WorldWideScience

Sample records for casks

  1. Testing of Metal Cask and Concrete Cask

    International Nuclear Information System (INIS)

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

  2. A cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    1988-03-01

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

  3. A cask fleet operations study

    International Nuclear Information System (INIS)

    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

  4. Cask storing facility

    International Nuclear Information System (INIS)

    The present invention provides a facility suitable to keeping and storing of casks for transporting and storing spent fuels generated from power plants and radioactive wastes generated from spent fuel reprocessing plants. Namely, the casks are transported in and out by a portal crane when they are stored. The cask storage space is disposed underground and soils are used as a portion of shielding materials. Then, a portal crane gives less load on the storage building when it is used compared with a case of using an overhead traveling crane. Since the storage pits are disposed underground, the radiation released from the casks in lateral and downward directions can be shielded by the soils. If shielding lids are disposed on the upper portion of the cask storage pits, upward radiation released from the casks can be shielded. Accordingly, there is no need to ensure thickness of walls of the building and ceilings for shielding. As a result, construction cost for the building can be reduced. (I.S.)

  5. Cask development, testing, and licensing

    International Nuclear Information System (INIS)

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

  6. Cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

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

  7. Cask fleet operations study

    International Nuclear Information System (INIS)

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

  8. Design and operational experience of dry cask storage systems

    International Nuclear Information System (INIS)

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

  9. 78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-12-27

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

  10. SNF shipping cask shielding analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-01-01

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

  11. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

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

  12. GNS spent fuel cask experience

    International Nuclear Information System (INIS)

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

  13. GNS spent fuel cask experience

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-05-01

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

  14. Initiatives in transport cask licensing

    International Nuclear Information System (INIS)

    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)

  15. Initiatives in transport cask license

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  16. Low-cost/high-integrity waste casks

    International Nuclear Information System (INIS)

    The MOSAIK cast iron casks for storage and transportation of waste have the following advantages: much higher activity content with a lower total volume compared with concrete waste packages; good shielding in connection with automated filling or underwater loading techniques leads to dose exposure reduction of the operating personnel; high cask integrity guarantees a tight containment and makes an additional fixation of the waste in the cask cavity unnecessary; and the low serial production costs of cast iron casks and the resulting volume reduction using these casks lead to a cost advantage under German licensing conditions. 5 figures

  17. Source storage and transfer cask: Users Guide

    International Nuclear Information System (INIS)

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

  18. Rationalizing transport operations: The TN 24 transport storage cask approach

    International Nuclear Information System (INIS)

    The number of transports of spent fuel interim storage casks can be reduced by improved standardized cask design. Optimization of cask design is based on two main technological choices: shielding and spent fuel support basket design. The approaches to optimizing cask design to improve payload is described for the Transnucleaire TN24 family of dual purpose transport and storage casks. (author)

  19. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

    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

  20. Development of high capacity transportable storage cask

    International Nuclear Information System (INIS)

    Mitsubishi Heavy Industries have developed high performance and reliable transportable storage casks, MSF series casks. The casks have employed newly developed materials that have been expressly developed to obtain long-term stability and quality. Furthermore, the casks have been employed newly designed structure to maximize payload of accommodating fuel assemblies in order to increase economic efficiency of storing spent fuels. The casks have been applied the following technologies. Basket assembly of the cask is made of newly developed boronated aluminum. The boronated aluminum is manufactured by power metallurgy process to provide uniformity of metallic structure and artificial aging which causes deterioration under high temperature condition is not applied to provide the boronated aluminum with high stability for long-term use. For the cask for BWR fuel, simplified basket whose design is that basket consists of some individual squire pipes without assembling is adopted in the cask. Neutron shielding material of the cask is made of newly resin of which raw materials have been modified to improve durability. Monolithic forging method which is how to shape steel into vessel form is developed to skip welding process between body shell and base plate and to improve reliability. Internal face of the body forging is machined to provide steps' in its cross section in order to fit the external shape of basket assembly and so heat dissipation performance is greatly improved. The new technologies have been done demonstration test in order to confirm that MSF series casks satisfy transport regulations. (author)

  1. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    International Nuclear Information System (INIS)

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

  2. 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.

  3. Genetics Home Reference: CASK-related intellectual disability

    Science.gov (United States)

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

  4. Development of cask and transportation system

    International Nuclear Information System (INIS)

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

  5. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

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

  6. Transportation capabilities of the existing cask fleet

    International Nuclear Information System (INIS)

    This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 465 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the DOE consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated

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

    International Nuclear Information System (INIS)

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

  8. Used Fuel Cask Identification through Neutron Profile

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Laboratory

    2015-11-20

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

  9. Life cycle cost report of VHLW cask

    International Nuclear Information System (INIS)

    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

  10. Shielding Analysis of the 5320 Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-01

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

  11. Design review report FFTF interim storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.L.

    1995-01-03

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

  12. Design review report FFTF interim storage cask

    International Nuclear Information System (INIS)

    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

  13. 78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

  16. BR-100 spent fuel shipping cask development

    International Nuclear Information System (INIS)

    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

  17. Transport and interim storage casks in Switzerland

    International Nuclear Information System (INIS)

    Full text: The Swiss utilities have chosen two different ways for the management of their spent fuel after initial on-site cooling: either reprocessing at La Hague plant (COGEMA) and Sellafield plant (COGEMA); or interim storage at the Central Interim Storage Facility called 'Zwischenlager Wuerenlingen AG' ( ZWILAG). Following international call for tenders, COGEMA LOGISTICS were awarded contracts for the supply of dual-purpose transport and storage casks for the interim storage of various spent fuel assemblies. All these casks belong to the family of the TN 24 dual purpose spent fuel storage casks in operation in the USA and in Belgium as well. They offer utilities a modular solution for the interim storage of spent fuel in robust metal casks which are fully suitable for off site transports. This flexible product can be readily adapted to suit individual user needs. The Leibstadt Nuclear Power Plant (KKL) has purchased six licensed dual-purpose TN97L spent fuel casks (97 BWR type fuel assemblies capacity). Three of them are already in operation at ZWILAG. COGEMA LOGISTICS has also delivered a dual-purpose TN52L spent fuel casks (52 BWR type fuel assemblies capacity) presently used for transport of spent fuel for reprocessing. The Goesgen Nuclear Power Plant (KKG) has purchased four licensed dual-purpose TN24G spent fuel casks (37 PWR type fuel assemblies capacity). They are all in operation at ZWILAG. The Muehleberg Nuclear Power Plant (BKW/KKM) has purchased 2 TN24BH spent fuel casks (69 BWR type fuel assemblies capacity). At the time of this abstract, cold trials are carried out involving the shuttle transport cask TN9/4 procured by COGEMA LOGISTICS as well. This paper will present the main features of these casks and the main steps of their development and implementation: 1) Main features of the casks: - The basic structure is a thick steel cylindrical forging with a welded on forged bottom and two forged steel lids. Containment and gamma shielding features of

  18. Design report for cask transportation equipment

    International Nuclear Information System (INIS)

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

  19. Cooling performance evaluation of the concrete cask

    International Nuclear Information System (INIS)

    The concrete cask storage system stores spent fuel by first sealing it within canisters and then containing such canisters inside a concrete cask. This report describes the results of a full-size model test performed to examine the heat dissipation characteristics of the concrete cask and to ascertain its ability to deal with elevated temperature. The specification to which a full-size concrete cask model was fabricated assumed an interim storage of 17x17UO2 fuel that was burned in PWR, estimating the heating value of spent fuel containing canister to be approximately 20 kW apiece. The test, which actually covered canister heating values ranging from 10 kW to 30 kW per unit to allow for temperature variations likely to be experienced in actual operation, verified that the concrete cask member did not exceed temperature limits. Test condition anticipated highest air temperature inside the spent fuel storage facility to be 30degC and, with reference to existing standard, set temperature limits of 65degC or less for the main body of concrete and 90degC or less for the local part as criteria. Preliminary 3-D thermo hydrodynamic analysis done prior to the test indicated that the temperature of the local part of the concrete cask member would be below 90degC. It also confirmed that steel material used as the structural member of the canisters or concrete cask would remain around 200degC even in an area where it was highest, validating that the integrity of such material would pose no problem from the analytical point of view. Heat dissipation performance test conducted in steady state verified that the concrete cask was able to have a sufficient cooling capacity against per-canister heating values in the 10 kW to 30 kW range which had been chosen in anticipation of temperature variation thought to be encountered in actual service. Also, to examine the consequence of the concrete cask having lost its cooling ability, another heat dissipation test was carried out under

  20. 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

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

    International Nuclear Information System (INIS)

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

  2. SNS Inner Plug Shipping Cask Analysis

    International Nuclear Information System (INIS)

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

  3. A cask maintenance facility feasibility study

    International Nuclear Information System (INIS)

    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

  4. Cask Processing Enclosure Specification/Operation - 12231

    International Nuclear Information System (INIS)

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

  5. 78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

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

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

    International Nuclear Information System (INIS)

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

  7. Advanced handling technology project and implications for cask design

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.)

  11. Spent Fuel Dry Storage Cask Thermal Test

    International Nuclear Information System (INIS)

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

  12. NUHOMS registered - MP197 transport cask

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    2013-12-06

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

  14. Economic evaluation of nuclear waste transportation casks

    International Nuclear Information System (INIS)

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

  15. Hexagonal absorption cask for nuclear power

    International Nuclear Information System (INIS)

    A hexagonal absorption cask for compact spent fuel storage is designed. The cask is made of austenitic stainless steel with a high boron content. One of the two sides of each of the six wall plates is longitudinally chamfered and attached to the inner face of the next wall plate in the hexagonal arrangement. The whole is welded together. This design secures that the absorption of the neutron flux in the radial direction will not be deteriorated if the boron content of the weld metal is reduced. (Z.S.). 2 figs

  16. Safety evaluation for packaging (onsite) SERF cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-10-24

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

  17. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  1. Seismic stability of unanchored spent nuclear fuel storage casks

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  3. Simplified computer codes for cask impact analysis

    International Nuclear Information System (INIS)

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

  4. Surface storage cask test summarization report

    International Nuclear Information System (INIS)

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

  5. Safety analysis of casks under extreme impact conditions

    International Nuclear Information System (INIS)

    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

  6. Evaluation of improvement potential for spent fuel cask handling

    International Nuclear Information System (INIS)

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

  7. Cask containing method for spent fuel assembly and subcriticality measuring device for a cask containing system

    International Nuclear Information System (INIS)

    An area for a spent fuel storage pool is sectioned into an ordinary rack area for disposing spent fuel assemblies taken out from a reactor core and a preliminary storage rack area having the same constitution as a cask for containing spent fuel assemblies. Preceding to cask-containment, the spent fuel assemblies are temporarily transferred once in the preliminary storing rack area from the ordinary rack area to ensure subcriticality and then contained in casks. In addition, those fuels having a higher burn-up degree are disposed coaxially to the central portion and those having not higher burn-up degree are disposed at the outer circumferential portion. The spent fuel assemblies can surely be contained in the casks, or the process of containing the spent fuel assemblies to the casks or the subcriticality after the containment can be evaluated thereby capable of further ensuring the subcriticality. The spent fuel assemblies can be transferred or stored safely and reliably at a good efficiency. (N.H.)

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

    International Nuclear Information System (INIS)

    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

  9. Inspection of Used Fuel Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-01

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

  10. Drop test of transportable storage cask

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  13. Interfacing the existing cask fleet with the MRS

    International Nuclear Information System (INIS)

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

  14. Facility for the decontamination of fuel element transport casks

    International Nuclear Information System (INIS)

    Before transporting them to the reprocessing plant the transport casks containing fuel assemblies are cleaned on the outside. This is achieved by means of a tank which can be closed and into which the transport cask is inserted. Within the tank the cask is standing on a roller drive mechanism with the aid of which it can be revolved. In the bottom, shell, and cover region there are adjustable nozzles by means of which the transport cask can be sprayed with washing water all-round. After cleaning, drying is performed at subatmospheric pressure by means of stationary hot-air nozzles also arranged in the tank. (DG)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 TNTM81 casks currently in use in Switzerland and the TNTM85 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 TNTM81 and TNTM85 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 TNTM81 and the TNTM85 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 TNTM28 cask and TS 28 V cask, all along the logistics chain of loading, unloading, transport and maintenance. In addition, years of feedback and experience in design and operations - together with ever improved materials - have allowed finding further optimization of this type of cask design. In order to increase the loading capacity in terms of radioactive source terms and heat load by 40%, the cask design relies on innovative solutions and benchmarks from the current shipping campaigns. Currently, TNTM81 and TNTM85 are the only licensed casks that can transport and store 28 canisters with a total decay heat of 56 kW. It contributes to optimise the number of required transports to bring back high level waste residues to their producers. Three units have already been loaded and transported to

  17. Rail tiedown tests with heavy casks for radioactive shipments

    International Nuclear Information System (INIS)

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

  18. MCO loading and cask loadout technical manual

    International Nuclear Information System (INIS)

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

  19. MCO loading and cask loadout technical manual

    Energy Technology Data Exchange (ETDEWEB)

    PRAGA, A.N.

    1998-10-01

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

  20. Concrete Spent Fuel Cask Criticality Calculation

    International Nuclear Information System (INIS)

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

  1. Dry storage cask - DIORIT - Swiss experience

    International Nuclear Information System (INIS)

    A new approach that uses wet-dry-dry loading technology has been successfully demonstrated, including remote viewing and loading control. The experience gained allowed for comprehensive expeditious licensing, including transport and storage permissions. A measurements campaign of more than two years has been made on the loaded cask, confirming: heat transfer, shielding, leaktightness and fuel behavior; in May 1985 the cask was transported from the reactor DIORIT to a new Away From Reactor-AFR-storage facility in the authors premises, which by the way is the first of its kind in operation licensed. In spite of the rather complex situation with the DIORIT spent fuel and the extreme limitations in the reactor handling and loading facilities, the complete project was achieved in only 22 months; accounting for two years of monitoring and measurements, the whole project took only 3.8 years. This shows the high degree of maturity achieved in spent fuel dry storage in transport cask, and reflects the high degree of technology innovation that has been demonstrated and which can be transferred as required

  2. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    International Nuclear Information System (INIS)

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

  4. Cermet Spent Nuclear Fuel Casks and Waste Packages

    International Nuclear Information System (INIS)

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

  5. Drop test of reinforced concrete slab onto storage cask

    International Nuclear Information System (INIS)

    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.)

  6. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... COMMISSION 10 CFR Part 72 RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc... U.S. Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by...

  7. Optimization of cask capacity for long term spent fuel storage

    International Nuclear Information System (INIS)

    Within the framework of the IAEA Subprogramme of Spent Fuel Management, a new project was conceived, focusing on issues associated with the optimization of cask/container loading (capacity) with respect to long term storage and the related integrity of fuel. An initial Consultants Meeting held in November 2002 identified and discussed principal issues regarding the optimization of cask/container assembly capacity and burnup/age capability in the design of systems for long term spent fuel storage and the related integrity of fuel. Based on resulting working materials, a Technical Meeting was held in March 2003 to obtain country-specific views from both regulators and implementers on this topic. Discussions focused on the following issues relevant to cask loading optimization: fuel integrity, retrievability, zoning, burnup credit, damaged fuel, computer code verification, life of cask components, cask maintenance, performance confirmation, and records management. Follow-on actions and meetings will be pursued to develop a TECDOC on this subject. (author)

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

    Science.gov (United States)

    2013-12-06

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

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

    Science.gov (United States)

    2013-12-26

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

  10. 75 FR 41369 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD Revision 1; Withdrawal

    Science.gov (United States)

    2010-07-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS HD Revision 1... HD cask system listing within the list of approved spent fuel storage casks to include Amendment No... ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to the CoC. Amendment No....

  11. 75 FR 41404 - List of Approved Spent Fuel Storage Casks: NUHOMS®

    Science.gov (United States)

    2010-07-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS HD Revision 1... cask system listing within the list of approved spent fuel storage casks to include Amendment No. 1 to... within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to the CoC....

  12. US cask requirements and industry capability survey

    International Nuclear Information System (INIS)

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

  13. Feasibility and incentives for burnup credit in spent-fuel casks

    International Nuclear Information System (INIS)

    The spent-fuel carrying capacities of previous-generation spent-fuel shipping casks have been primarily thermal and/or shielding limited. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced considerably and cask capacities become criticality limited. Using burnup credit in the design of future casks can result in increased cask capacities as well as reduced environmental impacts and savings in time and money

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  16. Nuclear cask testing films misleading and misused

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Nondestructive evaluation of monolithic transportation casks for spent nuclear fuel

    International Nuclear Information System (INIS)

    When spent fuel from nuclear reactors must be transported by rail or truck, Federal regulations require that it be enclosed in shipping casks that satisfy a number of stringent requirements. One configuration that is under consideration for such casks consists of monolithic metal cylinders approximately 17 ft. (5 m) long, 8 ft. (2.5 m) in diameter, with 14-in. (35-cm) thick walls. The casks are to be fabricated by casting or forging with one integrally closed end. The materials being considered for this application are austenitic steel, ferritic steel, and nodular cast iron. The thick walls are needed in order to absorb most of the radiation emitted by the contents. In addition, the casks must be capable of withstanding severe transportation accidents without a breach of the cask walls that would permit the escape of any radiation. The National Bureau of Standards conducted a study to evaluate the inspectability of the casks. The study showed that current NDE technology is adequate for inspecting the casks, provided that the inspection personnel are well trained in their respective methods, and that they are experienced with the equipment and their specific techniques, and have been properly qualified in this application. The capabilities of many NDE techniques were evaluated in the study. These techniques were based upon all of the principal NDE methods in use today, including ultrasonics, acoustic emission, radiology, liquid penetrants, magnetic particles, eddy currents, and visual inspection

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

    International Nuclear Information System (INIS)

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

  20. DOE procurement activities for spent fuel shipping casks

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. 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.

  4. Transportation cask contamination weeping: A program leading to prevention

    International Nuclear Information System (INIS)

    This paper describes the problem of cask contamination weeping, and efforts to understand the phenomenon and to eliminate its occurrence during spent nuclear fuel transport. The paper summarizes analyses of field experience and scoping experiments, and concentrates on current modelling and experimental validation efforts. The open-quotes weepingclose quotes phenomenon associated with spent fuel transportation casks (also known as open-quote sweatingclose quotes) is believed to be due to the conversion of fixed contamination on the external surface of the cask to a removable form. Spent fuel transportation casks are loaded under water at nuclear power plants in a spent fuel storage pool, exposing the cask surfaces to contamination by radionuclides present in the pool water including 137Cs, 134Cs, and 60Co. The external surfaces of loaded casks are routinely surveyed for removable contamination and decontaminated to 1/10 of the US and IAEA regulatory limits prior to being released for shipment (49CFR 1983, IAEA 1989). However, 3% to 8% of US spent fuel casks have arrived at final destinations with removable surface contamination in excess of that allowed by regulation, though many preshipment surveys have shown contaminant levels to be within allowable limits (Grella 1987). Attempts to reduce the incidence of weeping have met with limited success and resulted in time-consuming operational constraints and procedures that significantly increase cask processing times and occupational composure at loading facilities. As the US Department of Energy (DOE) moves toward a high volume spent fuel transportation campaign beginning in 1998, the elimination of weeping occurrence and minimization of operational constraints has received increased attention. A DOE program is underway at Sandia National Laboratories (SNL) to determine the physical and chemical processes involved in radionuclide contamination and release on transportation cask surfaces

  5. 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.

  6. Interim storage and transport casks in Switzerland. COGEMA logistics experience

    International Nuclear Information System (INIS)

    The Swiss utilities have chosen two different ways for the management of their spent fuel after initial on-site cooling: (1) reprocessing at La Hague plant (COGEMA) and Sellafield plant (BNFL); (2) interim storage at the Central Interim Storage Facility called 'Zwischenlager Wuerenlingen AG' ( ZWILAG). Following international call for tenders by Swiss utilities, COGEMA LOGISTICS has been awarded several contracts for the supply of dual-purpose transport and storage casks for the interim storage of various spent fuel assemblies. All these casks belong to the family of the TN 24 dual purpose spent fuel storage casks in operation in the USA and in Belgium as well. They offer utilities a modular solution for the interim storage of spent fuel in robust metal casks which are fully suitable for off site transports. This flexible product can be readily adapted to suit individual user needs. The Leibstadt Nuclear Power Plant (KKL) has purchased nine licensed dual-purpose TN 97L spent fuel casks (97 BWR type fuel assemblies capacity). Three of them are already in operation at ZWILAG. COGEMA LOGISTICS has also delivered a dual-purpose TN 52L spent fuel casks (52 BWR type fuel assemblies capacity) presently used for transport of spent fuel for reprocessing. The Goesgen Nuclear Power Plant (KKG) has purchased four licensed dual-purpose TN 24G spent fuel casks (37 PWR type fuel assemblies capacity). They are all in operation at ZWILAG. The Muehleberg Nuclear Power Plant (BKW/KKM) has purchased two TN 24BH spent fuel casks (69 BWR type fuel assemblies capacity). At the time of this abstract, cold trials are carried out involving the shuttle transport cask TN 9/4 procured by COGEMA LOGISTICS as well. (author)

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

    International Nuclear Information System (INIS)

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

  8. SCANS, Shipping Cask Design Safety Analysis

    International Nuclear Information System (INIS)

    1 - Description of program or function: SCANS (Shipping Cask Analysis System) is a microcomputer-based system of computer programs and databases for evaluating safety analysis reports on spent fuel shipping casks. SCANS calculates the global response to impact loads, pressure loads, and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. Analysis options are based on regulatory cases described in the Code of Federal Regulation (1983) and Regulatory Guides published by the NRC in 1977 and 1978. The system is composed of a series of menus and input entry cask analysis, and output display programs. An analysis is performed by preparing the necessary input data and then selecting the appropriate analysis: impact, thermal (heat transfer), thermally- induced stress, or pressure-induced stress. All data are entered through input screens with descriptive data requests, and, where possible, default values are provided. Output (i.e., impact force, moment and shear time histories; impact animation; thermal/stress geometry and thermal/stress element outlines; temperature distributions as iso-contours or profiles; and temperature time histories) is displayed graphically and can also be printed. 2 - Method of solution: Impact analyses use a one-dimensional dynamic beam model. Each node in the beam model has two translational and one rotational degrees of freedom. The impact code uses an explicit time-history integration scheme in which equilibrium is formulated in terms of the global external forces and internal force resultants. This formulation allows the code to track large rigid- body motion. Thus, the oblique impact problem can be calculated from initial impact through essentially rigid-body rotation to secondary impact. Lateral pressure due to lead-slump can also be calculated. Appropriate two-dimensional finite-element meshes are automatically generated for thermal, thermal-stress, and pressure- stress analyses, based on

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. ALARA studies on spent fuel and waste casks

    Energy Technology Data Exchange (ETDEWEB)

    Sutherland, S.H.

    1980-04-01

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

  12. ALARA studies on spent fuel and waste casks

    International Nuclear Information System (INIS)

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

  13. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    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 keff 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

  14. New generation legal weight spent fuel shipping cask

    International Nuclear Information System (INIS)

    GA Technologies has proposed two new spent fuel shipping casks that have a capacity four times greater than comparable existing designs. The new casks, for legal weight truck shipments, can carry four PWR or nine BWR spent fuel assemblies. They were offered in response to the recent request for proposals issued by the Office of Civilian Radioactive Waste Management (OCRWM). The RFP addressed a new generation of truck and rail shipping casks that could transport intact spent fuel assemblies from nuclear reactors to a repository or a Monitored Retrievable Storage (MRS) facility. Our primary goal has been to maximize the number of fuel elements of each fuel type that a LWT cask can carry, while ensuring that the design meets all licensing requirements

  15. Transport/storage cask TN 1300 technical description

    International Nuclear Information System (INIS)

    The TN 1300 cask - developed by Transnuklear GmbH, Hanau - serves as a cask for the dry transport and storage of spent fuel elements from 1300 MW light water reactors. The cask is classified as - Typ B(U) package - fissile class II. The application is filed with the PTB, the German competent authority. The cask has a maximum capacity of 12 PWR fuel elements (Biblis type) or 33 BWR fuel elements. The maximum heat dissipation (natural convection) amounts to about 50 kW. This corresponds to a cooling period of about 2.5 years for PWR fuel elements. The handling weight of the TN 1300 is approx. 116.5 t. (orig./HW)

  16. Handbook for structural analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. What drives Greek consumer preferences for cask wine?

    DEFF Research Database (Denmark)

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

    2012-01-01

    , taste, origin) as opposed to extrinsic cues (brand, price, convenience packaging). Research limitations/implications – Two main strategic directions are suggested to Greek cask wine producers: they can either maintain the current approach to the market by providing a “simple”, not particularly refined...... of the cask wine consumer is. This study aims at filling this gap. Design/methodology/approach – Based on a web-based survey, the best-worst scaling (BWS) method was applied to measure the importance of attributes that Greek consumers assign when choosing cask wine. Then, a latent class clustering...... analysis based on the importance ratings of the attributes was applied in order to segment the Greek cask wine market. Findings – The most important attributes were found to be price, quality and convenience packaging, whereas brand, grape variety and origin were found to be the least important ones. In...

  20. Operating Experience in Spent Fuel Storage Casks

    International Nuclear Information System (INIS)

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

  1. 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

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

    International Nuclear Information System (INIS)

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

  3. A cask maintenance facility feasibility study

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    CARLSON, A.B.

    1999-02-24

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

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

    International Nuclear Information System (INIS)

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

  6. 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)

  7. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

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

  8. TRANSPORTATION CASK RECEIPT AND RETURN FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-02-01

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

  11. Verification tests on cask-storage method for storing spent fuel at reactor

    International Nuclear Information System (INIS)

    CRIEPI has conducted a feasibility study on spent-fuel storage and has shown that shipping and storage casks are the best method for storing less than 500 tons of spent fuel at the reactor. The cask-storage facility is composed of a storage house and casks. The cask has sealing, heat conduction, shielding and criticality-prevention functions. The storage house is used for managing casks in it. In consideration of the above functions, we confirmed the integrity of cask and spent fuel under normal conditions and in hypothetical accident conditions. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    C. Sanders; D. Kimball

    2005-08-02

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

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

    International Nuclear Information System (INIS)

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

  15. 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)

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

    International Nuclear Information System (INIS)

    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)

  17. The Dry-Cap spent fuel storage/transport cask

    International Nuclear Information System (INIS)

    Increasing inventories of spent fuel and decreasing storage capacities at reactors are prompting development of alternative storage technologies. In the United States of America, the Department of Energy is engaged in the development of a geological repository and is committed to begin accepting fuel for permanent storage by 31 January 1998. Until this time, US utilities have assumed the responsibility for handling this material. The storage situation is also recognized in Japan and several utilities are now engaged in the development of alternative storage options. In recognition of these situations, Combustion Engineering, Inc. and Sumitomo Heavy Industries Ltd are engaged in a programme to develop and manufacture a cask capable of safety storing and transporting spent nuclear fuel. The cask is designed in accordance with US 10CFR71 and 10CFR72 criteria and has one of the largest capacities of spent fuel casks, with the ability to hold 24 PWR or 60 BWR spent fuel bundles and remain under the 125 t crane capacity of most power plants. The Dry-Cap spent fuel storage cask consists of a 16.5 ft. (5 m) long by 7.5 ft (2.27 m) diameter thick-walled steel cylinder surrounded by shielding material. Dry-Cap is a relatively simple design, easily manufactured and, unlike other cask designs, requires no external fins for cooling. Dissipation of decay heat is accomplished by natural convection between the fuel and its helium environment and the cask and its surrounding environment. One of the most important features of the Dry-Cap design is that it does not require poison material for criticality control, since the basket design utilizes credit for burnup. Taking credit for the known irradiation heating of discharged fuel, and the fact that it has a low residual reactivity, can simplify and minimize the maintenance and monitoring requirements for long term storage. The Dry-Cap cask is designed to fulfil the long and short term storage needs for utilities. (author)

  18. Development of the GA-4 and GA-9 legal weight spent fuel casks

    International Nuclear Information System (INIS)

    GA is nearing the completion of the final design of two legal weight truck spent fuel shipping casks, the GA-4 Cask for PWR fuel and the GA-9 Cask for BWR fuel. GA is developing the casks under contract to the US Department of Energy (DOE) Field Office, Idaho, as part of the Office of Civilian Radioactive Waste Management (OCRWM) Cask Systems Development Program (CSDP). The casks will transport intact spent fuel assemblies fro commercial nuclear reactors sites to a monitored retrievable storage facility or a permanent repository. The DOE initiated the Cask Systems Development Program in response to the Nuclear Waste Policy Act of 1982 which made DOE responsible for managing the program for permanent disposal of spent nuclear fuel and high-level waste. This paper describes developmental and design verification testing programs, and the present status of the GA-4 and GA-9 Cask designs

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. Vestibule and Cask Preparation Mechanical Handling Calculation

    Energy Technology Data Exchange (ETDEWEB)

    N. Ambre

    2004-05-26

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

  2. Final version dry cask storage study

    International Nuclear Information System (INIS)

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

  3. Initial version, dry cask storage study

    International Nuclear Information System (INIS)

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

  4. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Response of spent fuel transportation casks to explosive loadings

    International Nuclear Information System (INIS)

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

  7. Shielding benchmark calculations of selected spent fuel storage cask experiments

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.; Tang, J.S.; Parks, C.V. (Oak Ridge National Lab., TN (United States)); Taniuchi, H. (Kobe Steel Ltd. (Japan))

    1993-01-01

    This paper describes the application of the three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of benchmark spent fuel storage cask measurements performed at the Idaho National Engineering Laboratory. A total of five storage cask problems were analyzed to determine the expected accuracies of computational analyses using well-established Monte Carlo codes. The results presented herein represent the current status of the work. Predicted neutron dose results generally compare very favorably (within 30%) with the measurements for the cask lid, bottom, and along the cask side. Gamma-ray dose rates exhibit differing trends, depending on the measurement location. For lid and bottom doses, as well as side doses near the endfittings, agreement is again within 30%, although several exceptions are seen. However, for gamma doses along the cask side and adjacent to the active fuel, a factor of 2 overprediction is noted. Investigations into the cause of these discrepancies are currently in progress.

  8. Shielding benchmark calculations of selected spent fuel storage cask experiments

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.; Tang, J.S.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Taniuchi, H. [Kobe Steel Ltd. (Japan)

    1993-03-01

    This paper describes the application of the three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of benchmark spent fuel storage cask measurements performed at the Idaho National Engineering Laboratory. A total of five storage cask problems were analyzed to determine the expected accuracies of computational analyses using well-established Monte Carlo codes. The results presented herein represent the current status of the work. Predicted neutron dose results generally compare very favorably (within 30%) with the measurements for the cask lid, bottom, and along the cask side. Gamma-ray dose rates exhibit differing trends, depending on the measurement location. For lid and bottom doses, as well as side doses near the endfittings, agreement is again within 30%, although several exceptions are seen. However, for gamma doses along the cask side and adjacent to the active fuel, a factor of 2 overprediction is noted. Investigations into the cause of these discrepancies are currently in progress.

  9. Estimated risk contribution for dry spent fuel storage cask

    International Nuclear Information System (INIS)

    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)

  10. Shipping and storage cask data for spent nuclear fuel

    International Nuclear Information System (INIS)

    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

  11. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Application of FELTRAN to NEACRP TN12 shipping cask benchmark. [Shielding of spent nuclear fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Evans, A.M.; Winstanley, D.D.; Watmough, M.H. (British Nuclear Fuels plc, Risley (United Kingdom)); Gerber, R. (Salford Univ. (United Kingdom). Dept. of Pure and Applied Physics)

    1991-01-01

    British Nuclear Fuels plc and Imperial College have collaborated in developing the finite element neutron shielding design code FELTRAN to near production code status. FELTRAN solves the even parity form of the Boltzmann equation using a functional approach. The solution is found in one or two spatial dimensions using various orders of finite elements to specify the problem geometry. The angular dependence of the even parity flux is expressed using spherical harmonics. FELTRAN has been interfaced to ANISN formatted nuclear data libraries such as CASK and BUGLE. Anisotropic scattering may be specified to any order. Methods have been incorporated within the code to analyse systems with voids. FELTRAN is currently undergoing further development. The purpose of this paper is to consider the application of FELTRAN to a practical shield design problem. The OECD have adopted a benchmark experiment to measure the neutron and gamma ray radiation dose rates around a spent fuel transport flask. As part of an international collaboration the physical details of the flask design and contents have been provided to the nuclear industry. The objective is to perform an international comparison of the methods used in the analysis of cask shielding. BNFL is one of the companies involved, using the well established codes RANKERN and MCBEND. The FELTRAN calculations are performed using the same source and geometry data and equivalent angular flux expansions as for these two codes. FELTRAN is then compared with experimental and calculated results. (author).

  14. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

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

  15. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Conditions for spent fuel storage cask reapproval. 72.240... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.240 Conditions for spent fuel storage cask...

  16. 75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-07

    ... REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS HD... Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN), NUHOMS HD System listing within the ``List of Approved Spent Fuel Storage Casks''...

  17. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Science.gov (United States)

    2010-01-01

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

  18. 75 FR 33736 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

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

  1. Influence of local regulations on TN dual purpose BWR casks

    International Nuclear Information System (INIS)

    Transnucleaire (Paris, France) and Transnuclear, Inc. (Hawthorne, New York, United Sates) have both developed Pressurized Water Reactor (PWR) spent fuel casks for storage and transport purposes. The products are supplied in Europe by Transnucleaire and in the United States by Transnuclear, Inc. Now the TN Group is working on a design for Boiling Water Reactor (BWR) spent fuel assemblies: the TN 52 L cask is designed for transport and storage is Switzerland, the TN 68 cask is designed for transport and storage in the United States. For storage purpose, national regulatory requirements have to be met: each country has specific demands and criteria. As a consequence, differences between the TN 52 L design and the TN 68 design for rather similar contents appear in several fields: the design work, the licensing process, the manufacturing and the operational life. (author)

  2. High temperature performance limit of containment system of transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Osamu; Saegusa, Toshiari [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Abiko Research Lab.

    1998-03-01

    The containment performance of a containment system using elastomer gaskets for transport casks under a high temperature and high pressure was clarified. Major results are as follows; (1) The deformation characteristics of the gaskets were represented by the compressive permanent strain rate (Dp). The temperature and time dependence was shown by Larson-Miller Parameter (LMP). (2) Generally, the high temperature performance limit is obtained by a value of LMP when the Dp value reaches 80%. However, the gaskets (FKM, VMQ, EPDM) used for real transport casks were not damaged and the containment performance was not deteriorated as a conservative condition. (3) Assuming that the service period of the gaskets for transport casks is 3 months or 1 year, the high temperature performance limit of the gasket made of fluorine rubber (FKM) is 202degC or 182degC, respectively, which includes safety margin. (author)

  3. High temperature performance limit of containment system of transport cask

    International Nuclear Information System (INIS)

    The containment performance of a containment system using elastomer gaskets for transport casks under a high temperature and high pressure was clarified. Major results are as follows; (1) The deformation characteristics of the gaskets were represented by the compressive permanent strain rate (Dp). The temperature and time dependence was shown by Larson-Miller Parameter (LMP). (2) Generally, the high temperature performance limit is obtained by a value of LMP when the Dp value reaches 80%. However, the gaskets (FKM, VMQ, EPDM) used for real transport casks were not damaged and the containment performance was not deteriorated as a conservative condition. (3) Assuming that the service period of the gaskets for transport casks is 3 months or 1 year, the high temperature performance limit of the gasket made of fluorine rubber (FKM) is 202degC or 182degC, respectively, which includes safety margin. (author)

  4. Plutonium detection in casks of compactable solid waste

    International Nuclear Information System (INIS)

    This report describes a method for determining plutonium in casks of compactable solid waste; it can be applied to amounts of plutonium varying from 2 to 200 grams. The principle of the method is the counting of the 380 keV γ photons from the plutonium 239; a correction is required if both zirconium 95 and niobium 95 are present in the cask. The maximum amount of zirconium 95 + niobium 95 which can be tolerated is 5 microcuries per gram of plutonium, and 300 microcuries per cask. Under the best conditions the accuracy of the measurement appears to be of the order of ±30 per cent, but experience has shown that the method is very useful as a guide to the recovery of the plutonium in the waste. In effect, for a batch of fifty measurements, the difference between the plutonium measured by this method and the plutonium recovered from the waste was equal to 10 per cent. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-11-07

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

  7. Thermal Evaluation of a KRI-BGM Shipping Cask

    International Nuclear Information System (INIS)

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

  8. Optimization of cask capacity for long term spent fuel storage

    International Nuclear Information System (INIS)

    Full text: Long term storage of spent fuel is a priority topic within the Member States of the IAEA. Long term spent fuel storage was previously addressed in an IAEA Co-ordinated Research Project /1/, which recognized the growing challenge of extending the life of storage facilities. Dry cask storage of spent fuel is playing a steadily increasing role in this regard. Storage practices should comply with IAEA safety requirements 'International Basic Safety Standards for Protection Against Ionizing Radiation and for the Safety of Radiation Sources' /2/, including maintaining doses as low as reasonably (taking economic/social/etc aspects. into account) achievable [i.e., the ALARA principle]. Within the framework of the IAEA Subprogramme of Spent Fuel Management, a new project was conceived, focusing on issues associated with the optimisation of cask/container loading (capacity) with respect to long term storage and the related integrity of fuel, see IAEA /3. Optimization is a part of the design process in which the combination of application objectives, regulatory limits and design margins are innovatively addressed and judiciously balanced in the final design. A primary result of a successful design optimization is a cask of superior assembly and burnup/age capacity that minimizes the total number of required cask loadings. An equally important and parallel benefit is that this process also results in reduced radiation exposure, thereby contributing significantly to maintaining doses as low as reasonably achievable (ALARA objectives). In this sense, both cask designers and regulators have the common ultimate goal of improving cask performance, and thus facilitating optimization. An initial Consultants Meeting held in November 2002 identified and discussed principal issues regarding the optimization of cask/container assembly capacity and burnup/age capability in the design of systems for long term spent fuel storage and the related integrity of fuel. Working

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

    International Nuclear Information System (INIS)

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

  10. Quality assurance in a cask fleet parts control system

    International Nuclear Information System (INIS)

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

  11. Quality assurance in a cask fleet parts control system

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Performance testing and analyses of the VSC-17 ventilated concrete cask

    International Nuclear Information System (INIS)

    This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power ampersand Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained

  14. 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)

    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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Asymmetric temperature profiles in transport and storage casks for radioactive materials

    International Nuclear Information System (INIS)

    Transport and storage casks for spent fuel elements or vitrified radioactive waste are exposed to radioactive radiation and additional thermal load due to the radioactive inventory. A reliable heat removal is required in order to avoid material degradation of the shielding and the cask. The calculation procedures of maximum temperatures in the casks need structural modeling of the cask inventory and the environmental conditions with respect to the heat removal. The authors show that simplified models with homogeneous heat load distributions underestimate the real conditions. Detailed models of the cask internals and the air circulation around the cask under a protection cover and sun irradiation have to be taken into account. The calculations methods have to be adapted to the safety relevant conditions of each cask type.

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

    International Nuclear Information System (INIS)

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

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

    CERN Document Server

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Impact analysis of spent fuel dry casks under accidental drop scenarios

    International Nuclear Information System (INIS)

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister (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)

  6. Experience Gained From Skoda VPVR/M Casks Use For RRRFR Program

    International Nuclear Information System (INIS)

    Full text: The aim of the paper is to present the Nuclear Research Institute Rez plc, Czech Republic (NRI) cask owner experience gained from the use of unique high capacity transport and storage SKODA VPVR/M cask technology. Cask is licensed for nearly all types of Russian origin research reactor fuel (package design license, transport permission). NRI is participating in the Russian Research Reactor Fuel Return (RRRFR) program incorporated in the Russian Federation - United States common activity Global Threat Reduction Initiative (GTRI) supported by IAEA. Within the scope of this project, the high enriched uranium (HEU) and low enriched uranium (LEU) spent nuclear fuel (SNF) from NRI was returned back to RF Mayak facility in November 2007 using 16 of these casks (total 549 fuel assemblies and hermetic canisters with SNF, only one combined road and rail secure shipment needed in total). Now NRI is supporting the SNF shipments from further countries research reactors - cask Users (in 2008 from Hungary and Bulgaria already completed, in 2009 from Ukraine and Poland in progress, in the following years from Serbia, Byelorussia in preparation, and next coming). The NRI role and activity in the program is based on performing the transport packaging system inspections and maintenance, assuring the transportation of the empty SKODA VPVR/M casks in special ISO containers and cask handling auxiliary equipment / accessories to the user, providing instruction and recommendations for the research reactor facility modifications, foreign country research reactor staff education and training to handle with the casks, reviewing the user cask handling operational procedures, performing special technical support (cask drying and He-leak testing), supervising the SNF loading into the casks, and loaded casks mounting into special ISO containers. These activities are covered by a long term Cask Custodian Contract between the NRI and US DOE valid until the year 2014 and Service

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Interim Dry Storage of Spent Fuel in Casks

    International Nuclear Information System (INIS)

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

  10. Structural analysis of closure bolts for shipping casks

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

  15. Design of casks: incorporating operational feedback from maintenance

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  16. Design of casks: incorporating operational feedback from maintenance

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  18. Heat transfer investigations within dry spent fuel casks

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  2. Dry cask spent fuel storage at JAPC Tokai No.2 Power Station

    International Nuclear Information System (INIS)

    The Dry cask spent fuel storage project at Tokai No.2 power station started with a geological examination of the facility site, and a design of the cask and storage facility in the mid-1990s. Considering economical efficiency and suitability for site conditions, a cask with large capacity is designed. The cask can accommodate 61 BWR fuel assemblies and is used for on-site storage only. 24 casks can be stored in the storage facility, which consists of a concrete facility building, an overhead crane and some monitoring systems. The foundation of the facility building is supported on bedrock with steel piles. Air inlets and outlets for passive natural circulation cooling are installed on the walls. The construction of the facility building and the fabrication of 7 casks began in 1999, and were completed in 2001. 8 casks for the second stage were fabricated in 2004. 2 casks for the third stage are fabricating. And 4 casks for the forth stage are in process of design. The first loaded 4 casks have been stored safely in the facility for three years since December 2001, and followed another 9 dry casks as of the end of 2007. In addition to the above spent fuel storage management at reactor site, a spent fuel storage away from reactor (AFR) is projected to start operation by 2010. We think our experiences on Tokai No.2 power station will be able to apply to the AFR storage project, such as the design of the cask and the facility. Outline of the Tokai No.2 project, experiences on the fuel loading and cask storage conditions including the monitoring data will be reported on this paper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Margeanu, C.A.; Iorgulis, C. [Reactor Physics, Nuclear Fuel Performances and Nuclear Safety Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Ciocanescu, M. [Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Prava, M. [Design Department, Institute for Nuclear Research Pitesti, Campului Str, no.1, 115400 Mioveni (Romania); Margeanu, S. [Radiation Protection Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania)

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

  5. Analysis technology on the thick plate free drop impact of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    The package used to transport radioactive materials, which is called by cask, must maintain the structural integrity for the requirements of hypothetical accident conditions, 9m free drop of the thick plate impact. These requirements for the cask design should be verified through test or finite element analysis to confirm the regulatory guide. In this paper, three dimensional impact analysis using ABAQUS/Explicit code under 9m free drop of the thick plate impact condition for the KSC-4 cask is performed. As the results, maximum stress intensity on each part of the cask and deformation shape of the cask is calculated and the structural intensity of the cask is evaluated by NRC Regulatory Guides. (orig.)

  6. A study on impact problems of a spent fuel shipping cask

    International Nuclear Information System (INIS)

    The design method is described of preventing a spent fuel shipping cask from being damaged in case of drop accidents. This subject was experimentally and analytically studied toward development of a new type of cask. The cask consists of shock absorbers, a cask shell structured in three layers of steel-lead-steel, internal structures and cooling water. A toroidal shell type shock absorber was employed in this case, and its characteristics were presented. The dynamic response of each structural element was also investigated. Tests were carried out using the drop tower which was constructed in conformity with IAEA regulation. A one-dimensional, lumped-mass, nonlinear spring system was used for dynamic anayses. A 1/4 scale model of the 100 ton cask which was designed on the basis of the above results was put to the test, and as a result, the validity of the design method and the structural integrity of the new cask were confirmed. (author)

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-12

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

  10. Comparison of structural integrity of casks for spent-fuel transportation

    International Nuclear Information System (INIS)

    This paper presents the results of a series of finite element analyses comparing the structural integrity of two shipping casks for transportation of high-level waste (HLW). The objective of this project is to assess the advisability of utilizing ductile iron (DI) for type-B transport cask construction by investigating its structural response under severe loading conditions. This response is compared to that of a stainless steel (SS) cask under comparable loading conditions

  11. Dry cask spent fuel storage at JAPC Tokai No.2 power station

    International Nuclear Information System (INIS)

    The Dry cask spent fuel storage project at Tokai No.2 power station started with a geological examination of the facility site, and a design of the cask and storage facility in the mid-1990s. Considering economical efficiency and suitability for site conditions, a cask with large capacity is designed. The cask can accommodate 61 BWR fuel assemblies and is used for on-site storage only. 24 casks can be stored in the storage facility, which consists of a concrete facility building, an overhead crane and some monitoring systems. The foundation of the facility building is supported on bedrock with steel piles. Air inlets and outlets for passive natural circulation cooling are installed on the walls. The construction of the facility building and the fabrication of 7 casks began in 1999, and were completed in 2001. 8 casks for the second stage were fabricated in 2004. And 6 casks for the third stage are in process of design. The first loaded 4 casks have been stored safely in the facility for three years since December 2001, and followed another 6 dry casks as of the end of 2004. In addition to the above spent fuel storage management at reactor site, a spent fuel storage away from reactor (AFR) is projected to start operation by 2010. We think our experiences on Tokai No.2 power station will be able to apply to the AFR storage project, such as the design of the cask and the facility. Outline of the Tokai No.2 project, experiences on the fuel loading and cask storage conditions including the monitoring data will be reported on this paper. (author)

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

    International Nuclear Information System (INIS)

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

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

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

    Science.gov (United States)

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

    2014-02-01

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

  16. Bonner sphere neutron spectrometry at spent fuel casks

    CERN Document Server

    Rimpler, A

    2002-01-01

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

  17. Fuel Element Transfer Cask Modelling Using MCNP Technique

    Science.gov (United States)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

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

  18. Safety engineering achievements in handling casks at La Hague

    International Nuclear Information System (INIS)

    Interest is focused on safety aspects of some new trends in commercial fuel reprocessing plants at La Hague. The first is the dry cask unloading, unique in size, avoiding several meters height handling and associated risks. Moreover, improvements were introduced about contamination retention, effluent decrease and contact work time, resulting in lower operators' doses. Extensive use of standard equipment, whose replacement using a special cask is foreseen as a common maintenance operation, is another major improvement for an industrial process, increasing plant availability with low personal doses compared to contact intervention. Associated crane use led to systematic studies of falling accidents and, where necessary, improved reliability crane design. It has been shown that the design and quality of corresponding elements is such that the prevention of risk is sufficient to reach a high level of safety. 1 fig

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

    International Nuclear Information System (INIS)

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

  20. Stress analysis of closure bolts for shipping casks

    International Nuclear Information System (INIS)

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

  1. Storage cask drop test on reinforced concrete slab

    International Nuclear Information System (INIS)

    The test results obtained may be summarized as follows: (1) The strain and acceleration during oblique dropping are sufficiently small compared with those during vertical and horizontal dropping. The strain and acceleration due to the secondary collision after dropping are also sufficiently small as compared with those due to the primary collision. For evaluation of integrity against vertical and horizontal orientation, therefore, it can be considered that dropping in the oblique orientation will pose no problem in making such evaluation. (2) The structural integrity of the cask against its dropping at the normal operating height and up to the maximum lifting height which is determined by the construction of storage facilities was verified. (3) Since the estimated critical drop height is sufficiently heigh as compared with the above-mentioned drop height, it was verified that the cask had a sufficient margin against a falling accident during operation. (J.P.N.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Safety evaluation of dry-cask storage facility for spent fuel during earthquake

    International Nuclear Information System (INIS)

    Design criteria of storage facilities were established considering the special circumstances of Japan, such as limited site area and strong earthquakes. Therefore, it is necessary to confirm the integrity in the rare case of the collapse of storage building and gantry crane, the cooling performance of cask buried in concrete rubbish, and the resistance for overture of dry storage cask during earthquake. This report evaluated the security for impact load of falling body such as concrete wall or gantry crane, the stability of the dry storage cask during earthquake, and the cooling performance of cask. (author)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Kawaguchi, Shohei; Shirai, Koji; Kanazawa, Kenji

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

  8. Human factors engineering applications to the cask design activities of the Civilian Radioactive Waste Management Program

    International Nuclear Information System (INIS)

    The use of human factors engineering (HFE) in the design and use of spent fuel casks being developed for the Department of Energy's Civilian Radioactive Waste Management Program is addressed. The safety functions of cask systems are presented as background for HFE considerations. Because spent fuel casks are passive safety devices they could be subject to latent system failures due to human error. It is concluded that HFE should focus on operations and verifications tests, but should begin, to the extent possible, at the beginning of cask design. Use of HFE during design could serve to eliminate or preclude opportunity for human error

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

    International Nuclear Information System (INIS)

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

  10. Sampled control of vibration in suspended cask by using vibration manipulation functions

    International Nuclear Information System (INIS)

    Safe and reliable operation is most important for decommissioning the Fukushima 1 nuclear power plant. Especially it requires for transferring spent nuclear fuels from fuel pool to storage cask. Since the heavy cask will be suspended during the transferring operation, there is a risk of dropping it in case of the strike of large earthquakes. In this study, we introduce analytical functions to suppress residual vibration of a suspended cask by using vibration manipulation function. Hence the oscillation of the cask can be feedforward or sampled-data controlled by moving a trolley with analog actuator, the possible risk could be reduced. (author)

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

    International Nuclear Information System (INIS)

    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 % ΔK subcriticality margin is always met

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

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

    International Nuclear Information System (INIS)

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

  16. Experience from transport casks management in the Swedish AFR-facility CLAB

    International Nuclear Information System (INIS)

    From the 12 Swedish reactors, all located at the coast, about 250 tonnes of spent fuel are transported to the central intermediate storage facility, CLAB, yearly. The fuel assemblies are transported in TN 17/2 type B(U) casks. The sea transport system consists of the special designed ship M/S SIGYN, 4 terminal vehicles, 12 load carriers, 10 TN 17/2 casks and 2 core component casks TN 17-CC. 26 steel containers, IP-2, for low and intermediate wastes are also part of the system. During a normal year, about 95 casks are transported to and handled at CLAB. Following the ''Green book'' manual, every month 1 cask is going through a preventing maintenance and inspection programme at the maintenance workshop in CLAB. The central intermediate storage facility for spent nuclear fuel, CLAB, was taken into active operation in July 1985. Until December 1988, 362 transport casks have been handled at CLAB and 847 tonnes of spent fuel were stored in the water pools. The CLAB facility is designed in such a way that the possibility for surface contamination of the casks is very small. If this should happen there are different systems and facilities for decontamination. So far only 1 cask has been contaminated during the unloading operation in CLAB. The successful management and the experience of handling the transport casks in CLAB, will be described in the paper. (author). 6 figs

  17. Safety margins of spent fuel transport and storage casks considering aircraft crash impacts

    International Nuclear Information System (INIS)

    The safety of spent fuel transport casks in severe accident conditions is always a matter of concern. This paper surveys German missile impact tests that have been carried out in the past to demonstrate that German cask designs for transport and interim storage are safe even under conditions of an aircraft crash impact. A fire test with a cask beside an exploding propane vessel and temperature calculations concerning prolonged fires also show that the casks have reasonably good safety margins in thermal accidents beyond regulatory fire test conditions. (author)

  18. Numerical fracture analysis for the structural design of CASTOR casks

    International Nuclear Information System (INIS)

    The numerical implementation of the dynamic J-Integral is presented as one method to compute the dynamic stress intensity factor (DSIF). The applicability of the computational method is demonstrated by a finite element simulation of a free drop test of a ductile cast iron CASTOR cask with a pre-crack. The results of the simulation are contrasted with the data from the real experiment. (author)

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

    International Nuclear Information System (INIS)

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

  20. Structural evaluation of spent fuel dry storage cask

    International Nuclear Information System (INIS)

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

  1. CLASSIFICATION OF THE MGR CARRIER/CASK TRANSPORT SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) carrier/cask transport system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  2. Cask operation and maintenance for spent fuel storage

    International Nuclear Information System (INIS)

    Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The development of the Source Term Analyses for Containment Evaluations (STACE) methodology provides a unique means for estimating the probability of cladding breach within transport casks, quantifying the amount of radioactive material released into the cask interior, and calculating the releasable radionuclide concentrations and corresponding maximum permissible leakage rates. Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source team has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volitile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking in which experimental validation is planned. Finally, the ANSI N14.5 recommendation that 3% and 100% of the fuel rods fail during normal and hypothetical accident conditions of transport, respectively, has been show to be overly conservative by several degrees of magnitude for these example analyses. Furthermore, the maximum permissible leakage rates for this example assembly under normal and hypothetical accident conditions are significanly higher that the leaktight requirements. By relaxing the maximum permissible leakage rates, the source term methodology is expected to significantly improvecask economics and safety

  5. Cask system maintenance in the Federal Waste Management System

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-23

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 (DUO2, Al2O3, Gd2O3, 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 (DUO2) 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 (Al2O3) 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 DUO2 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 as a function of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

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

  13. 76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-06-08

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

  14. 76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

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

  15. 78 FR 22411 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections

    Science.gov (United States)

    2013-04-16

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. 76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... COMMISSION 10 CFR Part 72 RIN 3150--AI89 List of Approved Spent Fuel Storage Casks: NUHOMS HD System Revision... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN) NUHOMS HD System listing within the ``List of Approved Spent Fuel...

  19. 78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

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

  20. 76 FR 2243 - List of Approved Spent Fuel Storage Casks: NUHOMS ® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... 3150-AI89 List of Approved Spent Fuel Storage Casks: NUHOMS HD System Revision 1 AGENCY: Nuclear... Commission) is amending its spent fuel storage regulations by revising the Transnuclear, Inc. (TN) NUHOMS HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1...

  1. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Science.gov (United States)

    2011-11-14

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

  2. 75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

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

  3. 75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

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

  4. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Specific requirements for spent fuel storage cask approval... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.236 Specific requirements...

  5. 76 FR 70374 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising... 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL...

  6. 77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2012-01-27

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

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

    International Nuclear Information System (INIS)

    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

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

    Directory of Open Access Journals (Sweden)

    Justin eSlawson

    2014-11-01

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

  9. The safety of transport operations and transport casks for LWR and VVER spent fuel

    International Nuclear Information System (INIS)

    The title topics are discussed, covering the following items: safety as a basic requirement for customers and operators, regulations (which should be stringent following IAEA recommendations), Quality Assurance (which is compulsory following IAEA documents), wide transport experiences, the TN 12 spent fuel shipping cask, and the TN 120 transport/storage cask for WWER-440 spent fuel assemblies. (P.A.)

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

    Science.gov (United States)

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

    2016-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S

    2004-03-18

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

  12. HI-storm dry storage cask tip-over event structural response

    International Nuclear Information System (INIS)

    Current regulations in the United States (10CFR Part 72) allow the power reactor spent fuel and other radioactive materials associated with the spent fuel to be stored at an independent spent fuel storage installation, using a free-standing dry cask storage system, approved by the U. S. Nuclear Regulatory Commission. Even though a cask is designed to preclude tip-over during a design basis earthquake event, structural integrity of the cask is required to be evaluated for a non-mechanistic tip-over event. Additionally, a cask may experience a tip-over event at an angular impact velocity greater than during a design basis earthquake event, due to a potential deliberate act of terrorism of a jetliner impact into a cask storage facility. To understand how a cask storage system would perform at angular impact velocities greater than at an impact velocity greater than during an earthquake event, a study was undertaken to examine the behavior of one of the dry cask storage systems (HI-STORM 100) for a tip-over event at various angular impact velocities. Effects of changes in foundation stiffness on the cask responses were also examined. Behavior of the structural integrity of the HI-STORM 100 cask was examined using a finite-element method of analysis in a computer program, ANSYS/LS-DYNA. A detailed model of the foundation and the cask, including the exterior concrete overpack, the multi-purpose canister and the fuel basket with the spent-fuel, was developed for the explicit method of dynamic analysis. The analyses were performed for the cask tip-over impact on a concrete pad foundation at velocities of 1.7 radians/sec to 5.0 radians/sec. Additional analyses were performed for impact velocities of 1.7 radians/sec and 5.0 radians/sec with the foundation stiffness properties changed by ±50 percent. Results of the analyses were evaluated to understand the behavior of the cask, and relationship of the cask response to the impact velocity and the foundation stiffness. This

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

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

    International Nuclear Information System (INIS)

    On 27 April 1999 a fire test was performed with a 45 m3 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  18. Heat transfer tests by slice model of high performance spent fuel shipping cask

    International Nuclear Information System (INIS)

    In consideration of high burn-up plan for LWR fuel, the HP-CASK has been designed to transport more contents than the existing casks. The high performance cask required a high heat transfer performance inside the cask body, because the strong intensity of the neutron source due to the contents required use of a thick resin shielding layer. This led to a design in which internal fins welded at both ends were provided in the resin layer to ensure heat transfer performance, which was verified by means of a slice model heat transfer test. It was shown, as a result, that the heat transfer performance in the cask body had a performance as originally designed. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-07-14

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. The NCS 45 cask family: an updated design replaces an old design. Lessons learned during design, testing and licensing

    International Nuclear Information System (INIS)

    The NCS 45 cask family is intended to replace the cask types R52, TN6/1 and TN6/3. These packagings - country of origin France - were in operation worldwide since mid 1970. In the late nineties prolongations of the certificates of package approval became more and more difficult and time consuming. Finally only special arrangements for restricted contents were issued by the competent French authority which caused considerable problems when validations in other countries were applied for. To guarantee the availability of such a cask in the future for its customers NCS decided to replace the old casks by an updated design, the NCS 45 cask family

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Shielding benchmarks analysis for transport/storage casks

    International Nuclear Information System (INIS)

    The dose rate measurements of the TN 12/2, TN 28 VT and FS 65 has been used to evaluate the calculational procedures of Transnuclaire (TN). The three-dimensional (3D) Monte Carlo code TRIPOLI-3.4 which is used to optimize the shielding of TN casks, is applied to the analysis of a series of benchmarks. In the same cases the one-dimensional (1D) Sn code SN1D and the point kernel code MERCURE-V (3D) used for the more simplified calculations, are checked by the comparison with the measurements. The multi-group approximation used by the above codes, in order to reduces nuclear data, introduces errors due to the neutron cross-sections resonance treatment and the repartition of the gamma-ray spectrum (discrete) into an energy group structure. For a cask consisting of an iron shell of 250 mm of thickness, neutron dose rates can been underestimated of 50% if the resonances of the iron cross sections for high energy (above 1 MeV) are not taken into account. Also, depending on the energy group structure, gamma-ray dose rates can be over-estimated or under-estimated by the repartition of the gamma rays. The comparisons between measured and calculated dose rates are closer than 20% for the Monte Carlo calculations, 50% for the Sn calculations (1D) and a factor of 2 for the point kernel calculations. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The design of the new transport and storage cask type CASTORR 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Halstead, R. J.; Dilger, F.

    2003-02-25

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

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

    International Nuclear Information System (INIS)

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

  11. Barge shipment and reactor handling of a castor V/21 cask

    International Nuclear Information System (INIS)

    The results of this study consist of a complete handling time/dose assessment for barge transport and reactor loading of a Castor V/21 storage cask. Observations are based on the barge transport and spent-fuel loading of storage casks at the Surry, Virginia, nuclear power plant during 1987. The minimum time required to perform all storage cask-handling activities from ship off-loading through placement of the loaded cask in at-reactor storage was 43.8 h. The addition of delays (due to backshifts not worked, etc.) resulted in a total turnaround time for the operation of ∼6 days (24 h/day). Total labor requirement was 136 person-hours. Occupational dose for these activities totaled 416 person-mrem of exposure was due to background dose, representing ∼40% of total dose. The highest dose-producing activity consisted of those steps involved with draining the loaded storage cask (i.e., installing drain pipe and pumping water from cask). This activity resulted in 75 person-mrem of exposure. Lid installation and vacuum drying of the cavity resulted in 56 person-mrem of exposure. The actual loading of spent-fuel assemblies into the storage cask was the third highest dose-contributing activity, resulting in 38 person-mrem of exposure

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

    International Nuclear Information System (INIS)

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

  13. Optimization of cask for transport of radioactive material under impact loading

    International Nuclear Information System (INIS)

    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®. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Experience in complying with quality assurance requirements for cask lifting devices

    International Nuclear Information System (INIS)

    The Nuclear Assurance Corporation (NAC) owns and operates four NAC-1 truck casks. These casks are used to ship spent reactor fuel assemblies and radioactive reactor-core components. The casks have been loaded or unloaded at a total of fifteen nuclear facilities in the United States. In addition, NAC has used another large, overweight-truck cask to ship radioactive reactor core components from a reactor to a waste burial site. There are many individual differences in the cask handling facilities at each of the reactor stations, nuclear research facilities and the storage and burial sites serviced. Various types of auxiliary lifting and handling devices for on-site cask operations have been required. The quality assurance requirements for the equipment used in interfacing casks with nuclear power plant facilities have become more stringent. This paper presents details on the type of special equipment being employed, the quality assurance requirements that are imposed, and the quality assurance audits that are being performed. The paper presents NAC's experiences in the development and procurement of a variety of cask-facility equipment and the implementation of quality assurance procedures for the design, manufacture, acceptance and in-service inspection and test of the equipment. Also, experiences in working with customers' engineering and quality assurance organizations are discussed with specific attention given to the establishment of interface equipment requirements and the documentation that must be developed. The paper discusses a number of factors that must be considered in the development of design criteria for cask lifting devices. In addition to the criteria that are important to the functional safety of the equipment, other considerations important to the equipment utilization and effectiveness are presented

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Burnup credit application in criticality analysis of storage casks with spent RBMK-1500 nuclear fuel

    International Nuclear Information System (INIS)

    Nuclear criticality safety analysis of two types of the casks CASTOR RBMK-1500 and CONSTOR RBMK-1500 was performed using the SCALE 4.3 computer code system. These casks are planned for an interim dry storage of spent nuclear fuel at Ignalina nuclear power plant. Effective neutron multiplication factor keff was calculated for different density of the water inside the casks for unfavorable operational cases and for assumed hypothetical accident conditions when fuel in the system is fresh and fuel is depleted (i.e. burnup credit taken into account). Results show that for all cases effective neutron multiplication factor keff is less then allowable value 0.95. (author)

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

    International Nuclear Information System (INIS)

    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

  4. Certification challenges in the development of an innovative high payload capacity spent fuel transportation cask

    International Nuclear Information System (INIS)

    The design approach and certification strategy used in the development of an innovative transportation cask for legal weight truck shipments of spent nuclear fuel is presented. The proposed approach represents a significant departure from conventional cask designs in that it uses titanium alloy, a material with a high strength-to-weight ratio which has no precedent in transportation cask certification. The significant increase in payload obtainable with the proposed approach, and the associated benefits such as reduced life cycle costs, lower personnel exposure, and lower transportation accident risks are discussed. 8 refs., 3 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-01

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

  9. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-26

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    In Japan, utilities are preparing to initiate an independent interim storage facility (ISF) for spent fuel at Mutusi-city in Aomori prefecture in 2010. In the ISF, dual purpose metal casks which are used for both transportation and storage will be adopted, because no direct handling of spent fuel is necessary at the ISF, thereby reducing risks. The metal cask will be handled without impact limiters in the ISF. Therefore, supposing a hypothesis cask drop accident without the limiter, cask drop tests using an actual size simulated cask were analyzed and the leak characteristics from the flange with the metal gasket were investigated. The tests were conducted without the limiter, and the conditions were a horizontal drop and rotational impact with the supporting point at a trunnion. Before the calculation of this cask drop event, based on examination of results obtained from small scale tests for seal performance of flange with aged metal gasket, a correlation curve between total sliding movement of lid and leak rate was obtained. The relation between the total sliding movement of the lid and the leak rate obtained from the cask dropping tests without the impact limiter was compared with the correlation. Considering the leak rate increase due to aging of the gasket which is assumed to be ranging from 100 to 1000, the result from the cask drop tests agreed to the correlation with a 95% confidence level. Then, a general non-linear dynamic simulation computer code, LS-DYNA was used in the calculation of the cask drop tests. In the calculation, a half of the cask, considering axial symmetry, and a concrete floor were modelled. The calculation for the horizontal drop test was initiated just before a trunnion impacts the floor. For the rotational impact test, the calculation was initiated just before the edge of the outer flange impacting the floor. The impacting velocity of the cask was calculated assuming a free drop from the original position for both horizontal drop and

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Development of the GA-4 and GA-9 legal weight truck spent fuel shipping casks

    International Nuclear Information System (INIS)

    General Atomics (GA) is developing two legal-weight-truck spent-fuel shipping casks for transporting commercial reactor spent fuel. The GA-4 pressurized-water-reactor (PWR) and the GA-9 boiling-water-reactor (BWR) casks are stainless steel with non-circular cross sections. Depleted uranium (DU) and boron polypropylene are used for gamma and neutron shielding. Solid pellets of boron carbide contained in a removable stainless steel fuel support structure provide criticality control. The GA-4 Cask utilizes burnup credit to maintain a capacity of four spent fuel assemblies for enrichments greater than 3.0 U-235 wt %. Aluminum honeycomb impact limiters and a dedicated semitrailer contribute to the overall efficiency and safety of the system. Design verification testing of a half-scale model cask will confirm the adequacy of the structural design

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-11-01

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

  2. COBRA-SFS modifications and cask model optimization

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Effect of Loading Pattern on Thermal and Shielding Performance of a Spent Fuel Cask

    International Nuclear Information System (INIS)

    This study analyzes the effect of non.-uniform load patterns on peak fuel cladding temperatures and cask surface dose rates using previously validated analytical methods. The study was performed using a spent fuel storage cask that was designed to hold 24 spent fuel assemblies with a decay heat load of 24 kW. The fuel was selected to have cooling times of 3.5 to 10 years, burnups of 20 to 60 GWd/MTU, and enrichments of 2.4 to 4.8%. Three radial power distributions were considered in the study: uniform loading, hotter assemblies in the center of the cask, and hotter assemblies near the wall of the cask. Seventeen different load patterns were selected. For a given decay heat load in the cask, loading assemblies with higher decay heat output around the outside of the cask results in lower peak fuel cladding temperatures than loading hotter assemblies in the center of the cask. Several of the load patterns resulted in a peak cladding temperature that was lower than for a uniformly loaded cask. Seven source terms were selected to provide the thermal output used in the thermal analysis. A constant power density of 32 MW/MTU was used for all irradiation calculations. Cooling times were selected to provide the decay heat values used in the thermal analysis. Photon dose rates are dominated by the cobalt-60 in the bottom-end fittings, top-end fittings, and plenum and are proportional to fuel burnup. For short cooling times, photon dose rates on the side of the cask are somewhat higher due to short-lived fission products. Cask loadings with high decay heat assemblies near the periphery exhibit increased photon dose rates on the side surface and top and bottom surfaces away from the centerline. Near the centerline, on the top and bottom of the cask, the dose rates are reduced substantially. Neutron dose rates increase exponentially with burnup and are nearly independent of cooling time.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

  9. Maintenance manual for the Beneficial Uses Shipping System cask. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-05-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    J. Bisset

    2005-02-14

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Experimental determination of radiation safety of spent nuclear fuel dry storage casks CASTOR and CONSTOR

    International Nuclear Information System (INIS)

    When Ignalina NPP was built it was planned that spent nuclear fuel (SNF) will be stored at the pools for 3-5 years and after that will be transported to Russia for reprocessing or disposal. But after reestablishment of independence the situation changed totally and an urgent need arose to solve the questions related with interim storage of spent nuclear fuel in Lithuania, because storage pools were almost totally filled. Various possibilities have been analysed and finally it was decided to use dry storage technology for interim storage (up to 50 years) of Ignalina NPP spent nuclear fuel. For this purpose GNB (Germany) duel-purpose casks have been chosen. The part of them are ductile cast iron CASTOR RBMK-1500 casks and the rest part are metal-concrete CONSTOR RBMK-1500 casks. In order to evaluate radiation characteristics of the casks, combined experimental investigations (measurements of the equivalent dose and γ-spectrum on the cask surface at dry storage) and computer modeling (calculations of the equivalent dose rates, activities of nuclides, etc.) were performed. The obtained results show that equivalent dose rate values on the surface of the casks are much less than the design criteria value of 1000 μSv/h. (author)

  1. Seismic response analysis of a free-standing model of spent fuel storage cask

    International Nuclear Information System (INIS)

    The seismic response analyses of a free-standing spent fuel storage cask are performed for an artificial time history acceleration generated on the basis of the US NRC RG1.60 response acceleration spectrum. This paper focuses on the structural stability regarding seismic loads to check the overturning possibility of a storage cask and the slip displacement on the concrete installation bed. A simple structural analysis model for the storage cask is developed to perform the parametric effect analyses regarding the seismic responses. Two parameters considered in the analyses are the magnitude of the seismic load and the interface friction between the cask's bottom surface and the upper surface of the concrete installation bed. The analyses results show that the seismic responses of the storage cask are influenced by a combination of the two parameters and the storage cask also has a large marginal integrity for the maximum overturning angle and the slip distance for the design and beyond design seismic loads. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Optimization strategies for cask design and container loading in long term spent fuel storage

    International Nuclear Information System (INIS)

    As delays are incurred in implementing reprocessing and in planning for geologic repositories, storage of increasing quantities of spent fuel for extended durations is becoming a growing reality. Accordingly, effective management of spent fuel continues to be a priority topic. In response, the IAEA has organized a series of meetings to identify cask loading optimisation issues in preparation for a technical publication on Optimization Strategies for Cask/Container Loading in Long Term Spent Fuel Storage. This publication outlines the optimisation process for cask design, licensing and utilization, describing three principal groups of optimization activities in terms of relevant technical considerations such as criticality, shielding, structural design, operations, maintenance and retrievability. The optimization process for cask design, licensing, and utilization is outlined. The general objectives for the design of storage casks, including storage casks that are intended to be transportable, are summarized. The nature of optimization within the design process is described. The typical regulatory and licensing process is outlined, focusing on the roles of safety regulations, the regulator, and the designer/applicant in the optimization process. Based on the foregoing, a description of the three principal groups of optimization activities is provided. The subsequent chapters of this document then describe the specific optimization activities within these three activity groups, in each of the several design disciplines

  6. Long-term containment performance of storage cask for spent fuel

    International Nuclear Information System (INIS)

    Previous papers [1-4] reported that the performance life of metallic gaskets coated with aluminum or silver will be more than 190 years, respectively, based on an accelerated tests and Larson-Miller's estimation method. This paper describes demonstrative tests on long-term containment of full-scale cask lid models. The cask models were selected from various types of storage casks, taking account of the influential structure such as lid shape, gasket groove, and gasket structure. The tests have continued from 1990 for more than 9 years. In addition, it was noted that the casks experience temperature variation of seasons in the storage building. It will be necessary to confirm any influence of such environmental temperature variation, on the containment performance of the casks. Primary, a cyclic creep characteristics should be investigated. Factors for the cyclic creep will be number of cycles, temperature, velocity of cycle, etc. Taking account of those factors, temperature-cyclic tests were carried out to investigate the effect on the containment of the cask. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  9. Analysis for Spent Nuclear Fuel Multi-Canister Overpack (MCO) Drop into the Cask from the Multi-Canister Overpack - Handling Machine (MHM) with Air Cushion

    International Nuclear Information System (INIS)

    The purpose of this report is to investigate the potential for damage to the MCO during impact from an accidental drop from the MHM into the shipping cask. The MCO is dropped from a height of 8.2 feet above the cask enters the cask concentrically and falls the additional 12.83 feet to the cask bottom. Because of the interface fit between the MCO and the cask and the air entrapment the MCO fall velocity is slowed. The shipping cask is resting on an impact absorber at the time of impact. The energy absorbing properties of the impact absorber are included in this analysis

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    International Nuclear Information System (INIS)

    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 TNTM24 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, TNTM9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TNTM9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TNTM24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TNTM9/4 round trips are performed, and one TNTM24BH is loaded. 5 additional TNTM24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TNTM24BH high capacity dual purpose cask and the TNTM9/4 transport cask characteristics and capabilities will then be detailed. (authors)

  13. Inspection of Fuel Cladding and Metal Gasket in Metallic Dry Cask at Tokai No. 2 Power Station

    International Nuclear Information System (INIS)

    The metallic dry cask storage of spent fuel started in December 2001 at TOKAI No.2 power station. The cask that had served for 7 years was inspected in January 2009. The objective of this inspection is confirmation of fuel cladding and metal gasket integrity. This cask accommodates 8 × 8 zirconium liner type fuel. The gasket applied to this cask consists of aluminum outer lining and Inconel spring. This inspection confirmed that there had been no damage in fuel cladding and metal gasket during the storage for 7 years. (author)

  14. Research on spent fuel storage and transportation in CRIEPI. Part 1. Metal cask and vault storages, and transportation

    International Nuclear Information System (INIS)

    For metal cask storage method, containment safety of the metal gasket was demonstrated using a full-scale metal cask without shock absorbers subjected to drop accidents during handling work at a storage facility. The instantaneous leakage was negligible. Long-term containment of the metal gasket has been demonstrated using full-scale cask lid models under a high and constant temperature for more than 17 years. Taking account of temperature decay in the real cask, the containment for more than 60 years has been evaluated. Hypothetical airplane crash against a cask storage building was studied by analysis and tests. The mechanical impact on the containments of the metal gaskets of the lid structure of the cask was analyzed and demonstrated by tests. Vault storage method may be economical for a large capacity of spent fuel storage. A design concept of the vault storage at a shallow underground was developed and the licensability of the underground's space was studied. Transport cask may deteriorate with respect to its elastomer gaskets as a result of creep deformation. The deformation and reduction of resilience of the gasket was studied by means of an analysis of finite element method. Transport ship of casks on the sea was assumed to shipwreck hypothetically and the casks loose their containment in the sea. Radiation dose under the hypothetical accident was evaluated by means of an analysis using an oceanic circulation model of the sea water. (author)

  15. Conceptual design report for a transportable DUCRETE spent fuel storage cask system

    International Nuclear Information System (INIS)

    A conceptual design has been developed for a spent fuel dry storage cask that employs depleted uranium concrete (DUCRETE) in place of ordinary concrete. DUCRETE, which uses depleted uranium oxide rocks rather than gravel as the concrete's heavy aggregate, is a more efficient overall radiation shield (gamma and neutron) than either steel or ordinary concrete. Thus, it allows the cask weight and size to be substantially reduced. Also, using DUCRETE as shielding avoids, or at least defers, disposal of the depleted uranium as waste. This report focuses on DUCRETE cask transportation issues. The approach studied involves placing the storage cask into a simple steel transportation overpack. Preliminary analyses were performed to demonstrate the transportation system's ability to meet the structural, thermal, and shielding transportation criteria. Conservative manual calculations were performed to demonstrate the adequacy of the DUCRETE transportation overpack with respect to structural requirements. Two-dimensional thermal analyses were performed on the system (the DUCRETE storage cask inside the steel overpack) using the ANSYS thermal analysis code. Two-dimensional shielding analyses were performed on the system with the MCNP code. Effects of the fuel axial burnup profile and solar radiation are considered. The analyses show that the proposed system can meet the transportation structural criteria and can easily meet the transportation shielding criteria. The thermal criteria are not as easy to meet because when the storage cask is placed horizontally in the transportation overpack, the DUCRETE storage cask's ventilation duct becomes an insulating dead air space. The maximum allowable temperature for the DUCRETE, which is not yet known, will be the limiting factor

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

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  17. Radiological and thermal characteristics of CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks for spent nuclear fuel storage at Ignalina nuclear power plant

    International Nuclear Information System (INIS)

    The main characteristics, such as temperatures of the fuel rod cladding and cask surface, dose rates at the surface and at the some distance for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks loaded with spent nuclear fuel are presented. These casks are used for an interim dry storage of spent nuclear fuel at Ignalina Nuclear Power Plant. Numerical modeling (calculation of the equivalent dose rates, activities of nuclides, etc.) and experimental measurements of the equivalent dose and gamma spectrum on the cask surface at the dry storage facility were performed for assessment of radiation characteristics. Temperatures were evaluated using only numerical modeling. Rather good agreement between experimentally determined and calculated dose rates for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks was obtained. Also it was revealed that maximum fuel rod cladding temperature is higher for CONSTOR RBMK-1500 cask, but never exceeds the maximum allowable value. The cask surface temperatures are similar for both cask types. (orig.)

  18. Conceptual Design Report - Cask Loadout System Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform, 105 K West Basin, Project A.5/A.6

    International Nuclear Information System (INIS)

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

  19. 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.

  20. Optimization of radiation protection by optimizing technology of CASTOR-Cask loading

    International Nuclear Information System (INIS)

    Full text: Germany Optimization of Protection is one of the basic principles of the ICRP System of Radiation Protection. Often this principle is misunderstood and people try to achieve minimal doses irrespective of the amount of manpower or money they have to afford to reach this aim. The better way of optimization is to optimize the technology or the practise which is the cause of radiation exposure and at the same time reduce the dose uptake. Three measures have been used for this purpose in the management of spent fuel in Germany in preparation for the dry storage in CASTOR-Casks. The casks have to be loaded with the spent fuel in the pond of the power plant. After the loading the cask has to be dewatered and dried. The remaining humidity has to be checked with respect to a given maximum residual humidity to avoid corrosion during the long-term storage. Initially a measuring device using the dew point mirror method was used. The mirror was often polluted and needed recalibration. This led to a large variety of measuring times, the time period needed for the above mentioned three steps ranged from 55 to 120 hours. Thus the work could not be reliably planned. To solve this problem we now use a pressure-rise method to measure the humidity within the cask. The time needed is now nearly equal and reliable for all cask loadings and considerably lower than using the dew point method. Thereby the dose uptake of the cask handling staff could be reduced to 2.5 man mSv on average in comparison to the former collective dose of 4 to 5 man mSv. A second step for reducing the dose of the staff is the introduction of remotely controlled valves for the drying process, the humidity measurement and the subsequent filling with Helium. The valves are located at the lid of the cask where a remarkable dose rate could be. The equipment for the remote valve handling has been successfully tested. In the same line is a third measure: to record the process data by computer. The supervising

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  2. A CFD analysis of thermal behaviour of transportation cask under fire test conditions

    International Nuclear Information System (INIS)

    Highlights: → Melting and natural convection of lead in cask has been studied using CFD for the first time. → The role of turbulent natural convection on melting was pronounced. → The study establishes the importance of natural convection for accurate thermal design of cask. - Abstract: Thermal design of transportation cask for shipping radioactive waste needs strict compliance with the guidelines of the regulatory bodies. Lead shielding is usually provided in these casks for arresting gamma rays and reducing hazardous emissions to the environment below permissible limits. During transportation, accidental fire may break out and cause melting of lead for a prescribed duration. The present analysis reports, for the first time, a comprehensive CFD analysis of the thermal behaviour of melting of lead under high Rayleigh number convection during the fire test. The study reveals a substantial influence of natural convection on the thermal state and melting behaviour of lead which may have a great bearing on the safety and security of public during transportation of cask.

  3. A study on the free drop impact of a cask using commercial FEA codes

    International Nuclear Information System (INIS)

    The package used to transport radioactive materials, which is called a cask, must be designed to keep its contents safe under normal and hypothetical accident conditions. The design requirements of the cask are verified by test or finite element analysis (FEA). Comparing evaluation procedures for the safety of a new cask, the cost of FEA is generally much less than that test. Therefore, FEA is mainly used to verify safety of a cask under the considered conditions. However, one commercial FEA code may show different results from another FEA code for the same problem due to the modeler's several assumptions for simplifying actual states into the FE model and due to modeling technique. Materials of the components of a cask display elastic-plastic or elastic-perfectly plastic behavior under the considered conditions in which large deformation, impact and contact mechanism are included. The behavior is simulated with difficulty and may have different results depending on FEA codes. In this paper, finite element analysis is carried out for the 9-m free drop and the puncture condition under the hypothetical accident condition by using LS-DYNA3D and ABAQUS/Explicit. Energy and effective stress on each component are presented and compared between the two FEA codes, where the effective stress designates the maximum von Mises stress on inner and outer shells

  4. Thermal analysis of spent fuel storage cask using the fluent code

    International Nuclear Information System (INIS)

    Thermal analysis for spent fuel storage cask loaded with 24 spent PWR fuel assemblies has been carried out using the Fluent code to verify the reliability of analysis method and procedure. And the temperature distribution for storage cask loaded with 24 metalized fuels equivalent to 96 PWR fuels has been also calculated. It is found that the storage volume of PWR assembly is reduced to a quarter and the heat load is reduced to a half by the preferential elimination of Sr-90 and Cs-137 through the metalization process of spent PWR fuel. Total decay heat from 24 spent PWR fuels and 24 metalized spent fuels are 28 kW and 54 kW, respectively. The calculated temperatures for 24 spent PWR fuels were compared with the proven data presented from the safety analysis report of spent fuel storage cask. It has good agreement between the two results, and it is also found that the feasibility of the analysis method and procedure has been confirmed by the results to estimate the temperature for the spent fuel storage cask. The maximum fuel temperature for 24 metalized spent fuel assemblies inside the cask is calculated at 617 .deg. C

  5. Integrated cask storage systems for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Since 1979 Tennessee Valley Authority TVA has participated in conceptual design studies of dry storage vaults, silos, casks, ad dry wells, and, with DOE and others, has undertaken limited demonstrations of rod consolidation and cask dry storage at TVA's Browns Ferry Nuclear Plant in Alabama. TVA believes the integrated storage cask concept is worthy of consideration as an alternative for spent fuel management. Placing spent fuel in a secure passive storage mode at an early date and avoiding unnecessary handling and repackaging reduces the potential for occupational and public radiological exposure. Therefore the notion of a universal cask used for storage, shipment, and disposal is appealing from a safety, environmental, and public perception standpoint. The universal cask can also serve as a dispersed monitored retrievable storage (MRS), thus eliminating the need for redundant facilities, and it does not foreclose future options. It also appears that this concept would simplify repository design, ease retrievability, and provide greater flexibility in repository siting. 2 figures, 2 tables

  6. 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.

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

    International Nuclear Information System (INIS)

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

  8. First experience in international air transportation of RR SFA in Russian-made TUK-19 casks

    International Nuclear Information System (INIS)

    Traditionally, spent fuel assemblies (SFA) have been transported across the Russian Federation by rail in special railcars. New conditions required SFA shipments by other conveyance, i.e. road, sea and even air transport. The air shipment of the VVR-S research reactor SNF in TUK-19 casks from Magurele, Romania in June 2009 was the first experience after new Russian and international regulations for the safe transport of radioactive material came into effect. The preparatory stage of the shipment focused on the issues associated with radiation and nuclear safety both during the loading and transport operations. The project covered development of a technology and equipment for SFA loading into TUK-19 casks and that for the air shipment. The SFAs were loaded into the TUK-19 casks with a specially designed transfer cask, and the SFA-containing packages were transported in specialized freight 20-foot ISO-containers. The safety of the loading and transport operations was ensured both by reliable engineering solutions, and selected conveyances and routes. The paper shows that the loading and the air shipment of the Romanian SFAs in TUK-19 casks does not contradict Romanian, Russian and international regulations for the safe transport of radioactive material. The outcomes of the SNF shipment from Romania confirmed correctness of the solutions and demonstrated high environmental safety. (author)

  9. 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.

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

    International Nuclear Information System (INIS)

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

  11. MicroShield analysis to calculate external radiation dose rates for several spent fuel casks

    International Nuclear Information System (INIS)

    The purpose of this MicroShield analysis is to calculate the external radiation, primarily gamma, dose rate for spent fuel casks. The reason for making this calculation is that currently all analyses of transportation risk assume that this external dose rate is the maximum allowed by regulation, 10 mrem/hr at 2 m from the casks, and the risks of incident-free transportation are thus always overestimated to an unknown extent. In order to do this, the program by Grove Software, MicroShield 7.01, was used to model three Nuclear Regulatory Commission (NRC) approved casks: HI-STAR 100, GA-4, and NAC-STC, loaded with specific source material. Dimensions were obtained from NUREG/CR-6672 and the Certificates of Compliance for each respective cask. Detectors were placed at the axial point at 1 m and 2 m from the outer gamma shielding of the casks. In the April 8, 2004 publication of the Federal Register, a notice of intent to prepare a Supplemental Yucca Mountain Environmental Impact Statement (DOE/EIS-0250F-S1) was published by the Office of Civilian Radioactive Waste Management (OCRWM) in order to consider design, construction, operation, and transportation of spent nuclear fuel to the Yucca Mountain repository [1]. These more accurate estimates of the external dose rates could be used in order to provide a more risk-informed analysis. (authors)

  12. Dry storage of the BR3 spent fuel in the CASTOR BR3 cask

    International Nuclear Information System (INIS)

    The BR3 reactor was the first PWR plant installed in Europe. Started in 1962, BR3 was definitely shut down on June 30th, 1987. Used at the beginning of its life as a training device for commercial plant operators, it was also used during its whole life as test-reactor for new fuel types and assemblies. Most of the spent fuel was stored in the deactivation pool of the plant for more than 15 years. The reactor being now in decommissioning, it was decided to remove the spent fuel from the plant. After comparison of different solutions, the long term storage in dual purpose storage casks was selected in 1997. The selected CASTOR-BR3 cask is designed as a transport and storage cask for accommodating 30 spent fuel assemblies. As a type B(U) cask fitted with shock absorbers, it meets the transport requirements according to the IAEA guidelines and fulfils also the conditions for cask storage. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

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

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

    International Nuclear Information System (INIS)

    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)

  15. Behavior of Full-Scale Model Cask Under 9 m Drop Test and Simulation

    International Nuclear Information System (INIS)

    The nuclear spent fuel transport and storage cask is used for transport of the spent fuel from a nuclear power station to an intermediate storage facility. Leak tightness and subcriticality on transportation required from IAEA TS-R1[1] have to be assured by a 9 m drop test and its numerical simulation. This paper describes the drop test using a full-scale prototype test cask The test was conducted by German Federal Institute for Materials Research and Testing (BAM) at their test facility in Horstwalde, Germany and comparison of the test result with the 'MH1 (Mitsubishi Heavy Industries, Ltd.)' numerical simulation using LS-DYNA code. The drop orientations of the tests were slap down and vertical. From the drop test the following is demonstrated: The leak rate of He gas after the drop tests satisfied the IAEA's criteria. The numerical simulation which modeled the cask body enabled dynamic response such as acceleration and strain of the cask body. This means the simulation method qualified the relation of dynamic response of the cask body and leakage behavior. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181... 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear... of Approved Spent Fuel Storage Casks.'' This direct final rule allows the holders of power...

  1. An analysis of contingencies for making casks available for use during the early years of federal waste management system operations

    International Nuclear Information System (INIS)

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

  2. Long-term experience with the storage of spent fuel and vitrified high level waste in CASTOR and CONSTOR casks

    International Nuclear Information System (INIS)

    Two decades ago the world-wide first transport and storage cask-a CASTOR Ic-DIORIT-was loaded in Wurenlingen/Switzerland. Meanwhile CASTOR casks are used at 24 sites on four continents. Spent fuel assemblies of PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste containers are stored in these kinds of full metal casks. Also MOX fuel of PWR and BWR has been stored. Starting in the mid of the 90ies, GNB developed the new CONSTOR cask concept, which is based on a double liner technology with heavy concrete as shielding material in between. This CONSTOR cask concept fulfils as well all design criteria for transport and for storage given by the IAEA recommendations and by national authorities. The advantages of the CONSTORR casks are the much simpler manufacturing requirements compared to the fabrication of full metal casks. The CASTOR and CONSTOR casks have been thoroughly investigated by many experiments. There have been around 100 drop tests, a lot of them with full scale casks, fire tests, simulations of airplane crash, investigations with anti tank weapons, and an explosion of a railway tank with liquid gas beside a CASTOR cask. The casks are stored in especially designed buildings, on simple concrete pads, or on pads with additional shielding walls around it. Most of the casks are stored vertically, but some of them horizontally in a storage cradle. Up to now, around 700 CASTOR and CONSTOR casks have been loaded for long-term storage. This results in around 4,800 cask years and 24,000 metallic gasket years, allowing to draw conclusions with respect to the safety of dry storage, especially for the safe confinement of the radioactive inventory. The storage experience shows the excellent behaviour of the metallic gaskets and of the tightness control system. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in

  3. Neutron measurements around storage casks containing spent fuel and vitrified high-level radioactive waste at ZWILAG.

    Science.gov (United States)

    Buchillier, T; Aroua, A; Bochud, F O

    2007-01-01

    Spectrometric and dosimetric measurements were made around a cask containing spent fuel and a cask containing high-level radioactive waste at the Swiss intermediate waste and spent fuel storage facility. A Bonner sphere spectrometer, an LB 6411 neutron monitor and an Automess Szintomat 6134A were used to characterise the n-gamma fields at several locations around the two casks. The results of these measurements show that the neutron fluence spectra around the cask containing radioactive waste are harder and higher in intensity than those measured in the vicinity of the spent fuel cask. The ambient dose equivalents measured with the LB 6411 neutron monitor are in good agreement with those obtained using the Bonner spheres, except for locations with soft neutron spectra where the monitor overestimates the neutron ambient dose equivalent by almost 50%. PMID:17494980

  4. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    International Nuclear Information System (INIS)

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs

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

    International Nuclear Information System (INIS)

    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 significant 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

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

    International Nuclear Information System (INIS)

    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

  7. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  8. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  9. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    Science.gov (United States)

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

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

    International Nuclear Information System (INIS)

    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

  11. Preliminary assessment of the benefits of derating a cask for increasing age/burnup capability

    International Nuclear Information System (INIS)

    This study was performed to determine the extent to which the age/burnup capability of the Babcock and Wilcox BR-100 rail cask could be extended by reducing the number of fuel assemblies. Since cask shielding was seen as the limiting design feature, the criterion used to assess the derating was the calculated dose 2 m from the rail car. The reference calculations were based on the 70% design of the BR-100 cask with 21 PWR fuel assemblies. Seven different basket/assembly loading configurations were investigated. The results indicate that both an alternate 18-assembly basket configuration and a 17-assembly/4-empty-hole configuration for the 21-element basket offer substantial gains over the fully loaded reference 21-element basket configuration

  12. Criticality safety analysis of WWER-440 spent fuel cask with radial and axial burnup profile implementation

    International Nuclear Information System (INIS)

    The impact of radial and axial burnup profile on the criticality of WWER-440 spent fuel cask is presented in the paper. The calculations are performed based on two AER Benchmark problems for WWER-440 irradiated fuel assembly. The radial zonewise dependent spent fuel inventory has been calculated by the NESSEL - NUKO code system. The axial dependent isotope concentrations have been determined by the modular code system SCALE4.4. For criticality calculations the SCALE4.4 has been applied. Calculations have been carried out for cask with 30 WWER-440 fuel assemblies with initial enrichment 3.6% of 235U and burnup up to 40 MWd/kgU. The influence of radial and axial burnup credit on the cask criticality has been evaluated

  13. Development of boronated aluminum alloy for basket of cask for nuclear spent fuel

    International Nuclear Information System (INIS)

    Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities and competent authorities 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 cask uses new-developed boronated aluminum as basket material. This boronated aluminum has been developed to improve characteristics of material. To achieve this object, powder metallurgy method has been adopted for manufacturing boronated material. It is well known that this method provides excellent characteristics for the material and this boronated aluminum alloy has obtained excellent both mechanical and neutron absorbing characteristics. In addition, in order to maintain material properties for long-term use this boronated material is not strengthened by aging treatment. This paper summarizes an outline of the boronated aluminum alloy for basket assemblies by powder metallurgy. (author)

  14. Heat transfer investigations for spent fuel assemblies in a dry cask

    International Nuclear Information System (INIS)

    For studying the heat transfer processes and predicting the maximum spent fuel element surface temperature in a spent fuel assembly (SFA) transported in a dry cast, model experiments have been performed with a gas filled model cask containing a simplified electrically heated model of a WWER-type SFA with 90 fuel elements. The temperature distribution of the SFA model is measured for different heat rates under vacuum in the model cask, and under normal pressure and overpressure (0.1-0.7 MPa) for several cooling gases (air, argon, helium) in order to separately investigate heat transfer processes by radiation and convection/conduction. The results were compared with the calculations. Computer programs as well as simplified calculation methods for temperature prediction were developed and checked. The results obtained are also useful for thermal analyses in the field of the dry storage of SFAs in a cask or a can. (author)

  15. Lesson Learned from Drop and Tip-Over Test of a Dry Storage Cask System

    International Nuclear Information System (INIS)

    To study an appropriate storage system with the consideration of a nuclear power plant situation, a status evaluation of the technology and a feasibility study has been performed in Korea. As a part of doing these, a dry spent fuel storage cask has been developed. The dry storage cask under development consists of a cask body, a canister, and an in-canister structure. 24 fuel assemblies are stored in canister structure. The spacer disks in this in-canister structure are designed to dissipate the heat from the baskets and to provide a lateral support to the baskets. The support rods keep the spacer disks at an even interval. To assess a structural integrity after a hypothetical accident condition of this dry storage system, a 1/4 scale model is fabricated for the drop test. Drop tests of this test model were performed and the test results were also discussed

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  18. Development of Aircraft Impact Scenario on a Concrete Cask in Interim Storage Facility

    International Nuclear Information System (INIS)

    This paper provides a method for determining the failure criteria in global and local damage responses for the concrete cask under extreme mechanical impact condition. IAEA safety guide No. SSG-15 mentions the hypothetical initiating events of SNF storage. Among the external initiating events, the aircraft strike on a storage cask is considered one of the dominant contributions to the risk during storage phase. Although the probability of aircraft crash on ISF is extremely small, it is important to develop the accident scenario caused by an intentional malicious acts launched towards the storage facility in terms to improve inherent security. Thus, the probabilistic approach to develop aircraft impact scenarios on a storage cask is needed

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Behaviour of a spent fuel transport-storage cask during an airplane crash

    International Nuclear Information System (INIS)

    TRANSNUCLEAIRE has got an order for the design and manufacturing of dual purpose, transport and storage, casks for spent fuel.An original item of the qualification of the design of this cask, for the storage aspect, is the necessity to demonstrate the resistance to an air crash.The typical case taken into account for design is the crash of a military fighter (F16) with a total mass of 14600kg and an impact speed of 150ms-1. The demonstration of the ability of the cask to withstand this test is provided by both calculation and test.Two cases were considered. For the first one, the projectile hits the cask at the centre of the anti-crash lid. For the second one, it hits the cask in the plane of the closure system.The first step of the qualification is based on calculations performed with a code designed to study the effects of crashes. The aim of the calculations is, mainly, to determine the missile which has to be shot, and to select the worst orientation for the impact.To provide a full justification of the acceptability of the impact as concerned leaktightness, a test has been performed on a one-third scale model. It has shown that it was not altered by the impact.The paper provides a full description of the method of analysis, results of the numerical analysis, conclusion of the test and how the combination of calculation and test demonstrates the ability of the cask to withstand an airplane crash. ((orig.))

  3. Constor steel concrete sandwich cask concept for transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    A spent nuclear fuel transport and storage sandwich cask concept has been developed together with the Russian company CKTI. Special consideration was given to an economical and effective way of manufacturing by using conventional mechanical engineering technologies and common materials. The main objective of this development was to fabricate these casks in countries not having highly specialized industries. Nevertheless, this sandwich cask concept fulfills both the internationally valid IAEA criteria for transportation and the German criteria for long-term intermediate storage. The basic cask concept has been designed for adaptation to different spent fuel specifications as well as handling conditions in the NPP. Recently, adaptations have been made for spent fuel from the RBMK and VVER reactors, and also for BWR spent fuel. The analyses of nuclear and thermal behaviour as well as of strength according to IAEA examination requirements (9-m-drop, 1-m-pin drop, 800 deg. C-fire test) and of the behaviour during accident scenarios at the storage site (drop, fire, gas cloud explosion, side impact) were carried out by means of recognized calculation methods and programmes. In a special experimental programme, the mechanical and thermodynamic properties of heavy concrete were examined and the reference values required for safety analyses were determined. The results of the safety analysis after drop tests according to IAEA-regulations as well as after 1 m-drops at the storage site were confirmed by means of a test programme using a scale model. The fabrication technology has been tested with help of a half scale cask model. The model has been prefabricated in Russia and completed in Germany. It has been shown that the CONSTOR cask can be fabricated in an effective and economic way. (authors)

  4. GNS experience of CASTOR cask loading for storage and transport of spent fuel assemblies

    International Nuclear Information System (INIS)

    With over 25 years of experience in the design, manufacturing, assembly and loading of CASTOR registered casks, GNS is one of the worldwide leading suppliers of casks for the transport and storage of spent fuel assemblies as well as for canisters with vitrified high level wastes. GNS products are used at around 30 sites worldwide for a wide range of inventories from pressurized and boiling water reactor fuels (PWR and BWR), thorium high-temperature reactor fuels (THTR) and research reactor fuels to vitrified high-active wastes (HAW) from reprocessing plants

  5. TMI-2 (Three-Mile Island-Unit 2) rail cask and railcar maintenance

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-02-01

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

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

    International Nuclear Information System (INIS)

    General Atomics (GA), under contract to the Idaho Operations Office of the US 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. Our primary goal is 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. 4 figs

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  9. Dry Storage Casks Monitoring by Means of Ultrasonic Tomography

    Science.gov (United States)

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

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

  10. CASK/MSC/WP PREPARATION SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the Cask/MSC/WP preparation system and their bases to allow the design effort to proceed to license application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design, as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This type of SDD both leads and trails the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. This SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. This SDD addresses the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]) requirements. Additional PRD requirements may be cited, as applicable, to drive the design of specific aspects of the system, with justifications provided in the basis. Functional and operational requirements applicable to this system are obtained from the ''Project Functional and Operational Requirements'' (F and OR) (Curry 2004 [DIRS 170557]) document. Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (BSC 2004 [DIRS 171599]) and the preclosure safety analyses

  11. Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

    Energy Technology Data Exchange (ETDEWEB)

    Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky; Miroslav Picek; Alexey Smirnov; Sergey Komarov; Edward Bradley; Alexander Dudchenko; Konstantin Golubkin

    2007-10-01

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing, testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.

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

    International Nuclear Information System (INIS)

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

  13. Development of New Transportation/Storage Cask System for Use by DOE Russian Research Reactor Fuel Return Program

    International Nuclear Information System (INIS)

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian supplied high-enriched uranium (HEU) fuel currently stored at Russian-designed research reactors throughout the world to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions for these research reactor facilities. This new cask system is named the SKODA VPVR/M cask. The design, licensing, testing, and delivery of this new cask system are the results of a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: (1) Introduction/Background; (2) VPVR/M Cask Description; (3) Ancillary Equipment, (4) Cask Licensing; (5) Cask Demonstration and Operations; (6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, (7) Summary and Conclusions.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  17. Design and scale model testing of the NuPac 125-B rail cask

    International Nuclear Information System (INIS)

    The NuPac 125-B package was developed for defueling the damaged Three Mile Island Unit II (TMI-2) reactor. The nature and impact of these requirements on the design and licensing of a transport package are discussed in this paper. Loading and unloading of the NuPac 125-B differs from conventional fuel cask handling procedures due to facility features and limitations at both TMI-2 and the receiving station. All transfers are ''dry'' and the cask is never placed in a conventional fuel loading pool. In addition, the cask design was affected by the unique requirements for double containment of the TMI-2 fuel material, an accelerated development schedule, and limits imposed on impact loads experienced by the fuel debris canisters. Licensing activities involved analyses correlated with drop and puncture tests conducted on a 1/4 scale cask model. All structural details of the NuPac 125-B were accurately represented in these tests and analyses. Excellent correlation was found between analytical predictions and model behavior on the impact events, and basic structural design and analysis assumptions were validated. Used together, integrated test and analysis demonstrations are shown to accelerate the design and licensing process

  18. ITER Transfer Cask System: Status of design, issues and future developments

    International Nuclear Information System (INIS)

    The Remote Handling tasks scheduled during the ITER maintenance shutdown require transportation of in-vessel components from the Vacuum Vessel ports, at all levels of the Tokamak building, to the docking stations in the Hot Cell building. This transfer of radioactive, contaminated components represents a task of unprecedented complexity for a nuclear device like ITER. A Transfer Cask System (TCS) has been adopted as a reference solution. The TCS is a mobile, leak-tight unit, which can be divided into: (1) the cask itself, i.e., the container for the components to be transferred, able to avoid spread of contamination outside its envelope, equipped with in-cask handling devices; (2) the interface pallet that assists the docking operations of the cask and, underneath; (3) an Air Transfer System (ATS), i.e., a mobile platform floating on air-cushions with drive and steering wheels powered by electric motors and batteries on-board. The system will be remotely controlled, moving along previously defined paths. This paper focuses on the present status of the ATS design, the issues to be faced and the future developments foreseen.

  19. Full-scope simulation of a dry storage cask using computational fluid dynamics

    International Nuclear Information System (INIS)

    Graphical abstract: Display Omitted Research highlights: → A computational fluid dynamics (CFD) analysis of a TN24P cask was performed through a full-scope simulation using FLUENT. → The CFD predictions of the TN24P cask were compared with experimental data and COBRA-SFS results. → The comparisons between the experiment and FLUENT were in good agreement. → By sensitivity studies of various parameters, it was found that the basket gap sizes were the most sensitive parameter in the modeling. - Abstract: A computational fluid dynamics (CFD) analysis of a TN24P cask was performed through a full-scope simulation using FLUENT. In order to establish the analysis methodology while minimizing the computational burden, the sensitivities of various parameters were investigated by constructing a small-scale model. The full-scale CFD predictions of the TN24P cask were compared with the experimental data and COBRA-SFS results. There was good agreement between the FLUENT predictions and the experimental data. FLUENT showed similar temperature predictions to COBRA-SFS, while there were deviations between FLUENT and COBRA-SFS in the velocity predictions. By conducting sensitivity studies for the application uncertainties using a full-scale simulation, it was found that the basket gap size was the most sensitive parameter in the analysis.

  20. Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

    International Nuclear Information System (INIS)

    Highlights: • Design on Intelligent Air Transfer System (IATS) for Transfer Equipment Cask (TECA). • A rhombic-like parallel robot for docking with minimum misalignment. • Design on electro-hydraulic servo system of the TECA for Tokamak Equator Port Plug (TEPP) manipulation. • A control architecture with several algorithms and information acquired from sensors could be used by the TECA for Remote Handling (RH). - Abstract: The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER

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

    International Nuclear Information System (INIS)

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

  2. 75 FR 49813 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of...

    Science.gov (United States)

    2010-08-16

    ... 3150-AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of..., 2010, for the direct final rule that was published in the Federal Register on June 15, 2010 (75 FR 33678). This direct final rule amended the NRC's spent fuel storage regulations at 10 CFR 72.214...

  3. 77 FR 9515 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear... Commission) is amending its spent fuel storage regulations by revising the Holtec International HI-STORM...

  4. 75 FR 24786 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-06

    ..., entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990... COMMISSION 10 CFR Part 72 RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS HD System Revision... Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the......

  5. 75 FR 42292 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ... for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also... COMMISSION 10 CFR Part 72 RIN 3150-AI88 List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6... Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International Inc....

  6. 76 FR 9381 - Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks

    Science.gov (United States)

    2011-02-17

    ... COMMISSION Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks AGENCY... Gordon, Structural Mechanics and Materials Branch, Division of Spent Fuel Storage and Transportation... performing technical reviews of spent fuel storage and transportation packaging licensing actions.'' This...

  7. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false List of approved spent fuel storage casks. 72.214 Section... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved...

  8. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ... License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also... 3150-AJ22 List of Approved Spent Fuel Storage Casks: MAGNASTOR System AGENCY: Nuclear Regulatory... spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced...

  9. Analysis of the non-cylindrical GA-4 and GA-9 spent fuel casks

    International Nuclear Information System (INIS)

    DOE's Office of Civilian Radioactive Waste Management (OCRWM) has awarded General Atomics (GA) a contract to develop the GA-4 and GA-9 legal weight truck (LWT) transportation system to transport pressurized-water-reactor (PWR) and boiling-water-reactor (BWR) spent fuels. The GA-4 and GA-9 Casks maximize cargo capacity while complying with the weight limits imposed on legal weight truck transport. These casks will carry up to 4 PWR or 9 BWR spent fuel assemblies, a capacity four times greater than comparable existing designs. The approach to the structural analysis of a non-cylindrical cross-section differs from a cylindrical design. These differences include: More orientations are evaluated for impact analyses; the structural analyses are more complex; and applicable criteria are more conservative and some require additional development. GA's structural analyses of the GA-4 and GA-9 Casks address each of these analytical differences, resulting in casks that meet the criteria imposed by applicable regulatory guides and the ASME Code. When necessary, we have confirmed our analytical approach with component testing. GA will obtain further confirmation during model testing in 1990. 2 refs., 7 figs

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

    International Nuclear Information System (INIS)

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

  11. Impact stress reduction by shell splitting in cask for transporting radioactive material

    International Nuclear Information System (INIS)

    Highlights: • High impact stress in shell of a container for transporting radioactive material. • Reduction of impact stress by splitting shell into multiple parts. • Impact simulations on simple objects to prove benefits of shell splitting. • Explanation based on theory of bending of simply supported beam. • Impact simulations on a simple cask showing up to 21% reduction in maximum stress. - Abstract: Casks designed for transporting radioactive material are mandated to withstand drop from specific heights on hard ground. The maximum internal stress in the shell of the cask after such an impact needs to be as low as possible to ensure safety of the material being transported. This paper investigates the concept of splitting the shell of the radioactive transport container into multiple layers to reduce these stresses after impact. Different geometrical configurations which are likely to be encountered while designing such containers have been studied through plane 2D and 3D finite element analysis and the efficacy of this idea has been explored on each of them. Considerable reduction of stress has been reported and an explanation based on elastic deformation of layered beams has been suggested. Simulations on a cask with the currently prevalent design also show the benefit of implementing this idea

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  14. Imaging Spent Fuel in Dry Storage Casks with Cosmic Ray Muons

    Energy Technology Data Exchange (ETDEWEB)

    Durham, J. Matthew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dougan, Arden [National Nuclear Security Administration (NNSA), Washington, DC (United States)

    2015-11-05

    Highly energetic cosmic ray muons are a natural source of ionizing radiation that can be used to make tomographic images of the interior of dense objects. Muons are capable of penetrating large amounts of shielding that defeats typical radiographic probes like neutrons or photons. This is the only technique which can examine spent nuclear fuel rods sealed inside dry casks.

  15. Separator assembly for use in spent-nuclear-fuel shipping cask. [Patent application

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1981-04-24

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

  16. Criticality calculations of various spent fuel casks - possibilities for burn up credit implementation

    International Nuclear Information System (INIS)

    A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion Keff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)

  17. Evaluation of burnup credit for accommodating PWR spent nuclear fuel in high-capacity cask designs

    International Nuclear Information System (INIS)

    This paper presents an evaluation of the amount of burnup credit needed for high-density casks to transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic 32-assembly cask and the current regulatory guidance were used as bases for this evaluation. By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based loading curves, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of SNF assemblies in high-capacity storage and transportation casks. The impact of varying selected calculational assumptions is also investigated, and considerable improvement in effectiveness is shown with the inclusion of the principal fission products (FPs) and minor actinides and the use of a bounding best-estimate approach for isotopic validation. Given sufficient data for validation, the most significant component that would improve accuracy, and subsequently enhance the utilization of burnup credit, is the inclusion of FPs. (author)

  18. Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Buyun, E-mail: ayun@iim.ac.cn [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Gao, Lifu [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Cao, Huibin; Sun, Jian [Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Sun, Yuxiang; Song, Quanjun; Ma, Chengxue; Chang, Li; Shuang, Feng [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2014-12-15

    Highlights: • Design on Intelligent Air Transfer System (IATS) for Transfer Equipment Cask (TECA). • A rhombic-like parallel robot for docking with minimum misalignment. • Design on electro-hydraulic servo system of the TECA for Tokamak Equator Port Plug (TEPP) manipulation. • A control architecture with several algorithms and information acquired from sensors could be used by the TECA for Remote Handling (RH). - Abstract: The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.

  19. Design of a spent fuel shipping cask for Korea Nuclear unit-1

    International Nuclear Information System (INIS)

    To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium considering the aspects of transportability, cost, fabricability and safety. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail car. The cask requires 33 cm thick steel shield and 27 cm thick water region to satisfy the 3 feet apart dose limit set forth in 10CFR 71, and 1.27 cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel temperature was calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. ksub(eff) was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radiation source terms were calculated using ORIGEN-79. (Author)

  20. Design study of a spent fuel shipping cask for Korea nuclear unit, 1

    International Nuclear Information System (INIS)

    To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail-car. The cask requires 33cm thick steel shield and 27cm thick water region to satisfy the 3 feet apart dose rate limit set forth in 10 CFR 71, and 1.27cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel cladding and centerline temperatures were calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. Keff was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radioactivity source terms were calculated using ORIGEN-79. (author)

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

    International Nuclear Information System (INIS)

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

  2. Structural dimensioning of dual purpose cask prototype; Dimensionamento estrutural de prototipo de casco de duplo proposito

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Luiz Leite da; Mourao, Rogerio Pimenta; Lopes, Claudio Cunha [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail: silvall@cdtn.br; mouraor@cdtn.br; ccl@cdtn.br

    2005-07-01

    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

  3. Studies and research concerning BNFP: operational assessment of the FSV-1 (HTGR) spent fuel shipping cask in alternate modes

    International Nuclear Information System (INIS)

    This report presents an operational assessment of the FSV-1 spent fuel shipping cask which was developed by General Atomic specifically for the Fort St. Vrain reactor. Of primary interest is the adaptation to underwater loading and unloading of light-water-reactor (LWR) fuel in this cask which was designed for the dry-handling of high temperature gas cooled reactor (HTGR) fuel. Also presented is a concept for a system to compare the pros and cons of wet and dry handling of this and other casks

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

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

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

    International Nuclear Information System (INIS)

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

  11. Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

    Energy Technology Data Exchange (ETDEWEB)

    Krementz, Dan; Rose, David; Dunsmuir, Mike

    2014-02-06

    The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit #2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24” diameter.[1, 2] This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18” in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit #2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24” in diameter and ~11 feet long from a dry transfer cask to the basin. The 18” and 24” applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit #2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with

  12. The experiences from interim spent fuel storage operation with CASTOR 440/84 CASKS in NPP Dukovany

    International Nuclear Information System (INIS)

    In this lecture are presented: principles of the CASTOR 440/84 design; design development works; commissioning of interim spent fuel storage facility; international transports of spent fuel utilising CASTOR 440/84 casks

  13. Operations of the LR56 radioactive liquid cask transport system at U.S. Department of Energy sites

    International Nuclear Information System (INIS)

    The LR56 cask system is licensed for use in France under Certificate of Compliance F/309/B(U)F for transport of 4,000-liter volumes of radioactive liquids. Three LR56 cask systems (with modifications for use at Department of Energy (DOE) sites) have been purchased for delivery at the Hanford Site, Oak Ridge National Laboratory (ORNL), and Savannah River Site (SRS). The LR56 cask systems will be used for on-site transfers of Type B quantities of radioactive liquid waste. The ORNL unit will also be used as a Type A packaging for transfers of radioactive liquids between DOE sites. This paper discusses LR56 operating features and the use of the cask system at the three DOE sites

  14. Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

    International Nuclear Information System (INIS)

    The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit no. 2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24'' diameter. This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18'' in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit no. 2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24'' in diameter and ∼11 feet long from a dry transfer cask to the basin. The 18'' and 24'' applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit no. 2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with

  15. Operations of the LR56 radioactive liquid cask transport system at U.S. Department of Energy sites

    Energy Technology Data Exchange (ETDEWEB)

    Davidson, J.S. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States); Hornstra, D.J. [Performance Development Corp., Oak Ridge, TN (United States); Sazawal, V.K. [NUMATEC, Inc., Bethesda, MD (United States); Clement, G. [SGN, St. Quentin en Yvelines (France)

    1996-06-01

    The LR56 cask system is licensed for use in France under Certificate of Compliance F/309/B(U)F for transport of 4,000-liter volumes of radioactive liquids. Three LR56 cask systems (with modifications for use at Department of Energy (DOE) sites) have been purchased for delivery at the Hanford Site, Oak Ridge National Laboratory (ORNL), and Savannah River Site (SRS). The LR56 cask systems will be used for on-site transfers of Type B quantities of radioactive liquid waste. The ORNL unit will also be used as a Type A packaging for transfers of radioactive liquids between DOE sites. This paper discusses LR56 operating features and the use of the cask system at the three DOE sites.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  18. 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.

  19. 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.

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

    Directory of Open Access Journals (Sweden)

    Rezaeian Mahdi

    2015-01-01

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

  1. A new type-B cask design for transporting 252Cf

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Full text: All operating nuclear power reactors in the United States (US) are storing spent fuel in NRC licensed on-site spent fuel pools (SFPs). Most reactors were not designed to store, in these pools, the full amount of spent fuel generated during the life of plant operation. Utilities originally planned for spent fuel to remain in the SFPs for a few years after discharge from the reactor core and then to be sent to a reprocessing facility. However, the US Government declared a moratorium on reprocessing in 1977. Although the ban was later lifted, reprocessing has not been pursued as a feasible option. Consequently, utilities expanded the storage capacity of SFPs by the use of high-density storage racks. Eventually, utilities needed additional storage capacity. In response to these needs, NRC provided a regulatory alternative for interim spent fuel storage in dry cask storage systems. For spent fuel management, both pool storage and dry storage are safe methods, but there are significant differences. Pool storage requires a greater operational vigilance on the part of the nuclear power plant to maintain the performance of electrical and mechanical systems using pumps, piping and instrumentation. Dry storage technology uses passive cooling systems with robust cask designs requiring minimal operational vigilance. The United States Nuclear Regulatory Commission (NRC), through the combination of a rigorous licensing and inspection program, ensures the safety and security of dry cask storage. NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site-specific licensing and general licensing. In July 1986, the NRC issued the first site-specific license to the Surry Nuclear Power Plant in Virginia, authorizing the interim storage of spent fuel in a dry storage cask configuration. Today, there are over 30 ISFSIs currently licensed by the NRC with over 700 loaded dry casks. Current projections

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. Integrity assessment of dual-purpose metal CASK after long term interim storage - seal performance under transport conditions

    International Nuclear Information System (INIS)

    Spent fuels generated in nuclear power plants (NPPs) shall be stored until reprocessing as recyclable energy resources in Japan. The quantity of spent fuels stored at each NPP site is increasing, and early realization of the interim storage is expected. Dual-purpose metal cask will be used there and will not be reopened until it is delivered to a reprocessing plant in order to for example minimize personal exposure of radiation. Japan Nuclear Energy Safety Organization (JNES) was established on October in 2003 with the mission to ensure the public safety from the potential hazard of nuclear energy. Study of ''Metal Cask Storage Technology Verification'' was originally initiated in 1999 at Nuclear Power Engineering Corporation with Japanese government funds, and transferred to JNES and conducted up to the end of FY2003. In the study, many tests were conducted to investigate material property change for main components relating to safety of cask due to degradation during interim storage, furthermore, to verify containment safety during the subsequent transport because there was a possibility of providing the cask with degraded metal gasket for the transport after interim storage and the such cask should be considered fragileness of lid containment system, especially for transport that cask would be provided external force. This paper presents the results and consideration on seal performance during the subsequent transport

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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: (1) a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); (2) a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); (3) a small hot cell containing equipment for transferring fuel between shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept); and (4) 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. 1 reference, 4 figures, 2 tables

  10. TGC36 a dual purpose cask for the transport and interim storage of compacted waste (CSD-C)

    International Nuclear Information System (INIS)

    According to contractual and international obligations, the German Utilities have to return the residues resulting from the reprocessing of nuclear fuel assemblies (compacted hulls and ends) to Germany. The new dual purpose cask TGC36 is a joint product from the two leading companies in the field development and manufactory of nuclear casks in Europe, GNS and TN International, is intended for the transport to the interim storage facility Ahaus and to be stored there for up several years. For the development and the delivery of the TGC36 cask, GNS and TN International formed the AGC Consortium based on German law to combine the special know how of both partners in the most efficient way. The design and the licensing strategy of the TGC36 are introduced in this paper. In conclusions: GNS and TNI have formed a consortium named AGC to design, license and manufacture an innovative cask for the transport and the interim storage of the compacted wastes resulting from the reprocessing of the German spent fuel. This cask has been optimized in order to offer a high capacity of loading, and allows a payload of 36 canisters, leading to a total mass of approximately 116 Mg in transport configuration. The success of this project requires a special effort from both partner companies, members of the consortium, and implies also an efficient management of simultaneous tasks during the licensing period and the manufacturing time of the first items of the cask. (authors)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-02-01

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

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

    International Nuclear Information System (INIS)

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

  14. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    Energy Technology Data Exchange (ETDEWEB)

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

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

    International Nuclear Information System (INIS)

    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)

  16. Simulation of Multi Canister Overpack (MCO) Handling Machine Impact with Cask and MCO During Insertion into the Transfer Pit (FDT-137)

    International Nuclear Information System (INIS)

    The K-Basin Cask and Transportation System will be used for safely packaging and transporting approximately 2,100 metric tons of unprocessed, spent nuclear fuel from the 105 K East and K West Basins to the 200 E Area Canister Storage Building (CSB). Portions of the system will also be used for drying the spent fuel under cold vacuum conditions prior to placement in interim storage. The spent nuclear fuel is currently stored underwater in the two K-Basins. The K-Basins loadout pit is the area selected for loading spent nuclear fuel into the Multi-Canister Overpack (MCO) which in turn is located within the transportation cask. This Cask/MCO unit is secured.in the pit with a pail load out structure whose primary function is lo suspend and support the Cask/MCO unit at.the desired elevations and to protect the unit from the contaminated K-Basin water. The fuel elements will be placed in special baskets and stacked in the MCO that have been previously placed in the cask. The casks will be removed from the K Basin load out areas and taken to the cold vacuum drying station. Then the cask will be prepared for transportation to the CSB. The shipments will occur exclusively on the Hanford Site between K-Basins and the CSB. Travel will be by road with one cask per trailer. At the CSB receiving area the cask will be removed from the trailer. A gantry crane will then move the cask over to the transfer pit and load the cask into the transfer pit. From the transfer pit the MCO will be removed from the cask by the MCO Handling Machine (MHM). The MHM will move the MCO from the transfer pit to a canister storage tube in the CSB. MCOs will be piled two high in each canister Storage tube

  17. Design of a redundant-load-path lifting fixture for the Clinch River Breeder Reactor Plant spent-fuel shipping cask

    International Nuclear Information System (INIS)

    A newly developed concept for a redundant-load-path lifting fixture for spent fuel shipping casks is presented. The concept addresses remote attachment of the crane hook to the lifting fixture and remote attachment of the lifting fixture to this cask, and also allows the cask to be rotated from its shipping position to the vertical and lifted from its transport vehicle without requiring hands-on action

  18. Radiological Risk Assessment and Cask Materials Qualification for Disposed Sealed Radioactive Sources Transport

    International Nuclear Information System (INIS)

    The hazardous waste problem imposes to respect national and international agreed regulations regarding their transport, taking into account both for maintaining humans, goods and environment exposure under specified limits, during transport and specific additional operations, and also to reduce impact on the environment. The paper follows to estimate the radiological risk and cask materials qualification according to the design specifications for disposed sealed radioactive sources normal transport situation. The shielding analysis has been performed by using Oak Ridge National Laboratory's SCALE 5 programs package. For thermal analysis and cask materials qualification ANSYS computer code has been used. Results have been obtained under the framework of Advanced system for monitoring of hazardous waste transport on the Romanian territory Research Project which main objective consists in implementation of a complex dual system for on-line monitoring both for transport special vehicle and hazardous waste packages, with data automatic transmission to a national monitoring center

  19. Design and fabrication of the retube transfer cask for Bruce N.G.S

    International Nuclear Information System (INIS)

    The retubing of CANDU reactors is a complex process which involves the removal and disposal of highly activated and contaminated calandria tubes and pressure tubes. For Bruce 'A' N.G.S., old pressure tubes will be removed from the reactor by cutting them into three segments; two end fitting assemblies which measure up to 3.2 m in length, and one pressure tube segment which measures up to 6.3 m. in length. Calandria tubes are 6.2 m in length. The function of the retube transfer cask is to provide for shielded transfer of these components between the reactor face and the in-ground disposal facility. This paper describes the design and fabrication of this cask. (author) 1 tab., 7 figs

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

    International Nuclear Information System (INIS)

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

  1. Dual Purpose Cask for Dry Storage of Research Reactor Spent Fuel in Latin America

    International Nuclear Information System (INIS)

    Since 2001 Brazilian researchers have participated in a regional initiative, with researchers from other Latin American countries whom operate research reactors, to improve the regional capability in the management of spent fuel elements from these reactors. A dual purpose cask for transport and storage was selected as the best option for the long term dry storage of this material, and a half-scale model was designed, built and tested. Although the model failed the tests, its overall performance was considered very satisfactory and design and constructive features were changed as a result of the tests. A new test sequence with the modified cask model was scheduled for the first quarter of 2010. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    This Preliminary Design Report presents the results of General Atomics' (GA) preliminary design effort to develop legal weight truck From-Reactor Spent Fuel Shipping Casks. The report consists of a design description, preliminary drawings, and the results of the structural, thermal, containment and nuclear evaluations that support the design. Also included are the results of trade-off studies in which we considered the effect of changing several basic parameters on our baseline design as required by the contract. Our engineering test program supports the selection of the neutron shield material and the honeycomb impact limiter design. The design report also includes preliminary drawings and a structural analysis of a semitrailer designed specifically for the GA-4 cask. 24 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

  5. Extended Burnup Impact on the TN24 Spent Fuel Storage Cask Main Parameters

    International Nuclear Information System (INIS)

    In order to establish the capability of the TN24 cask for storage of spent fuel assemblies which are beyond the limits given by the manufacturer, a calculations of the dose and heat decay have been made for several cases of burnup higher than 35 GWd/MTU, using the SCALE 4.2 code package. The results were compared with the data obtained from the manufacturer. According to the results of the ORIGEN and SAS4 calculations and taking into the account limitations of the used model, it is possible to estimate that for 50 GWd/MTU burnup at least 15 years cooling time period is necessary to allow the use of TN24 cask. (author)

  6. Upgrading spent fuel shipping casks to meet higher burn-up

    International Nuclear Information System (INIS)

    In order to allow the transportability of high burn-up fuel and of MOX fuel in existing casks, TRANSNUCLEAIRE presents a two-step proven solution: (1) starting from 35/40 GWd/tU and 3.5 % enrichment, casks of the TN 12 family can be upgraded to 40/45 GWd/tU and 4.3 % enrichment by the use high performance baskets. (2) a second step consists in adding neutron shielding to allow transportation of fuel with a burn-up of 45/50 GWd/tU with a standard basket and of 50/55 GWd/tU with a high performance basket. (J.P.N.)

  7. Quality assurance on DCI transport/storage casks for radioactive material

    International Nuclear Information System (INIS)

    The Gesellschaft for Nuclear-Service (GNS), owned by German utilities, is responsible, on their behalf, for the management of spent fuel and radioactive waste, resulting from the operation of their nuclear power plants. The company's other responsibilities include the development, design and fabrication of the transport and storage casks thus required. Questions of quality assurance obviously play an important part here. Using the example of a spent fuel cask of the CASTOR family, the performance of quality assurance measures is explained. These start with the company's quality assurance system and include control activities during design and fabrication. In accordance with German regulations, fabrication and test sequences plans are used, which determine the inclusion of all parties involved at the time required. The interplay of different review authorities is being illustrated

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

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

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

  9. Benchmark calculations of neutron dose rates at transport and storage casks

    International Nuclear Information System (INIS)

    The application of numerical calculations methods for demonstration of sufficient radiation shielding of radioactive waste transport and storage casks requires a validation based on appropriate measurements of gamma and neutron sources. The results of the comparison of measured data and calculations using the Monte Carlo program MCNP show deviations dependent on the loading of the cask within the standard deviation which is dominated by the measuring method. Considering the neutrons scattered at the salt MCNP (in case of disposal in the salt) tends to underestimate the nominal values, but still within the double standard deviation. This accuracy is not reached with MAVRIC. Based on AHE (active handling experiments) data benchmark calculations were performed that can be used as reference value. The total accuracy results from the accuracy of the source term and the measurement of the neutron dose rate with a deviation of 15%.

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

    International Nuclear Information System (INIS)

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

  11. Direct disposal of transport an storage casks - status of the actual considerations

    International Nuclear Information System (INIS)

    For the final disposal of spent fuel elements and radioactive wastes from the spent fuel reprocessing two different concepts exist. The self-shielding POLLUX casks were developed for final disposal of spent fuels in underground repositories (gallery storage). For the high-level waste from reprocessing plants the concept of borehole storage of vitrified coquilles BSK3 was developed. for both concepts fuel elements and structural parts are supposed to be separated in conditioning facilities. An alternative concept (projects DIREGT) aimed to avoid conditioning is based on the direct final storage of transport and storage casks of the type CASTOR registered V in boreholes. The concepts have to consider the transport in the underground facility; the safety against criticality has to be demonstrated. An appropriate manipulation technique is to be developed.

  12. Development of a transport cask for spent fuel elements of research reactors

    International Nuclear Information System (INIS)

    This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

  13. Ductile iron cask with encapsulated uranium, tungsten or other dense metal shielding

    International Nuclear Information System (INIS)

    In a cask for the transportation and storage of radioactive materials, an improvement in the shielding means which achieves significant savings in weight and increases in payload by the use of pipes of depleted uranium, tungsten or other dense metal, encapsulating polyethylene cores, dispersed in two to four rows of concentric boreholes around the periphery of the cask body which is preferably made of ductile iron. Alternatively, rods or small balls of these same shielding materials, alone or in combination, are placed in these bore holes. The thickness, number and arrangement of these shielding pipes or rods is varied to provide optimum protection against the neutrons and gamma radiation emitted by the particular radioactive material being transported or stored. (author) 4 figs

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

    International Nuclear Information System (INIS)

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

  15. 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)

    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)

  16. Development of defect size determination procedure in a cask of WWER defective assembly detection system

    International Nuclear Information System (INIS)

    At present time an integrated approach to analysis of fuel failures in WWER reactor core is under development. It includes analysis of defective Fuel Assemblies (FAs) under operation conditions by data on primary coolant activity and analysis of failed FAs in leakage tests during refuelling. At refuelling a system of defective assembly detection (SDAD) is used in WWERs. The conventional technique for detection of leaking FAs is based on measurement of coolant activity for reference radioactive nuclides after pressure elevation and drop in the SDAD circuit. To make possible an evaluation of leak size in leakage tests, the conventional technique should be modified. For this purpose an out-of-pile experimental facility CASK has been built which is a small-scale analogue to the SDAD system. Experimental investigations at the CASK facility resulted in development of a new leakage test technique with pressure cycling and monitoring of activity release kinetics in the SDAD circuit. A mechanistic RTOP-LT code has been developed for scaling of the CASK experimental results to real parameters of the SDAD system at NPPs. The RTOP-LT code is capable of modelling the kinetics of activity release from failed fuel rods in leakage tests with different scenarios of pressure and temperature variation in the circuit. The present paper reviews experiments at the CASK facility for verification of the RTOP-LT code and verification results. Computational and experimental proof is presented. The leakage tests with pressure cycling and monitoring of activity release kinetics give a proper evaluation of the leak size. Results of testing of the modified technique at NPPs with WWER reactors are presented as well. (authors)

  17. Concept for an all-purpose transport, storage, and disposal cask for spent nuclear fuel management

    International Nuclear Information System (INIS)

    The Tennessee Valley Authority believes that taking a systems approach to overall integration of spent fuel management with respect to onsite storage and disposal is essential. Their studies show that development of an integrated dry cask system suitable for onsite storage, transportation, monitored retrievable offsite storage, and perhaps use as a disposal container in a geologic repository offers the potential of the lowest overall economic, environmental, and social cost related to spent fuel management. 5 figures, 4 tables

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

    International Nuclear Information System (INIS)

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

  19. Experimental verification of a dynamic finite element analysis for a ductile iron cask

    International Nuclear Information System (INIS)

    The paper summarizes the results of an instrumented 9 meter drop test of a Cylindrical Ductile Iron Cask with shock absorbers at the BAM test facility compared to a stress analysis performed with the dynamic finite computer code DYNA 3D (GNS). The comparison between the results obtained from the experiment and the calculation, which was performed before the drop test, is according to strain and acceleration time history during the impact and the deformation of the shock absorber. (author)

  20. Conceptual design for concrete storage cask and landscape design of its storage facility

    International Nuclear Information System (INIS)

    As an advantageous interim spent fuel dry storage system at a reactor site, the conceptual study of the vertical dry cask spent fuel storage system and the landscape evaluation of the facility are carried out. The system is concluded to be viable means for the spent fuel storage and landscape design is confirmed to be necessary for environment conservation and for avoiding the damage of a natural grand view. (author)