storage Facility of DWK at Gorleben in the FRG. The other one is the Final Disposal (Repository) Cask, which can be made suitable for long term storage before a final decision with respect to a repository application is taken. The licensing procedure for a Pilot Conditioning Facility with the Pollux Cask System as reference case will be initiated by DWK in the near future. Under the assumption that in addition to the present Multiple Transport/Storage Casks a license for a Final disposal Cask with respect to long term storage is available, the relative merits of different cask storage systems would have to be evaluated

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

International Nuclear Information System (INIS)

Funke, Th.; Henig, Ch.

2008-01-01

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

Shielding tests for a new ship for the transport of spent nuclear fuels

International Nuclear Information System (INIS)

Ito, D.; Kitano, T.; Akiyama, H.; Ueki, K.; Sanui, T.

1998-01-01

a new ship for the transport of spent nuclear fuels which uses serpentine concrete as its major shielding material has been constructed. The shielding calculations are based on DOT3.5 code (CCC-276) and the DLC23). Experiments with Cf-252 and Co-60 sources were carried out to confirm the validity of this method of calculating the shielding effectiveness of serpentine concrete. In these experiments, neutron and gamma-ray dose equivalent rates were measured in various arrangements to simulate the shielding structures of the ship, the calculations for each arrangement were performed by this shielding calculation method. For both neutron and gamma-rays, the calculation results agreed with the experiments very well, confirming that this calculation method used in the ship's shielding design is valid. Two kinds of on-board gamma-ray shielding tests were performed to confirm the ship's actual shielding effectiveness. In one kind of test, gamma-ray dose equivalent rates were measured for each shielding wall using Co-60 sources. In the other kind of test, gamma-ray dose equivalent rates in the ship's accommodation area were measured when a strong Co-60 source was placed in a loaded shipping cask's position. In both gamma-ray shielding tests all measured dose equivalent rates were less than the calculated values, confirming that the ship's actual shielding is sufficient to meet safety requirements. (authors)

European experience with spent fuel transport

International Nuclear Information System (INIS)

Hunter, I.A.

1995-01-01

Nuclear Transport Ltd has transported 5000 tonnes of spent fuel from 35 reactors in 8 European countries since 1972. Transport management is governed by the Quality Plan for: transport administration, packaging and shipment procedures at the shipping plant, operations at the power plant, and packaging and shipment organization at the power plant. Selection of a suitable carrier device is made with regard to the shipping plant requirements, physical limitations of the reactor, fuel characteristics, and transport route constraints. The transport plan is set up taking into account exploitation of the casks, reactor shut-down requirements, fuel acceptance plans at the reprocessing plant, and cask maintenance periods. A transport cycle involving spent fuel shipment to La Hague or to Sellafield takes typically two or four weeks, respectively. Most transports through Europe are by rail. A special-design railway ferry boat serves transports to the United Kingdom. Both wet or dry casks are employed. Modern casks are designed for high burnups and for oxide fuels. (J.B.)

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01