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. 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

  4. CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    E.F. Loros

    2000-06-23

    The Carrier/Cask Handling System receives casks on railcars and legal-weight trucks (LWTs) (transporters) that transport loaded casks and empty overpacks to the Monitored Geologic Repository (MGR) from the Carrier/Cask Transport System. Casks that come to the MGR on heavy-haul trucks (HHTs) are transferred onto railcars before being brought into the Carrier/Cask Handling System. The system is the interfacing system between the railcars and LWTs and the Assembly Transfer System (ATS) and Canister Transfer System (CTS). The Carrier/Cask Handling System removes loaded casks from the cask transporters and transfers the casks to a transfer cart for either the ATS or CTS, as appropriate, based on cask contents. The Carrier/Cask Handling System receives the returned empty casks from the ATS and CTS and mounts the casks back onto the transporters for reshipment. If necessary, the Carrier/Cask Handling System can also mount loaded casks back onto the transporters and remove empty casks from the transporters. The Carrier/Cask Handling System receives overpacks from the ATS loaded with canisters that have been cut open and emptied and mounts the overpacks back onto the transporters for disposal. If necessary, the Carrier/Cask Handling System can also mount empty overpacks back onto the transporters and remove loaded overpacks from them. The Carrier/Cask Handling System is located within the Carrier Bay of the Waste Handling Building System. The system consists of cranes, hoists, manipulators, and supporting equipment. The Carrier/Cask Handling System is designed with the tooling and fixtures necessary for handling a variety of casks. The Carrier/Cask Handling System performance and reliability are sufficient to support the shipping and emplacement schedules for the MGR. The Carrier/Cask Handling System interfaces with the Carrier/Cask Transport System, ATS, and CTS as noted above. The Carrier/Cask Handling System interfaces with the Waste Handling Building System for building

  5. 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

  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. 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

  8. 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...

  9. 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

  10. 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.

  11. 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.

  12. 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)

  13. 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

  14. 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.

  15. 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

  16. 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 ...

  17. Test Plan for Cask Identification Detector

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-29

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

  18. Radioactive fuel cask railcar humping study

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, L.T. (comp.)

    1978-01-01

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

  19. 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

  20. 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)

  1. 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.

  2. 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

  3. 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

  4. 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.

  5. Design review report FFTF interim storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.L.

    1995-01-03

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

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

    Science.gov (United States)

    2013-12-26

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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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...

  15. 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

  16. 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

  17. 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...

  18. 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

  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. 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

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ziock, K P; Vanier, P; Forman, L; Caffrey, G; Wharton, J; Lebrun, A

    2005-07-27

    This report documents a week-long measurement campaign conducted on six, dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. A gamma-ray imager, a thermal-neutron imager and a germanium spectrometer were used to collect data on the casks. The campaign was conducted to examine the feasibility of using the cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the germanium spectrometer and differences between thermal neutron images and neutron-dose meters, this result is thought to be due to the limitations of the extant imagers used, rather than of the basic concept. Results indicate that measurements with improved imagers could contain significantly more information. Follow-on measurements with new imagers either currently available as laboratory prototypes or under development are recommended.

  3. 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

  4. 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.)

  5. 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

  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. Safety analysis report for EPMA irradiated specimen cask

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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)

  13. 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.

  14. 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)

  15. 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.

  16. 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....

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

    Science.gov (United States)

    2013-12-26

    ...; ] 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...

  18. 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

  19. 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

  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. 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

  2. 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)

  3. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-28

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

  6. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-10-01

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

  7. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-10-01

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

  8. Nuclear cask testing films misleading and misused

    International Nuclear Information System (INIS)

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

  9. 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.

  10. 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.

  11. What drives Greek consumer preferences for cask wine?

    DEFF Research Database (Denmark)

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

    2012-01-01

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

  12. 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)

  13. 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)

  14. 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

  15. 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.)

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  18. 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

  19. 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)

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-10-01

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

  3. 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)

  4. 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)

  5. 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

  6. 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.

  7. 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.)

  8. 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

  9. 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....

  10. 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...

  11. 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''...

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

    Energy Technology Data Exchange (ETDEWEB)

    Samuel Bays; Ayodeji Alajo

    2010-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. 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.)

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. Thermoelectric Powered Wireless Sensors for Dry-Cask Storage

    Science.gov (United States)

    Carstens, Thomas Alan

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

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. 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.

  12. 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

  13. 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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

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

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. 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

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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

  10. 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).

  11. 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.

  12. 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...

  13. 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)

  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. 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. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  18. 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.

  19. 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.

  20. 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)

  1. 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

  2. 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)

  3. 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.

  4. 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...

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. 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.

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

    International Nuclear Information System (INIS)

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

  14. 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

  15. 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

  16. 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:...

  17. 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...

  18. 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''...

  19. 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)

  20. 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.

  1. 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

  2. 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...

  3. 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,......

  4. 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,...

  5. 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''...

  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. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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

  12. 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)

  13. 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

  14. CLASSIFICATION OF THE MGR CARRIER/CASK HANDLING SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    J.A. Ziegler

    2001-02-08

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

  15. 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.

  16. 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

  17. 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

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

    International Nuclear Information System (INIS)

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

  19. 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)

  20. 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

  1. 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.

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

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

    International Nuclear Information System (INIS)

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

  5. Transfer cask system design activities: status and plan

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

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

    International Nuclear Information System (INIS)

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

  7. 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)

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  10. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-01

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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. 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)

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B A; Gwinn, K W

    1984-05-01

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

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-04-01

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

  15. 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

  16. COBRA-SFS modifications and cask model optimization

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

  18. 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

  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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  20. 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.

  1. 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

  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. 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

  5. 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

  6. 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).

  7. 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.

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  12. 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

  13. 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

  14. 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

  15. An attempt to estimate gamma-ray dose rate from radioactive waste storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M.; Plecas, I.; Pavlovic, R.; Pavlovic, S. [Institute for Nuclear Sciences ' ' Vinca' ' , Belgrade (Yugoslavia); Sokcic-Kostic, M. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Hauptabteilung Zyklotron

    2001-07-01

    Radioactive waste - {sup 137}Cs and {sup 60}Co contaminated sludge from the irradiated fuel storage pool of the RA research reactor in the Vinca Institute of Nuclear Sciences, Belgrade, is conditioned and stored in specially designed casks in 1997. Main purpose of this paper is to describe an attempt to estimate a gamma-ray ambient dose equivalent rate from the cask with the conditioned sludge by reference Monte Carlo MCNP code and compare the result to the measuring data. The aim of the study is to master with a reliable computational tool that allows faithful estimation of the total ambient gamma-ray dose equivalent rates from the cask. Values of so obtained gamma-ray ambient dose equivalent rates are compared to the measured values at the same spots of the cask and acceptable agreement were found. These data could be used in a further study on minimisation of the total ambient dose equivalent rate of a regular array or random pile of casks, stored in the storage location. (orig.)

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L. (Sandia National Labs., Albuquerque, NM (United States)); Jordan, H. (EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant); Pasupathi, V. (Battelle, Columbus, OH (United States)); Mings, W.J. (USDOE Idaho Field Office, Idaho Falls, ID (United States)); Reardon, P.C. (GRAM, Inc., Albuquerque, NM (United States))

    1991-09-01

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

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-04-01

    The loading experienced by spent fuel dry storage casks and storage pads due to potential drop or tip-over accidents is evaluated using state-of-the-art concrete structural analysis methodology. The purpose of this analysis is to provide simple design charts and formulas so that design adequacy of storage pads and dry storage casks can be demonstrated. The analysis covers a wide range of slab-design parameters, e.g., reinforcement ratio, slab thickness, concrete compressive strength, and sub-base soil compaction, as well as variations in drop orientation and drop height. The results are presented in the form of curves, giving the force on the cask as a function of storage pad hardness for various drop heights. In addition, force-displacement curves, deformed shapes, crack patterns, stresses and strains are given for various slab-design conditions and drop events. The utility of the results in design are illustrated through examples.

  5. 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

  6. 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.

  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. 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

  10. 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%

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-11

    In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-15

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

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

    Science.gov (United States)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

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

  16. 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.

  17. 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

  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. 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

  20. 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

  1. 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.

  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 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....

  5. 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...

  6. 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...

  7. 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

  8. A Stylistic Analysis on Edgar Allan Poe's The Cask of Amontillado

    Institute of Scientific and Technical Information of China (English)

    杨赛菲

    2016-01-01

    The Cask of Amontillado is one of Poe's best-known horror short stories. Based on Stylistics, this paper attempts to analyze this story from the aspects of themes, characterization, point of view, syntactic and lexical features, to reveal Poe's excellent skills and the artistic charm.

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. 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

  12. 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

  13. (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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  15. 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.

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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

  20. 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)

  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. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    International Nuclear Information System (INIS)

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

  6. Incentives for the use of depleted uranium alloys as transport cask containment structure

    International Nuclear Information System (INIS)

    Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both nominal use and accident conditions to serve the dual-role of shielding and containment, the use of other structure materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describesa two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature. The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracmm resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as win be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term ''structural credit'' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.)

  7. 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

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-30

    Alkali-silica reaction (ASR) is a deleterious chemical process that may occur in cement-based materials such as mortars and concretes, where the hydroxyl ions in the highly alkaline pore solution attack the siloxane groups in the siliceous minerals in the aggregates. The reaction produces a cross-linked alkali-silica gel. The ASR gel swells in the presence of water. Expansion of the gel results in cracking when the swelling-induced stress exceeds the fracture toughness of the concrete. As the ASR continues, cracks may grow and eventually coalesce, which results in reduced service life and a decrease safety of concrete structures. Since concrete is widely used as a critical structural component in dry cask storage of used nuclear fuels, ASR damage poses a significant threat to the sustainability of long term dry cask storage systems. Therefore, techniques for effectively detecting, managing and mitigating ASR damage are needed. Currently, there are no nondestructive methods to accurately detect ASR damage in existing concrete structures. The only current way of accurately assessing ASR damage is to drill a core from an existing structure, and conduct microscopy on this drilled cylindrical core. Clearly, such a practice is not applicable to dry cask storage systems. To meet these needs, this research is aimed at developing (1) a suite of nonlinear ultrasonic quantitative nondestructive evaluation (QNDE) techniques to characterize ASR damage, and (2) a physics-based model for ASR damage evolution using the QNDE data. Outcomes of this research will provide a nondestructive diagnostic tool to evaluate the extent of the ASR damage, and a prognostic tool to estimate the future reliability and safety of the concrete structures in dry cask storage systems

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

    International Nuclear Information System (INIS)

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

  11. Localization of cask and plug remote handling system in ITER using multiple video cameras

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Vale, Alberto [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Zimmer, A.; Koploy, M.

    1991-12-01

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

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

    International Nuclear Information System (INIS)

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

  17. Alternative Splicing of a Novel Inducible Exon Diversifies the CASK Guanylate Kinase Domain

    Directory of Open Access Journals (Sweden)

    Jill A. Dembowski

    2012-01-01

    Full Text Available Alternative pre-mRNA splicing has a major impact on cellular functions and development with the potential to fine-tune cellular localization, posttranslational modification, interaction properties, and expression levels of cognate proteins. The plasticity of regulation sets the stage for cells to adjust the relative levels of spliced mRNA isoforms in response to stress or stimulation. As part of an exon profiling analysis of mouse cortical neurons stimulated with high KCl to induce membrane depolarization, we detected a previously unrecognized exon (E24a of the CASK gene, which encodes for a conserved peptide insertion in the guanylate kinase interaction domain. Comparative sequence analysis shows that E24a appeared selectively in mammalian CASK genes as part of a >3,000 base pair intron insertion. We demonstrate that a combination of a naturally defective 5 splice site and negative regulation by several splicing factors, including SC35 (SRSF2 and ASF/SF2 (SRSF1, drives E24a skipping in most cell types. However, this negative regulation is countered with an observed increase in E24a inclusion after neuronal stimulation and NMDA receptor signaling. Taken together, E24a is typically a skipped exon, which awakens during neuronal stimulation with the potential to diversify the protein interaction properties of the CASK polypeptide.

  18. Application of a chemical ion exchange model to transport cask surface decontamination

    International Nuclear Information System (INIS)

    Radionuclide contamination of stainless steel surfaces occurs during submersion in a spent fuel storage pool, Subsequent release or desorption of these contaminants from a nuclear fuel transportation cask surface under varying environmental conditions occasionally results in the phenomenon known as contamination 'weeping'. Experiments have been conducted to determine the applicability of a chemical ion exchange model to characterise the problem of cask contamination and release. Surface charge characteristics of Cr2O3 and stainless steel (304) powders have been measured to determine the potential for ion exchange at metal oxide-aqueous interfaces. The solubility of Co and Cs electrolytes at varying pH and the absorption characteristics of these ions on Cr2O3 and stainless steel powders in aqueous slurries have been studied. Experiments show that Co ions do reversibly absorb on these powder surfaces and, more specifically, that absorption occurs in the nominal pH range (pH = 4-6) of a boric acid moderated spent fuel pool. Desorption has been demonstrated to occur at pH≤3. Cs+ ions also have been shown to have an affinity for these surfaces although the reversibility of Cs+ bonding by H+ ion exchange has not been fully demonstrated. These results have significant implications for effective decontamination and coating processes used on nuclear fuel transportation casks. (author)

  19. Application of a chemical ion exchange model to transport cask surface decontamination

    International Nuclear Information System (INIS)

    Radionuclide contamination of stainless steel surfaces occur during submersion in a spent fuel storage pool. Subsequent release or desorption of these contaminants from a nuclear fuel transportation cask surface under varying environmental conditions occasionally results in the phenomenon known as contamination ''weeping.'' Experiments have been conducted to determine the applicability of a chemical ion-exchange model to characterize the problem of cask contamination and release. Surface charge characteristics of Cr2O3 and stainless steel (304) powders have been measured to determine the potential for ion exchange at metal oxide -- aqueous interfaces. The solubility of Co and Cs electrolytes at varying pH and the absorption characteristics of these ions on Cr2O3 and stainless steel powders in aqueous slurries have been studied. Experiments show that Co ions do reversibly adsorb on these powder surfaces and, more specifically, that adsorption occurs in the nominal pH range (pH = 4--6) of a boric acid-moderated spent fuel pool. Desorption has been demonstrated to occur at pH ≤ 3. Cs ions also have been shown to have an affinity for these surfaces although the reversibility of Cs+ bonding by H+ ion exchange has not been fully demonstrated. These results have significant implications for effective decontamination and coating processes used on nuclear fuel transportation casks. 8 refs., 5 figs

  20. Flexible path optimization for the Cask and Plug Remote Handling System in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Vale, Alberto, E-mail: avale@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Fonte, Daniel; Valente, Filipe; Ferreira, João [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Gonzalez, Carmen [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: ► Complementary approach for path optimization named free roaming that takes full advantage of the rhombic like kinematics of the Cask and Plug Remote Handling System (CPRHS). ► Possibility to find trajectories not possible in the past using the line guidance developed in a previous work, in particular when moving the Cask Transfer System (CTS) beneath the pallet or in rescue missions. ► Methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. -- Abstract: The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed.

  1. Onsite storage of spent nuclear fuel in metalic spent fuel storage casks

    International Nuclear Information System (INIS)

    Virginia Electric and Power Company (Vepco) owns and operates two nuclear power stations within its system: the North Anna Power Station located in Louisa County, Virginia; and the Surry Power Station located in Surry County, Virginia. Each of these power stations has two pressurized water reactor operating units which share a common spent fuel pool. Under the Nuclear Waste Policy Act of 1982, Vepco is responsible for providing interim spent fuel storage until availability of the Federal Repository. Vepco has studied a number of options and has developed a program to provide the required onsite interim spent fuel storage. Options considered by Vepco included reracking, pin consolidation, dry storage and construction of a new spent fuel pool to provide the increased spent fuel storage capacity required. Vepco has selected reracking at North Anna combined with dry storage in metal spent fuel storage casks at Surrey to provide the required onsite spent fuel storage. A dry cask storage facility design and license application were developed and the license application was submitted to the NRC in October, 1982. The selection of the option to use dry cask storage of spent fuel at Surry represents the first attempt to license dry storage of spent nuclear fuel in the United States. This storage option is expected to provide an effective option for utilities without adequate storage space in their existing spent fuel pools

  2. Flexible path optimization for the Cask and Plug Remote Handling System in ITER

    International Nuclear Information System (INIS)

    Highlights: ► Complementary approach for path optimization named free roaming that takes full advantage of the rhombic like kinematics of the Cask and Plug Remote Handling System (CPRHS). ► Possibility to find trajectories not possible in the past using the line guidance developed in a previous work, in particular when moving the Cask Transfer System (CTS) beneath the pallet or in rescue missions. ► Methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. -- Abstract: The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed

  3. Localization of cask and plug remote handling system in ITER using multiple video cameras

    International Nuclear Information System (INIS)

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building

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

    International Nuclear Information System (INIS)

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

  5. Gamma-ray control of metal and concrete cask radiation protection

    International Nuclear Information System (INIS)

    Metal and concrete cask for durable storage and transportation of the spent fuel is equipped by the remote control device for verification of radiation protection in particular concrete defects. Operation tenet is irradiation of the cask wall by gamma-rays with an exposure rate at the surface chart design. Introduction In compliance with the requirements of national standards and regulations which are valid in nuclear power engineering and also IAEA recommendations, transportation packaging modules (TPM) for long-term storage and shipment of the spent nuclear fuel (SNF) have to ensure rated protection against ionizing radiation and withstand emergency impacts while preserving integrity of tightness system and radiation protection. Special Mechanical Engineering Design Office (SMEDO) has developed to manufacture such a module on the basis of metal and concrete cask (TPM MCC) for spent nuclear fuel of RBMK- 1000 reactors, NPS (nuclear-powered submarines) etc. In general, the structure of MCC may be presented as three coaxial steel shells the space between them being filled with high-density (4 and 3.5 g/cm3) concrete of high ductility and reinforced with composite grid of bars, clamps and rings. We have developed a procedure to control radiation protection (RP) of this cask, RP integrity checks after dynamic testing which simulate emergency situation during transportation. Test bench of γ-control was designed and constructed. The task included assessment of the cask radiation protection parameters, correlation of estimated and pilot data, cask manufacturing quality control as well as assessment of its body concrete filling uniformity and finally, confirmation of RP integrity under applied dynamic loads. RP γ-control method has been selected for this task-solving. This method is based on radiometry of cask walls by irradiation from a radioactive source. So successive investigation and gamma-ray flow detection of cask wall are being performed in order to determine

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Witte, M.

    2000-03-28

    Applicants seeking a Certificate of Compliance for an Independent Spent Fuel Storage Installation (ISFSI) cask must evaluate the consequences of a handling accident resulting in a drop or tip-over of the cask onto a concrete storage pad. As a result, analytical modeling approaches that might be used to evaluate the impact of cylindrical containers onto concrete pads are needed. One such approach, described and benchmarked in NUREG/CR-6608,{sup 1} consists of a dynamic finite element analysis using a concrete material model available in DYNA3D{sup 2} and in LS-DYNA,{sup 3} together with a method for post-processing the analysis results to calculate the deceleration of a solid steel billet when subjected to a drop or tip-over onto a concrete storage pad. The analysis approach described in NUREG/CR-6608 gives a good correlation of analysis and test results. The material model used for the concrete in the analyses in NUREG/CR-6608 is, however, somewhat troublesome to use, requiring a number of material constants which are difficult to obtain. Because of this a simpler approach, which adequately evaluates the impact of cylindrical containers onto concrete pads, is sought. Since finite element modeling of metals, and in particular carbon and stainless steel, is routinely and accurately accomplished with a number of finite element codes, the current task involves a literature search for and a discussion of available concrete models used in finite element codes. The goal is to find a balance between a concrete material model with a limited number of required material parameters which are readily obtainable, and a more complex model which is capable of accurately representing the complex behavior of the concrete storage pad under impact conditions. The purpose of this effort is to find the simplest possible way to analytically represent the storage cask deceleration during a cask tip-over or a cask drop onto a concrete storage pad. This report is divided into three sections

  9. New generation of CASTOR registered casks for high enriched, high burn-up fuel from German NPP

    International Nuclear Information System (INIS)

    Requirements for new cask designs for transport and long-term dry storage of spent fuel assemblies (FA) from LWR-reactors are based on both increased source terms of the LWR FA including MOX FA, as well as the demand for economical optimisation of decommissioning costs by increased cask capacities. For this, cask development is the challenge to create and establish cask designs that can accommodate more FA with higher source terms, each under fixed boundary conditions (i.e. transport requirements and limitations of the power plants as crane loads and/or fixed maximum dimensions). This task has been elaborated by working simultaneously on different development actions each focussed to improve the cask performance. In the following a brief summary will be presented to give an overview which developments and investigations have been and are still will be performed for development and safety analyses of the new CASTOR registered -designs under the main subjects: material investigation and qualification, component tests and verifications, detailed design analysis and not at least design verification

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Chatzidakis, S.; Choi, C. K.; Tsoukalas, L. H.

    2016-08-01

    The potential non-proliferation monitoring of spent nuclear fuel sealed in dry casks interacting continuously with the naturally generated cosmic ray muons is investigated. Treatments on the muon RMS scattering angle by Moliere, Rossi-Greisen, Highland and, Lynch-Dahl were analyzed and compared with simplified Monte Carlo simulations. The Lynch-Dahl expression has the lowest error and appears to be appropriate when performing conceptual calculations for high-Z, thick targets such as dry casks. The GEANT4 Monte Carlo code was used to simulate dry casks with various fuel loadings and scattering variance estimates for each case were obtained. The scattering variance estimation was shown to be unbiased and using Chebyshev's inequality, it was found that 106 muons will provide estimates of the scattering variances that are within 1% of the true value at a 99% confidence level. These estimates were used as reference values to calculate scattering distributions and evaluate the asymptotic behavior for small variations on fuel loading. It is shown that the scattering distributions between a fully loaded dry cask and one with a fuel assembly missing initially overlap significantly but their distance eventually increases with increasing number of muons. One missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the distributions which is the case of 100,000 muons. This indicates that the removal of a standard fuel assembly can be identified using muons providing that enough muons are collected. A Bayesian algorithm was developed to classify dry casks and provide a decision rule that minimizes the risk of making an incorrect decision. The algorithm performance was evaluated and the lower detection limit was determined.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

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

    International Nuclear Information System (INIS)

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

  17. Interface Issues Arising Between Storage and Transport for Storage Facilities Using Storage/Transport Dual Purpose Dry Metal Casks

    International Nuclear Information System (INIS)

    The dual purpose dry metal casks were developed as a low cost and reliable design to handle spent fuel safely, not only in relation to storage, but also transportation. One of its main advantages is to enhance worker protection against radiation while reducing possible direct manipulation of the spent fuel. In order to define regulation and the use of this type of casks, a traditional approach can be used, based on the study of every individual aspect. However a new type of approach is possible, called the “holistic approach”, taking into account the different aspects as a whole. (author)

  18. Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    2001-03-08

    The U.S. Nuclear Regulatory Commission (NRC) is currently reviewing the technical specifications for spent fuel storage casks in an effort to develop standard technical specifications (STS) that define the allowable spent nuclear fuel (SNF) contents. One of the objectives of the review is to minimize the level of detail in the STS that define the acceptable fuel types. To support this initiative, this study has been performed to identify potential fuel specification parameters needed for criticality safety and radiation shielding analysis and rank their importance relative to a potential compromise of the margin of safety.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1983-01-01

    This report details many of the interrelated considerations involved in optimizing large Pb, Fe, or U-metal spent fuel shipping casks containing 1, 2, 3, 5, 7, or 10-year-old PWR fuel assemblies. Scoping analyses based on criticality, shielding, and heat transfer considerations indicate that some casks may be able to hold as many as 18 to 21 ten-year-old PWR fuel assemblies. In the criticality section, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. In another section, details of many n/..gamma.. shielding optimization studies are presented, including the optimal n/..gamma.. design points and the actual shielding requirements for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. Multigroup source terms based on ORIGEN2 calculations at these and other decay times are also included. Lastly, the numerical methods and experimental correlations used in the steady-state and transient heat transfer analyses are fully documented, as are pertinent aspects of the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. (While only casks for square, intact PWR fuel assemblies were considered in this study, the SCOPE code may also be used to design and analyze casks containing canistered spent fuel or other waste material. An abbreviated input data guide is included as an appendix).

  2. A Study on the Radiation Source Effect to the Radiation Shielding Analysis for a Spent-Fuel Cask Design with Burnup-Credit

    International Nuclear Information System (INIS)

    The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source (60 Co radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

  3. Development of a conditioning system for the dual-purpose transport and storage cask for spent nuclear fuel from decommissioned Russian submarines

    International Nuclear Information System (INIS)

    Russia, stores large quantities of spent nuclear fuel (SNF) from submarine and ice-breaker nuclear powered naval vessels. This high-level radioactive material presents a significant threat to the Arctic and marine environments. Much of the SNF from decommissioned Russian nuclear submarines is stored either onboard the submarines or in floating storage vessels in Northwest and Far East Russia. Some of the SNF is damaged, stored in an unstable condition, or of a type that cannot currently be reprocessed. In many cases, the existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing all of this fuel from remote locations. Additional transport and storage options are required. Some of the existing storage facilities being used in Russia do not meet health and safety and physical security requirements. The U.S. has assisted Russia in the development of a new dual-purpose metal-concrete transport and storage cask (TUK-108/1) for their military SNF and assisted them in building several new facilities for off-loading submarine SNF and storing these TUK-108/1 casks. These efforts have reduced the technical, ecological, and security challenges for removal, handling, interim storage, and shipment of this submarine fuel. Currently, Russian licensing limits the storage period of the TUK-108/1 casks to no more than two years before the fuel must be shipped for reprocessing. In order to extend this licensed storage period, a system is required to condition the casks by removing residual water and creating an inert storage environment by backfilling the internal canisters with a noble gas such as argon. The U.S. has assisted Russia in the development of a mobile cask conditioning system for the TUK-108/1 cask. This new conditioning system allows the TUK 108/1 casks to be stored for up to five years after which the license may be considered for renewal for an additional five years or the fuel will be shipped to

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

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

    International Nuclear Information System (INIS)

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

  6. Improvement of Operational Safety of Dual-Purpose Cask for SNF in Storage

    International Nuclear Information System (INIS)

    By now more than 100 dual-purpose packaging sets TUK-108/1 are in operation in the mode of interim storage and transportation of SNF from decommissioned nuclear powered submarines (NPSs). In accordance with certificate, spent fuel is stored in TUK-108/1 on the premises of plants involved in NPS dismantlement for 2 years, whereupon it is transported for processing to PO Mayak. At one Far Eastern plant Zvezda involved in NPS dismantlement there arose a complicated situation due to necessity to extend period of storage of SNF in TUK-108/1. To ensure safety over a longer period of storage of SNF in TUK-108/1 it is essential to modify conditions of storage by removing of residual water and filling the inner cavity of the cask with an inert gas. This report describes key issues of cask drying technology, justification of terms of dry storage of naval SNF in TUK- 108/1 and RBMK-1000 SNF in UKhK-109, design features of the mobile drying facility, results of operation of the pilot facility at the Far Eastern plant Zvezda and cold testing on the shop factory for Leningrad NPP. (author)

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

    International Nuclear Information System (INIS)

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

  8. High capacity cask (TN28V) and International Transport System for the return shipment of vitrified high activity wastes

    International Nuclear Information System (INIS)

    The reprocessing of spent fuel generates different kinds of wastes. Among them fission products and non fissile actinides represent 98% of the radioactivity; these wastes are separated, concentrated, mixed with molten glass and poured into stainless steel containers. For political reasons, it is necessary to return these vitrified high activity wastes to the foreign countries which have decided to have their spent fuel reprocessed in France. So the transport of vitrified waste is vital for both the reprocessor and the utilities that have trusted the reprocessor and this operation has to be securely performed to give satisfaction to all concerned particles. For that reason Cogema will control the whole transport activity from La Hague plants to the receiving facilities of the customers. Therefore cogema will be responsible of the transport whatever the cask type (transport or storage) and will subcontract the transport operation to experienced companies such as Transnucleaire, PNTL or NTL, who will act on behalf of Cogema. Cogema will be the owner of the transport casks while the storage casks will normally be owned by the customers. Both cask types will of course have to comply with the requirements of La Hague, as published by Cogema

  9. 75 FR 57841 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective...

    Science.gov (United States)

    2010-09-23

    ... RIN 3150-AI88 List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of... 42292). This direct final rule amended the NRC's spent fuel storage regulations at 10 CFR 72.214 to... spent fuel assemblies (36 undamaged Exxon fuel assemblies and up to 32 damaged fuel cans (in...

  10. Comparative changes in color features and pigment composition of red wines aged in oak and cherry wood casks.

    Science.gov (United States)

    Chinnici, Fabio; Natali, Nadia; Sonni, Francesca; Bellachioma, Attilio; Riponi, Claudio

    2011-06-22

    The color features and the evolution of both the monomeric and the derived pigments of red wines aged in oak and cherry 225 L barriques have been investigated during a four months period. For cherry wood, the utilization of 1000 L casks was tested as well. The use of cherry casks resulted in a faster evolution of pigments with a rapid decline of monomeric anthocyanins and a quick augmentation formation of derived and polymeric compounds. At the end of the aging, wines stored in oak and cherry barriques lost, respectively, about 20% and 80% of the initial pigment amount, while in the 1000 L cherry casks, the same compounds diminished by about 60%. Ethyl-bridged adducts and vitisins were the main class of derivatives formed, representing up to 25% of the total pigment amount in the cherry aged samples. Color density augmented in both the oak and cherry wood aged samples, but the latter had the highest values of this parameter. Because of the highly oxidative behavior of the cherry barriques, the use of larger casks (e.g., 1000 L) is proposed in the case of prolonged aging times. PMID:21548629

  11. Storing the Spent Nuclear Fuel in Dry Casks Licensed for a Century as an Alternative to Recycling Solution

    Science.gov (United States)

    Milincic, Radovan

    2010-02-01

    Management of spent nuclear power reactor fuels is one of the most urgent problems in nuclear technology. Yearly production of new spent fuel is in the range of thousands of tons, topping a couple of hundred thousand tons of spent fuel already. This material is extremely radioactive and currently there is no adequate international policy, control or management regarding it. I propose here an intermediate term solution to this problem, which will be technologically and economically sustainable: interim spent-fuel storage as an alternative to reprocessing. The reprocessing inherently increases the net amount of the plutonium, which can be used for production of nuclear arms. Moreover, it is an expensive process with the net effect of producing different type of radioactive waste. In particular, the development of a dry cask for nuclear waste storage on site and transport, licensed for a period of hundred years would provide a significantly less expensive solution in the recent future, giving a needed relief to crowded spent-fuel storage pools. Currently in the U.S, NRC licenses existing storage casks for 20 years; and licenses for some of the dry cask storage facilities in the U.S. are about to expire. The extended life dry casks will provide sufficient intermediate period toward a more efficient and/or technologically advanced solution for spent fuel. )

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

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

  16. How will regional brewers and their cask beer brands survive and prosper in the face of market changes resulting from Government intervention?

    OpenAIRE

    Jespersen, Paul

    2007-01-01

    This project is a study of the current state of the cask beer market in England, focussing on one particular regional cask beer producer, "A". The project has been undertaken with two main objectives in mind. Firstly, the project sets out to determine the nature and extent of the impact of recent Government intervention in the cask beer market; in particular, the introduction in 2002 of Progressive Beer Duty, or Small Relief, and, in the same year, the revocation of the Beer Orders legis...

  17. Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report Project A.5 and A.6

    Energy Technology Data Exchange (ETDEWEB)

    ARD, K.E.

    2000-04-19

    This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01.

  18. Experimental Study on Influence of Mechanical Vibration during Transport of Transport/Storage Cask for Spent Nuclear Fuel on Containment Performance of Metal Gasket during Storage in Japan

    International Nuclear Information System (INIS)

    Transport casks of spent nuclear fuel will receive mechanical vibration during transport. It is known that the containment performance of metal gaskets is influenced by large external load or displacement. Quantitative influence of such vibration during transport on the containment performance of the metal gasket has not been known, but is crucial information particularly if the cask is stored as it is after the transport

  19. Interim storage of spent fuel assemblies from VVER-reactors, taking as an example the cask dry storage for the Czech Dukovany Nuclear Power Plant

    International Nuclear Information System (INIS)

    The nuclear fuel cycle services from the former Soviet Union were fundamentally changed in 1989. The necessity of intermediate spent fuel storage increased in Czechoslovakia in short term. After performing an international comparison and request for proposals, Czechoslovakia represented by the electrical utility CEZ in Prague, decided in favor of a dry cask storage concept for the nuclear power plant Dukovany. The selection process among the offered solutions and the dry cask storage concept is discussed

  20. Direct final disposal of transport and storage casks. A realizable technical concept

    International Nuclear Information System (INIS)

    GNS and DBE TEC developed possible alternatives and supplementary concepts to the existing German reference concept POLLUX and the concept of direct final disposal in boreholes (BSK3) the concept of direct final disposal of transport and storage casks (DIREGT). Advantages of this include the avoidance of necessary elaborate segmentation of fuel elements and core structures, the reduction of waste package transfers and standardized technical equipment for the final disposal engineering. The tasks to be studied include the adaptation of the shaft lifting to the high workload, the adaptation of the underground hauling to the high loads and the development of an appropriate storage technology, considerations concerning the safety with respect to criticality for the demonstration of long-term safety. The basic feasibility of the concept has been demonstrated, the work to be done concerns the demonstration of approvability of the concept for licensing purposes.

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

    International Nuclear Information System (INIS)

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the U.S. Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for a shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

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

    International Nuclear Information System (INIS)

    It has been postulated that a degradation phenomenon, referred to as ''hot cell rot'', may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ''Hot cell rot'' refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ''hot cell rot'' phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical

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

    International Nuclear Information System (INIS)

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

  5. Development of dual purpose (storage and transport) metal casks in Spain

    International Nuclear Information System (INIS)

    The Spanish Nuclear Program consists of nine nuclear power plants with an overall capacity of 7.4 GWe, representing around 38% of the electricity share. In 1982 the National Energy Plan established the open cycle as the strategy to be followed thereby halting the reprocessing option. As a consequence of this Plan, a state-owned company, ENRESA, responsible to the ministry of Industry and energy, was created in 1984 with the responsibility to manage all kinds of radioactive wastes. To comply with its responsibilities for the interim storage of spent fuel, ENRESA has designed a strategy based on the use of dual purpose metal casks for the initial demand of additional storage capacity

  6. Quality assurance, fabrication, and accompanying quality control of CASTOR {sup registered} transport and storage casks

    Energy Technology Data Exchange (ETDEWEB)

    Beverungen, M.; Laug, R. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2004-07-01

    In Germany, the Federal Institute for Materials Research and Testing (BAM, Berlin) acts as the competent authority for the approval of quality assurance measures of packages used in the transport of radioactive material. For this purpose, the German Federal Ministry of Transport issued the ''Technical Guideline on Measures for Quality Assurance (QM) and Quality Surveillance (QUe) for packages for transport of radioactive materials (TRV 006)''. Due to this guideline, every applicant for a cask license is requested to issue a written programme for design, fabrication, testing, documentation, operation, maintenance, and recurrent testing of packages. This programme has to be approved by BAM. Therefore, the company implemented and maintains a programme on basis of the TRV 006 and a Quality Management System on basis of the ISO 9001:2000 which cover all requirements of the guideline to ensure a controlled processes. The system is regularly assessed by BAM.

  7. Application of surface complexation modeling to the understanding of transportation cask weeping

    Energy Technology Data Exchange (ETDEWEB)

    Granstaff, V.E.; Chambers, W.B.

    1993-11-01

    A new application for surface complexation modeling is described. These models, which describe chemical equilibria among aqueous and adsorbed species, have typically been used for predicting groundwater transport of contaminants by modeling the natural adsorbents as various metal oxides. We have shown that this type of modeling can also be used to explain stainless steel surface contamination and decontamination mechanisms. Stainless steel transportation casks that are submerged in a spent fuel storage pool at nuclear power stations, can become contaminated with radionuclides such as {sup 137}CS, {sup 134}Cs, and {sup 60}Co. Subsequent release or desorption of these contaminants under varying environmental conditions occasionally results in the phenomenon known as ``cask weeping.`` We have postulated that contaminants in the storage pool adsorb onto the hydrous metal oxide surface of the passivated stainless steel and are subsequently released during transportation, due to varying environmental factors, such as humidity, road salt, dirt, and acid rain. It is well known that 304 stainless steel has a chromium enriched passive surface layer; thus its adsorption behavior should be similar to that of chromium oxide. Presented here are adsorption data for Co{sup +2} on Cr{sup 2}O{sup 3} which simulate the stainless steel surface contamination. These data are interpreted using electrostatic surface complexation models. The FITEQL computer program was used to obtain the electrostatic model constants from the experimental data. Because the concentrations of contaminants in the storage pool are too low to be measured accurately by conventional chemical analysis techniques, MINTEQA2 can be used, with the fitted constants, to extrapolate the equilibria to the low concentrations representative of storage pool water.

  8. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage

    International Nuclear Information System (INIS)

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

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

    International Nuclear Information System (INIS)

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

  10. CASTOR registered 1000/19. Shipping and storage cask for dry intermediate storage of spent fuel elements from the Temelin nuclear power station

    International Nuclear Information System (INIS)

    The CASTOR registered 1000/19-type cask newly developed by GNS Gesellschaft fuer Nuklear-Service mbH has been designed for shipment and dry storage of 19 spent fuel elements from the two VVER-1000 pressurized water reactors of the Temelin nuclear power station. The design and expert approval followed the internationally recognized shipping regulations of IAEA and the Czech rules and regulations for dry intermediate storage. One June 21, 2010, the Czech regulatory authority issued a type approval (integral transport and storage permit) valid for a period of 5 years for the casks of the new CASTOR registered 1000/19 line. As early as on September 9, 2010, which is less than 4 years after the delivery contract had been signed by CEZ and GNS in November 2006, the first cask of the CASTOR registered 1000/19 line was loaded with 19 spent fuel elements in only 5 days and emplaced in the newly built intermediate cask store. Loading, handling, in-plant transport, and emplacement of the cask made successful use of the extensive handling and management equipment furnished by GNS. After successful commissioning of the equipment and cold handling in both power plant units and in the intermediate cask store, successful loading and emplacement represents one of the most important milestones in the project. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Dreesen, K.; Goetze, H.G.; Holst, L. [GNS, D-45127 Essen (Germany); Gerstenberg, H.; Schreckenbach, K. [Technical University of Munich, D-85748 Garching (Germany)

    2000-07-01

    According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was <1.0 x 10{sup -5} Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

  12. Full-scale prototyping of the Hitachi dual-purpose metal cask and verification of its heat transfer characteristics

    International Nuclear Information System (INIS)

    Hitachi has been developing dual-purpose metal casks for transport and storage of spent nuclear fuels. The Hitachi cask, HDP69B can store 69 BWR fuel assemblies. The cask features are as follows. 1) The fuel basket is assembled mainly with plates of borated stainless steel. The plates are not welded, but cross-inserted into each other like the dividers in an egg carton. Since the borated stainless steel has relatively low heat conductivity, aluminum alloy plates are inserted along with some stainless steel plates to enhance heat removal ability. 2) Cured resin blocks are fitted into the inner shell of the cask for neutron shielding of the cask body. The resin blocks are surrounded by an aluminum casing which transfers heat of stored fuel from the inner shell to the outer shell of the cask. The block type shield structure eliminates the need for welding the heat transfer fins to the inner and outer shells. The weldless structures of the HDP69B lead to its enhanced manufacturability, but they complicate the heat transfer characteristics because there are gaps between such components as the aluminum casing and inner/outer shells. We carried out full-scale prototyping of the HDP69B and ran a heat transfer test using the prototype. The purposes of the heat transfer test were to check the heat removal ability of the HDP69B and to verify the safety analysis model for heat removal. Results of the heat transfer test and optimized analysis model for heat transfer characteristics of the HDP69B are the focus of this paper. The heat transfer test is summarized as follows. Sixty nine heaters simulating the shape and heat power of spent fuel assemblies were inserted into the fuel basket. After replacing the inner atmosphere with 0.1 MPa of helium, the heat transfer test was started. About 7 days were required to equilibrate the temperature distribution. The temperature at the center of the basket was 194 C. The results confirmed the HDP69B had sufficient heat removal ability. The

  13. Full-scale prototyping of the Hitachi dual-purpose metal cask and verification of its heat transfer characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, N.; Ishida, N.; Ootsuka, M.; Kamoshida, M.; Hiranuma, T.; Doumori, S.; Hoshikawa, T.; Shimizu, M.; Kashiwakura, J.; Hayashi, M. [Hitachi, Ltd., Hitachi (Japan)

    2004-07-01

    Hitachi has been developing dual-purpose metal casks for transport and storage of spent nuclear fuels. The Hitachi cask, HDP69B can store 69 BWR fuel assemblies. The cask features are as follows. 1) The fuel basket is assembled mainly with plates of borated stainless steel. The plates are not welded, but cross-inserted into each other like the dividers in an egg carton. Since the borated stainless steel has relatively low heat conductivity, aluminum alloy plates are inserted along with some stainless steel plates to enhance heat removal ability. 2) Cured resin blocks are fitted into the inner shell of the cask for neutron shielding of the cask body. The resin blocks are surrounded by an aluminum casing which transfers heat of stored fuel from the inner shell to the outer shell of the cask. The block type shield structure eliminates the need for welding the heat transfer fins to the inner and outer shells. The weldless structures of the HDP69B lead to its enhanced manufacturability, but they complicate the heat transfer characteristics because there are gaps between such components as the aluminum casing and inner/outer shells. We carried out full-scale prototyping of the HDP69B and ran a heat transfer test using the prototype. The purposes of the heat transfer test were to check the heat removal ability of the HDP69B and to verify the safety analysis model for heat removal. Results of the heat transfer test and optimized analysis model for heat transfer characteristics of the HDP69B are the focus of this paper. The heat transfer test is summarized as follows. Sixty nine heaters simulating the shape and heat power of spent fuel assemblies were inserted into the fuel basket. After replacing the inner atmosphere with 0.1 MPa of helium, the heat transfer test was started. About 7 days were required to equilibrate the temperature distribution. The temperature at the center of the basket was 194 C. The results confirmed the HDP69B had sufficient heat removal ability. The

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. The role of sensor directed, model-based control in robotic handling of nuclear waste casks and materials

    International Nuclear Information System (INIS)

    This paper discusses the results from several projects investigating the application of intelligent machine technologies to remote handling of nuclear waste casks and materials. The importance of computer models of the robot, its environment and their interactions is focused upon. Integration of such models into the sensor based control of robot systems results in significant increases in the capabilities of commercial robots by allowing tuning of robot performance to the task

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. Interface Issues Arising in Interim Storage Facilities Using Storage/Transport Dual Purpose Dry Metal Casks in Japan. Annex VIII

    International Nuclear Information System (INIS)

    The annual amount of spent fuels (SFs) discharged by the operation of commercial reactors nowadays is estimated to be around 10 000 tU level worldwide. While the amount of SFs already reprocessed account about one-third, the rest are currently stored in storage facilities, typically, in wet pools attached to nuclear power plants (NPPs). Cumulative amount of SFs stored is estimated to be about 250 000 tU by 2010 (I. Hanaki, Japan). While wet pool system is dominant in storage facility designs, new design concepts for storage facilities have been continuously developed. One of these new designs is that using dual purpose dry metal casks. “Dual” here means that the casks are not only designed as storage containers, but also designed as transport containers that will satisfy relevant regulatory requirements for transport of radioactive materials such as TS-R-1. Advantage of adopting such “dual” design in storage facilities lies in that this could contribute to reduce the burden associated with handling operations, because, under such designs, SFs once loaded into casks can easily be “transported” to storage facilities, and after storage of several decades, they can again be “transported” to their destinations, regardless they are reprocessing facilities or final disposal sites. Other than these, adopting this kind of design can reduce the amount of radioactive wastes discharged through storage operation, thus can reduce operation costs while maintaining safety level. In Japan, where 53 commercial NPPs are now in operation and with the annual amount of SFs produced sums up to about 1000 tU, keen needs are perceived among SFs producers (namely, utilities) to secure adequate SFs storage capacity. Therefore, a new application for constructing storage facility of 3000 tU scale in Mutsu city, located in northern part of Aomori prefecture, has been submitted in March 2007 by a subsidiary company of utilities named RFS (Recyclable Fuel Storage Company), using

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

    International Nuclear Information System (INIS)

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  4. Large deformation inelastic analysis of impact for shipping casks. [DYNA3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Charman, C.M.; Grenier, R.M. (General Atomic Co., San Diego, CA (USA)); Nickell, R.E. (Applied Science and Technology, Poway, CA (USA))

    1982-09-01

    This paper describes the use of two- and three-dimensional nonlinear finite element computer programs to design a radioative material transportation cask to withstand a drop of 30 feet onto an unyielding surface. Because of recent advancement in the area of non-linear finite element code development, the use of such codes for an iterative design process is becoming practicable. The paper begins with a section dealing with a two-dimensional side drop analysis and is followed by a discussion of the general capabilities of DYNA3D and a brief discussion of the implementation of the code on a computational mainframe unlike any for which the developer had intended. Then, a section on three-dimensional models of center-of-gravity over a corner impact follows, which introduces design features such as bolted closures, internal impact limiter, seals and shear rings. Figs. showing the deformed model grids are included. Stress and strain results are given in the subsequent section. Finally, we interpret these results in terms of possible rules being developed by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code committees.

  5. Acoustic emission detection with fiber optical sensors for dry cask storage health monitoring

    Science.gov (United States)

    Lin, Bin; Bao, Jingjing; Yu, Lingyu; Giurgiutiu, Victor

    2016-04-01

    The increasing number, size, and complexity of nuclear facilities deployed worldwide are increasing the need to maintain readiness and develop innovative sensing materials to monitor important to safety structures (ITS). In the past two decades, an extensive sensor technology development has been used for structural health monitoring (SHM). Technologies for the diagnosis and prognosis of a nuclear system, such as dry cask storage system (DCSS), can improve verification of the health of the structure that can eventually reduce the likelihood of inadvertently failure of a component. Fiber optical sensors have emerged as one of the major SHM technologies developed particularly for temperature and strain measurements. This paper presents the development of optical equipment that is suitable for ultrasonic guided wave detection for active SHM in the MHz range. An experimental study of using fiber Bragg grating (FBG) as acoustic emission (AE) sensors was performed on steel blocks. FBG have the advantage of being durable, lightweight, and easily embeddable into composite structures as well as being immune to electromagnetic interference and optically multiplexed. The temperature effect on the FBG sensors was also studied. A multi-channel FBG system was developed and compared with piezoelectric based AE system. The paper ends with conclusions and suggestions for further work.

  6. Effects of temperature on concrete cask in a dry storage facility for spent nuclear fuels

    International Nuclear Information System (INIS)

    In the dry storage of spent nuclear fuels,concrete cask serves both as a shielding and a structural containment. The concrete in the storage facility is expected to endure the decay heat of the spent nuclear fuel during its service life. Thus, effects of the sustaining high temperature on concrete material need be evaluated for safety of the dry storage facility. In this paper, we report an experimental program aimed at investigating possible high temperature effects on properties of concrete, with emphasis on the mechanical stability, porosity,and crack-resisting ability of concrete mixes prepared using various amounts of Portland cement, fly ash, and blast furnace slag. The experimental results obtained from concrete specimens exposed to a temperature of 94 degree C for 90 days indicate that: (1) compressive strength of the concrete remains practically unchanged; (2) the ultrasonic pulse velocity, and dynamic modulus of elasticity of the concrete decrease in early stage of the high-temperature exposure,and gradually become stable with continuing exposure; (3) shrinkage of concrete mixes exhibits an increase in early stage of the exposure and does not decrease further with time; (4) concrete mixes containing pozzolanic materials,including fly ash and blast furnace slag, show better temperature-resisting characteristics than those using only Portland cement. (authors)

  7. Stress corrosion cracking of stainless-steel canister for concrete cask storage of spent fuel

    Science.gov (United States)

    Tani, Jun-ichi; Mayuzumi, Masami; Hara, Nobuyoshi

    2008-09-01

    Resistance to external stress corrosion cracking (ESCC) and crevice corrosion were examined for various candidate canister materials in the spent fuel dry storage condition using concrete casks. A constant load ESCC test was conducted on the candidate materials in air after deposition of simulated sea salt particles on the specimen gage section. Highly corrosion resistant stainless steels (SS), S31260 and S31254, did not fail for more than 46 000 h at 353 K with relative humidity of 35%, although the normal stainless steel, S30403 SS failed within 500 h by ESCC. Crevice corrosion potentials of S31260 and S31254 SS became larger than 0.9 V (SCE) in synthetic sea water at temperatures below 298 K, while those of S30403 and S31603 SS were less than 0 V (SCE) at the same temperature range. No rust was found on S31260 and S31254 SS specimens at temperatures below 298 K in the atmospheric corrosion test, which is consistent with the temperature dependency of crevice corrosion potential. From the test result, the critical temperature of atmospheric corrosion was estimated to be 293 K for both S31260 and S31254 SS. Utilizing the ESCC test result and the critical temperature, together with the weather station data and the estimated canister wall temperature, the integrity of canister was assessed from the view point of ESCC.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

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

  10. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  11. Calculative activation analysis of the transport rack for CASTOR {sup registered} casks; Berechnung der Aktivierung eines Transportgestells fuer CASTOR {sup registered} -Behaelter

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany); Biedermann, R. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Schmidt-Wohlfarth, Y.; Louia, A. [EnBW Kernkraft GmbH, Philippsburg (Germany)

    2011-07-01

    The transport rack for the internal transport of loaded CASTOR {sup registered} casks before the storage in the intermediate storage facility at the site of the NPP Philippsburg is exposed to neutron irradiation from the cask inventory. Using the Monte Carlo code MCNP the activation rates of the transport rack materials are calculated for typical storage times of the casks in the rack. The long-term activation was also calculated for the continuous use of the transport rack over 10 years. Further topics were the dose rate in the near surrounding of the transport rack after long-term activation and finally the disposability of rack components according to the legal regulations. The maximum contact dose rate was calculated to be below 1 micro Sv/h after 10 years of application. The transport rack can be disposed with large safety margins to the radiation protection limits.

  12. CAPSIZE: A personal computer program and cross-section library for determining the shielding requirements, size, and capacity of shipping casks subject to various proposed objectives

    International Nuclear Information System (INIS)

    A new interactive program called CAPSIZE has been written for the IBM-PC to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel shipping casks designed to meet those objectives. Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the loaded cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium-shielded cask meeting those objectives. The necessary neutron and gamma shield thicknesses are determined by the program in such a way as to meet the specified external dose rate while simultaneously minimizing the overall weight of the loaded cask. The one-group cross-section library used in the CAPSIZE program has been distilled from the intermediate results of several hundred 1-D multigroaup discrete ordinates calculations for different types of casks. Neutron and gamma source terms, as well as the decay heat terms, are based on ORIGEN-S analyses of PWR fuel assemblies having exposures of 10, 20, 30, 40, 50, and 60 gigawatt days per metric tonne of initial heavy metal (GWD/MTIHM). In each case, values have been tabulated at 17 different decay times between 120 days and 25 years. Other features of the CAPSIZE program include a steady-state heat transfer calculation which will minimize the size and weight of external cooling fins, if and when such fins are required. Comparisons with previously reported results show that the CAPSIZE program can generally estimate the necessary neutron and gamma shield thicknesses to within 0.16 in. and 0.08 in., respectively. The corresponding cask weights have generally been found to be within 1000 lbs of previously reported results. 13 refs., 20 figs., 54 tabs

  13. CAPSIZE: A personal computer program and cross-section library for determining the shielding requirements, size, and capacity of shipping casks subject to various proposed objectives

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1987-05-01

    A new interactive program called CAPSIZE has been written for the IBM-PC to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel shipping casks designed to meet those objectives. Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the loaded cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium-shielded cask meeting those objectives. The necessary neutron and gamma shield thicknesses are determined by the program in such a way as to meet the specified external dose rate while simultaneously minimizing the overall weight of the loaded cask. The one-group cross-section library used in the CAPSIZE program has been distilled from the intermediate results of several hundred 1-D multigroaup discrete ordinates calculations for different types of casks. Neutron and gamma source terms, as well as the decay heat terms, are based on ORIGEN-S analyses of PWR fuel assemblies having exposures of 10, 20, 30, 40, 50, and 60 gigawatt days per metric tonne of initial heavy metal (GWD/MTIHM). In each case, values have been tabulated at 17 different decay times between 120 days and 25 years. Other features of the CAPSIZE program include a steady-state heat transfer calculation which will minimize the size and weight of external cooling fins, if and when such fins are required. Comparisons with previously reported results show that the CAPSIZE program can generally estimate the necessary neutron and gamma shield thicknesses to within 0.16 in. and 0.08 in., respectively. The corresponding cask weights have generally been found to be within 1000 lbs of previously reported results. 13 refs., 20 figs., 54 tabs.

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

    International Nuclear Information System (INIS)

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

  15. First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

    International Nuclear Information System (INIS)

    The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C-E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA-AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Huerta, M.

    1981-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

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

  20. Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage

    Energy Technology Data Exchange (ETDEWEB)

    Tsoukalas, Lefteri H. [Purdue Univ., West Lafayette, IN (United States)

    2016-03-01

    This is the final report of the NEUP project “Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage”, DE-NE0000695. The project started on December 1, 2013 and this report covers the period December 1, 2013 through November 30, 2015. The project was successfully completed and this report provides an overview of the main achievements, results and findings throughout the duration of the project. Additional details can be found in the main body of this report and on the individual Quarterly Reports and associated Deliverables of the project, uploaded in PICS-NE.

  1. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    Energy Technology Data Exchange (ETDEWEB)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  2. Nonlinear dynamic impact analysis for installing a dry storage canister into a vertical concrete cask

    International Nuclear Information System (INIS)

    In this paper, a series of dynamic impact analysis for installing a dry storage canister into a vertical concrete cask (VCC) is performed. The dry storage system considered herein is called HCDSS-69, recently developed by INER and being capable of accommodating 69 bundles of BWR spent nuclear fuels. The impact accident is stemming from a conservative consideration of accidental movement when the canister is being hoisted into a VCC. According to NUREG-0554, the accidental movement is conservatively simulated by 80 mm- and 160 mm-height free-drop motions and then with straight and 2°-oblique impact to a pedestal in VCC. A symmetric fully 3-D finite element model is built and analyzed using the explicit finite element code, LS-DYNA. Geometrical, contact, and material nonlinearities are all taken into account. The analysis result concludes that the permanent deformations of the canister are not severe to affect fuel retrieve after the impact accident and the maximum stress intensity in the canister shell can meet the ASME code appendix F F-1340, preventing the leakage of radioactive materials. The study also found that with properly reducing the wall thickness of the pedestal cylinder, the maximum acceleration and permanent deformation of the canister can be much alleviated, even though the drop height is increased to the double of the required brake distance specified in NUREG-0554. The damages of the pedestal in each analysis are moderate so that the heat transfer condition after the impact accident can be bounded by the off-normal event for half-blockage of air inlets

  3. Developing a structural health monitoring system for nuclear dry cask storage canister

    Science.gov (United States)

    Sun, Xiaoyi; Lin, Bin; Bao, Jingjing; Giurgiutiu, Victor; Knight, Travis; Lam, Poh-Sang; Yu, Lingyu

    2015-03-01

    Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. In total, there are over 1482 dry cask storage system (DCSS) in use at US plants, storing 57,807 fuel assemblies. Nondestructive material condition monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health", and more importantly, to guarantee the safe operation of radioactive waste storage systems (RWSS) during their extended usage period. A state-of-the-art nuclear structural health monitoring (N-SHM) system based on in-situ sensing technologies that monitor material degradation and aging for nuclear spent fuel DCSS and similar structures is being developed. The N-SHM technology uses permanently installed low-profile piezoelectric wafer sensors to perform long-term health monitoring by strategically using a combined impedance (EMIS), acoustic emission (AE), and guided ultrasonic wave (GUW) approach, called "multimode sensing", which is conducted by the same network of installed sensors activated in a variety of ways. The system will detect AE events resulting from crack (case for study in this project) and evaluate the damage evolution; when significant AE is detected, the sensor network will switch to the GUW mode to perform damage localization, and quantification as well as probe "hot spots" that are prone to damage for material degradation evaluation using EMIS approach. The N-SHM is expected to eventually provide a systematic methodology for assessing and monitoring nuclear waste storage systems without incurring human radiation exposure.

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

    International Nuclear Information System (INIS)

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

  5. Feedback on the use of the MX6 MOX fuel transport cask: reduction of the doses during operation

    Energy Technology Data Exchange (ETDEWEB)

    Thierry Lallemant; Eric Pernice [COGEMA LOGISTICS, AREVA Group (France); Roland Wagner [RWE, Gundremmingen Kernkraftwerk (Germany); Gerhard Dietl [E.ON, ISAR Kernkraftwerk (Germany)

    2006-07-01

    Full text of publication follows: The MX6 cask was developed by COGEMA LOGISTICS for the transport of either BWR or PWR fresh MOX fuel assemblies. Replacing the previous SIEMENS type III and SIEMENS BWR packaging, the MX6 has been firstly used for the German Nuclear Power Plants. Complying with the TS-R-1 (IAEA 1996) regulations, the MX6 cask is based on innovative solutions implemented at each step of the design and the manufacturing. Its design includes an efficient neutron shielding for high Plutonium content and an easy use content restraining system. The large payload of the MX6, 6 PWR MOX fuel assemblies or 16 BWR MOX fuel assemblies, contributes to the optimisation of the doses uptake during unloading in the NPP. Moreover, new sequences of loading and unloading operations were proposed for testing and implementation in each Nuclear Facility. Concurrently, total dose uptake by the operators was assessed in order to prove the efficiency of the packaging and the proposed sequences. In this paper, the main contributors to the transports to Germany with the MX6 will present their involvement and feedback for the reduction of the dose uptake by the operators during the loading and unloading operations. Mainly used at Gundremmingen and Isar Nuclear Power Plants (NPPs), the use of the MX6 will be extended to other German and Swiss NPPs from 2006. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  7. Performance of the improved version of Monte Carlo Code A{sup 3}MCNP for cask shielding design

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, T. [Mitsubishi Heavy Industries, Yokohama (Japan); Ueki, K. [Tokai Univ., Kanagawa (Japan); Sato, O. [Mitsubishi Research Inst., Tokyo (Japan); Sjoden, G.E. [Dept. of Nuclear and Radiological Engineering, Univ. of Florida, Gainesville, FL (United States); Miyake, Y.; Ohmura, M.; Haghighat, A.

    2004-07-01

    A{sup 3}MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A{sup 3}MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A{sup 3}MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A{sup 3}MCNP (referred to as A{sup 3}MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A{sup 3}MCNPV for cask neutron and gamma-ray shielding problem.

  8. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Gonzalez, Jesus S. [Univ. Politecnica de Madrid (Spain); Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William BJ J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  11. Calculation methods for demonstration of cladding tube integrity during dry long-term interim storage of fuel elements in CASTOR registered V casks

    International Nuclear Information System (INIS)

    The interim storage of spent fuel elements at the plant site of German nuclear power plants is using CASTOR registered V casks. In the frame of inventory extension fuel elements with higher burn-up are stored that result in higher decay heat und higher EOL pressures. For demonstration of heat removal of the CASTOR registered V cask during the interim storage it has to be shown that the reached temperatures comply with the limiting values to ensure the safety objective. To exclude cladding tuve failure two criteria have to be fulfilled during the interim storage: (A) the maximum tangential stress in the cladding must not surmount 120 MPa. (B) the persistent tangential strain at the cladding at the end of the storage time must not surmount 1%. The authors describe the calculation methodology for the temperatures in the casks taking into account the heat removal from the cask surface by radiant heat transfer and convection, the maximum allowable internal pressure and the resulting tangential strain in the cladding tube.

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

    International Nuclear Information System (INIS)

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

  13. Spectrum of pontocerebellar hypoplasia in 13 girls and boys with CASK mutations: confirmation of a recognizable phenotype and first description of a male mosaic patient

    Directory of Open Access Journals (Sweden)

    Burglen Lydie

    2012-03-01

    Full Text Available Abstract Background Pontocerebellar hypoplasia (PCH is a heterogeneous group of diseases characterized by lack of development and/or early neurodegeneration of cerebellum and brainstem. According to clinical features, seven subtypes of PCH have been described, PCH type 2 related to TSEN54 mutations being the most frequent. PCH is most often autosomal recessive though de novo anomalies in the X-linked gene CASK have recently been identified in patients, mostly females, presenting with intellectual disability, microcephaly and PCH (MICPCH. Methods Fourteen patients (12 females and two males; aged 16 months-14 years presenting with PCH at neuroimaging and with clinical characteristics unsuggestive of PCH1 or PCH2 were included. The CASK gene screening was performed using Array-CGH and sequencing. Clinical and neuroradiological features were collected. Results We observed a high frequency of patients with a CASK mutation (13/14. Ten patients (8 girls and 2 boys had intragenic mutations and three female patients had a Xp11.4 submicroscopic deletion including the CASK gene. All were de novo mutations. Phenotype was variable in severity but highly similar among the 11 girls and was characterized by psychomotor retardation, severe intellectual disability, progressive microcephaly, dystonia, mild dysmorphism, and scoliosis. Other signs were frequently associated, such as growth retardation, ophthalmologic anomalies (glaucoma, megalocornea and optic atrophy, deafness and epilepsy. As expected in an X-linked disease manifesting mainly in females, the boy hemizygous for a splice mutation had a very severe phenotype with nearly no development and refractory epilepsy. We described a mild phenotype in a boy with a mosaic truncating mutation. We found some degree of correlation between severity of the vermis hypoplasia and clinical phenotype. Conclusion This study describes a new series of PCH female patients with CASK inactivating mutations and confirms that

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

    Energy Technology Data Exchange (ETDEWEB)

    CHASTAIN, S.A.

    2005-10-24

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

  15. Spent Nuclear Fuel Cask and Storage Monitoring with 4He Scintillation Fast Neutron Detectors

    International Nuclear Information System (INIS)

    review and conceptual design of a new system for the monitoring of spent nuclear fuel pool and cask storage. Previous experimental results using 4He scintillator detectors have demonstrated the potential for these new detectors to be implemented in a spent fuel monitoring role. Their ability to preserve neutron energy information, while at the same time reject interfering gamma radiation, provides a new method of signal analysis. By applying energy cuts as proposed above, the spent fuel's fission signal can be isolated, and an algorithm for assembly identification developed

  16. Recycle Experience of Dismantled Cask Handling Crane by Surface Removal Sampling at Kori Unit No.1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. D.; Baeg, C. Y.; Son, J. K.; Kim, H. S.; Ha, J. A.; Song, M. J.

    2002-02-25

    The Kori No.1, which began operation in 1978, replaced its cask handling crane in 2000. To prove the safety of recycling and reuse of crane scrap, a particular calculation method for surface contamination was used. Because surface radioactive contamination of steel is limited to a few-microns-thick layer, we can calculate the total(removable and fixed contamination) activity of the sample conservatively by this surface removal sampling means. If we multiply the ratio of total surface and the area of the selected surface by its activity, total activity of the scrap can be estimated. Conservatively, the sampled portion can be used as a representative sample of the scrap. Both the inner and outer part of the scrap was sampled separately, and gamma spectra were analyzed to check whether activation had occurred. Before sampling, the entire surface of the steel is scan surveyed by several kinds of GM and GP detectors. Contaminated parts were segregated, or decontaminated to the background. Almost one sample per one ton of steel was collected. Gamma spectra of 62 samples were analyzed by 100% efficiency HP Ge detector. Only 60Co was detected, and its highest activity was 0.01 Bq/g,. This level of activity is much lower than the ''clearance levels'' outlined in IAEA TecDoc-855.(4). The total alpha and total beta for 6 samples were measured in the laboratory by low background alpha, using a beta gas proportional counter. Activities were much lower than 0.005 Bq/g. A representative sample was taken from the complete mixture of 62 samples. Gamma activities of nuclides were measured to estimate the dose to the public. This study revealed that activities of nuclides were lower than 'clearance levels' if decontaminated until the lower limit of detection level of the portable field instrument. New surface removal sampling method was tested. This method allows us to easily calculate the specific activity for the solid material.

  17. A methodology to quantify the release of spent nuclear fuel from dry casks during security-related scenarios.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Luna, Robert Earl

    2013-11-01

    Assessing the risk to the public and the environment from a release of radioactive material produced by accidental or purposeful forces/environments is an important aspect of the regulatory process in many facets of the nuclear industry. In particular, the transport and storage of radioactive materials is of particular concern to the public, especially with regard to potential sabotage acts that might be undertaken by terror groups to cause injuries, panic, and/or economic consequences to a nation. For many such postulated attacks, no breach in the robust cask or storage module containment is expected to occur. However, there exists evidence that some hypothetical attack modes can penetrate and cause a release of radioactive material. This report is intended as an unclassified overview of the methodology for release estimation as well as a guide to useful resource data from unclassified sources and relevant analysis methods for the estimation process.

  18. Progress on the interface between UPP and CPRHS (Cask and Plug Remote Handling System) tractor/gripping tool for ITER

    International Nuclear Information System (INIS)

    Highlights: ► UPP interface requirements in the plug RH extraction/insertion for ITER. ► Analyze of maximum misalignment between port duct and port cell. ► Friction study between plug skids and VV port/ramp rails during the plug transfer. ► Definition of the tolerance in the plug skids to avoid the plug jamming. ► Concepts of gripping tools based on one gripping point and avoiding force feedback. -- Abstract: EFDA finances a training programme called Goal Oriented Training Programme for Remote Handling (GOT RH), whose goal is to train engineers in Remote Handling for ITER. As part of this training programme, the conceptual design of the mechanical interface between Upper Port Plug (UPP) and Cask and Plug Remote Handling System (CPRHS) as well as the conceptual design of the needed tools for UPP Remote Handling is carried out. The paper presents the conceptual design of the UPP/Gripping Tool Interface. This includes the conceptual design of the gripping tool for introducing/removing the UPP in/from the ITER port and the mechanical features on both sides of the UPP/Gripping Tool Interface (e.g. alignment features, mechanical connectors, fasteners). In order to develop the design of the interface between UPP and CPRHS it is necessary to first identify the functional requirements of the Transfer Cask System (TCS) and the CPRHS, such as required degrees of freedom (DoF), required performances of system, geometrical constraints, loading conditions, alignment requirements, RAMI requirements. These requirements are the input data for the design of the interface between UPP and gripping tool and some of them are also described in the paper

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

    International Nuclear Information System (INIS)

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

  20. Direct disposal of transport an storage casks - status of the actual considerations; Direkte Endlagerung von Transport- und Lagerbehaeltern. Stand der konzeptionellen Ueberlegungen

    Energy Technology Data Exchange (ETDEWEB)

    Graf, Reinhold; Brammer, Klaus-Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Filbert, Wolfgang [DBE Technology GmbH, Peine (Germany)

    2011-07-01

    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 {sup 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.

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

    International Nuclear Information System (INIS)

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

  2. Study on commercial realization of concrete cask for interim storage of spent nuclear fuel. Proposal of examination methodology for stainless steel canister lid weldment

    International Nuclear Information System (INIS)

    To realize the early utilization of economical and module storage methodology, the Japan Society of Mechanical Engineers (JSME) had already developed and issued a Code for Construction of SNF Facilities - Rules on Concrete Cask - (JSME S FB1-2003) in 2003. On the other hand, the former competent authority (Nuclear and Industrial Safety Agency) issued the safety requirements for concrete cask design in 2004. For confinement requirements, Ultrasonic Test (UT) should be used to inspect the lid weldment in addition to Dye Penetrant Test (PT). To contribute to the JSME codification activities for adjustment between the safety requirements and the JSME design code, we performed mock-up canister weldment tests considering the vapor from the hot water in the spent fuel canister, the UT tests with canister lid weldment at high temperature and the structural analysis of the canister under hypothetical drop conditions. As a result, the draft amendment of the JSME design code was proposed. (author)

  3. CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.

    2014-06-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  4. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-31

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  5. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  6. Numerical and Experimental Investigations of Polyurethane Foam for Use as Cask Impact Limiter in Accidental Drop Scenarios - 12099

    International Nuclear Information System (INIS)

    Rigid, closed-cell polyurethane foams are frequently used as cask impact limiters in nuclear materials and hazardous waste transport due to their high energy-absorption potential. When assessing the cask integrity in accidental scenarios based on numerical simulations, a description of the foam damping properties is required for different strain rates and for a wide temperature range with respect to waste heat generation in conjunction with critical operating and environmental conditions. Implementation and adaption of a respective finite element material model strongly relies on an appropriate experimental data base. Even though extensive impact experiments were conducted e.g. in Sandia National Laboratories, Savannah River National Laboratory and by Rolls Royce plc, not all relevant factors were taken into account. Hence, BAM who is in charge of the mechanical evaluation of such packages within the approval procedure in Germany, incorporated systematic test series into a comprehensive research project aimed to develop numerical methods for a couple of damping materials. In a first step, displacement driven compression tests have been performed on confined, cubic specimens at five loading rates ranging from 0.02 mm/s to 3 m/s at temperatures between +90 deg. C and -40 deg. C. Materials include two different polyurethane foam types called FR3718 and FR3730 having densities of 280 kg/m3 and 488 kg/m3 from the product line-up of General Plastics Manufacturing Company. Their data was used to adapt an advanced plasticity model allowing for reliably simulating cellular materials under multi-axial compression states. Therefore, an automated parameter identification procedure had been established by combining an artificial neural network with local optimization techniques. Currently, the selected numerical material input values are validated and optimized by means of more complex loading configurations with the prospect of establishing methods applicable to impact

  7. Experience Of Using Metal-and-Concrete Cask TUK-108/1 For Storage And Transportation Of Spent Nuclear Fuel Of Decommissioned NPS

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, E.; Dyer, R. [Environmental Protection Agency, Ronald Reagan Bldg. 3rd Floor 1200 Pennsylvania Av., NW Washington, D.C. 20024 (United States); Snipes, R. [Oak Ridge National Laboratories, VA (United States); Dolbenkov, V.G.; Guskov, V.D.; Korotkov, G.V. [Joint Stock Company ' KBSM' , 64 Lesnoy Av., St.Petersburg 194100 (Russian Federation); Makarchuk, T.F. [Joint Stock Company ' Atomstroyexport' , Potapovskiy str. 5, bld. 4, Moscow, 101990 (Russian Federation); Zakharchev, A.A. [State Corporation ' Rosatom' , 24-26 Ordinka St., Moscow, 100000 (Russian Federation)

    2009-06-15

    In past 10 years in Russia an intensive development of a new technology of management of spent nuclear fuel (SNF) has taken place. This technology is based on the concept of using a shielded cask which provides safety of its content (SNF) and meeting all other safety requirements to storage and transportation of SNF. Radiation protection against emission and non-propagation of activity outside the cask is ensured by the physical barriers such as all-metal or composite body, face work, inner structures to accommodate spent fuel assemblies (SFA), lids with sealing systems. Residual heat buildup is off-taken to the environment by natural way: emission and convection of surrounding air. The necessity in development of the cask technology of SNF management was conditioned by the situation at hand with defueling of Russian decommissioned nuclear-powered submarines (NPS) as the existed transport infrastructure and enterprises involved in fuel processing could not meet the demand for transportation and processing of SNF neither from reactors of all dismantled NPS, nor from reactors of NPS waiting for decommissioning. The US and Norway actively participated in the trilateral joint project with the Russian Federation aimed at creation of a cask prototype for interim storage and transportation of SNF of dismantled NPS. The 1.1 Project is a part of the Arctic Military Environmental Cooperation (AMEC) Program. In December 2000 the project was successfully completed by issuance of the certificate-permit for design and transportation of NP Submarine SNF. It was a first certified dual-purpose TUK from the MMC family. In these years 106 TUK-108/1 casks have been manufactured and supplied to PO Mayak, JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. The storage pads for interim storage of TUK-108/1 have been built and currently are in operation on sites of SNF unloading from submarine reactors and SNF cask-loading such as JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. In

  8. Development of a novel ultrasonic temperature probe for long-term monitoring of dry cask storage systems

    International Nuclear Information System (INIS)

    With the recent cancellation of the Yucca Mountain repository and the limited availability of wet storage utilities for spent nuclear fuel (SNF), more attention has been directed toward dry cask storage systems (DCSSs) for long-term storage of SNF. Consequently, more stringent guidelines have been issued for the aging management of dry storage facilities that necessitate monitoring of the conditions of DCSSs. Continuous health monitoring of DCSSs based on temperature variations is one viable method for assessing the integrity of the system. In the present work, a novel ultrasonic temperature probe (UTP) is being tested for long-term online temperature monitoring of DCSSs. Its performance was evaluated and compared with type N thermocouple (NTC) and resistance temperature detector (RTD) using a small-scale dry storage canister mockup. Our preliminary results demonstrate that the UTP system developed at Argonne is able to achieve better than 0.8 °C accuracy, tested at temperatures of up to 400 °C. The temperature resolution is limited only by the sampling rate of the current system. The flexibility of the probe allows conforming to complex geometries thus making the sensor particularly suited to measurement scenarios where access is limited.

  9. Exome sequencing identified mutations in CASK and MYBPC3 as the cause of a complex dilated cardiomyopathy phenotype.

    Science.gov (United States)

    Reinstein, Eyal; Tzur, Shay; Bormans, Concetta; Behar, Doron M

    2016-01-01

    Whole-exome sequencing for clinical applications is now an integral part of medical genetics practice. Though most studies are performed in order to establish diagnoses in individuals with rare and clinically unrecognizable disorders, due to the constantly decreasing costs and commercial availability, whole-exome sequencing has gradually become the initial tool to study patients with clinically recognized disorders when more than one gene is responsible for the phenotype or in complex phenotypes, when variants in more than one gene can be the cause for the disease. Here we report a patient presenting with a complex phenotype consisting of severe, adult-onset, dilated cardiomyopathy, hearing loss and developmental delay, in which exome sequencing revealed two genetic variants that are inherited from a healthy mother: a novel missense variant in the CASK gene, mutations in which cause a spectrum of neurocognitive manifestations, and a second variant, in MYBPC3, that is associated with hereditary cardiomyopathy. We conclude that although the potential for co-occurrence of rare diseases is higher when analyzing undefined phenotypes in consanguineous families, it should also be given consideration in the genetic evaluation of complex phenotypes in non-consanguineous families. PMID:27173948

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates

  12. Nuclear waste management systems issues related to transportation cask design: At-reactor spent fuel storage, Monitored Retrievable Storage and modal mix

    International Nuclear Information System (INIS)

    This report provides background information on nuclear waste transportation issues for an upcoming review of waste shipping cask designs. The focus of this report is related issues pertaining to at-reactor storage, monitored retrievable storage, and the mix of spent fuel transportation modes (railroad, highway and waterways) that will determine impacts of spent fuel transportation to a geologic repository. Part 1 traces the evolution of the civilian radioactive waste management program from its inception through passage of the 1987 amendments to the Nuclear Waste Policy Act (NWPA) of 1982. It emphasizes the factors that will influence the configuration of the transportation system for high-level nuclear waste and related cask design. Part 2 deals with at-reactor storage of wastes. Options for at-reactor storage of waste include expanding pool storage, consolidation and compaction of wastes in pool storage, and various forms of dry storage. Storage needs at-reactor are estimated, and storage options are evaluated on the basis of their ability to meet those needs. Part 3 deals with the MRS facility. The status of the MRS is reviewed starting with the 1987 Nuclear Waste Policy Amendments Act. Studies of the MRS by the State of Tennessee and DOE are reviewed. Alternatives to the MRS, such as the Integrated No-MRS waste management system configuration are discussed. The activities of the MRS Review Commission are also reviewed. Part 4 deals with transportation of wastes from reactors to the MRS or final disposal facility. Road, rail and water transport are evaluated, as are mixtures of different modes. The implications of transportation mode on cask design are discussed, as is the potential for transportation system optimization. The last section applies the as-low-as-reasonably-achievable (ALARA) standard to the problem of radioactive waste transportation. 106 refs., 14 figs., 14 tabs

  13. Determination of uncertainties in the calculation of dose rates at transport and storage casks; Unsicherheiten bei der Berechnung von Dosisleistungen an Transport- und Lagerbehaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Schloemer, Luc Laurent Alexander

    2014-12-17

    The compliance with the dose rate limits for transport and storage casks (TLB) for spent nuclear fuel from pressurised water reactors can be proved by calculation. This includes the determination of the radioactive sources and the shielding-capability of the cask. In this thesis the entire computational chain, which extends from the determination of the source terms to the final Monte-Carlo-transport-calculation is analysed and the arising uncertainties are quantified not only by benchmarks but also by variational calculi. The background of these analyses is that the comparison with measured dose rates at different TLBs shows an overestimation by the values calculated. Regarding the studies performed, the overestimation can be mainly explained by the detector characteristics for the measurement of the neutron dose rate and additionally in case of the gamma dose rates by the energy group structure, which the calculation is based on. It turns out that the consideration of the uncertainties occurring along the computational chain can lead to even greater overestimation. Concerning the dose rate calculation at cask loadings with spent uranium fuel assemblies an uncertainty of (({sup +21}{sub -28}) ±2) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are estimated. For mixed-loadings with spent uranium and MOX fuel assemblies an uncertainty of ({sup +24±3}{sub -27±2}) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are quantified. The results show that the computational chain has not to be modified, because the calculations performed lead to conservative dose rate predictions, even if high uncertainties at neutron dose rate measurements arise. Thus at first the uncertainties of the neutron dose rate measurement have to be decreased to enable a reduction of the overestimation of the calculated dose rate afterwards. In the present thesis

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Feed back on the use of the M.X.6 M.O.X. fuel transport cask: reduction of the doses during operation

    Energy Technology Data Exchange (ETDEWEB)

    Thierry, L.; Eric, P. [COGEMA Logistics (AREVA Group), 78 - Saint Quentin En Yvelines (France); Roland, W. [RWR, Gundremmingen (Germany); Gerhard, D. [E.ON, ISAR (Germany)

    2006-07-01

    The M.X.6 cask was developed by Cogema Logistics for the transport of either BWR or PWR fresh M.O.X. fuel assemblies. Replacing the previous Siemens type III and Siemens BWR packaging, the M.X.6 has been firstly used for the German Nuclear Power Plants. Complying with the T.S.-R-1 (IAEA 1996) regulations, the M.X.6 cask is based on innovative solutions implemented at each step of the design and the manufacturing. Its design includes an efficient neutron shielding for high Plutonium content and an easy use content restraining system. The large payload of the M.X.6, 6 PWR M.O.X. fuel assemblies or 16 BWR M.O.X. fuel assemblies, contributes to the optimisation of the doses uptake during unloading in the NPP. Moreover, new sequences of loading and unloading operations were proposed for testing and implementation in each Nuclear Facility. Concurrently, total dose uptake by the operators was assessed in order to prove the efficiency of the packaging and the proposed sequences. In this paper, the main contributors to the transports to Germany with the M.X.6 will present their involvement and feedback for the reduction of the dose uptake by the operators during the loading and unloading operations. Mainly used at Gundremmingen and Isar Nuclear Power Plant(NPPs), the use of the M.X.6 will be extended to other German and Swiss NPPs from 2006. (authors)

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

    International Nuclear Information System (INIS)

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

  18. Supplement to ORNL/Sub/86-SA094/1 on use of transportable storage casks in the nuclear waste management system

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses the use of transportable storage casks. 12 refs., 7 tabs

  19. Managing aging effects on dry cask storage systems for extended long-term storage and transportation of used fuel - rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y. (Decision and Information Sciences); ( EVS); ( NE)

    2012-07-06

    The cancellation of the Yucca Mountain repository program in the United States raises the prospect of extended long-term storage (i.e., >120 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Application for ISFSI license renewals must include the following: (1) Time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation; and (2) a description of the aging management program (AMP) for management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must also include design bases information as documented in the most recent updated final safety analysis report as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference provided that those references are clear and specific. The NRC has recently issued the Standard Review Plan (SRP) for renewal of used-fuel dry cask storage system (DCSS) licenses and Certificates of Compliance (CoCs), NUREG-1927, under which NRC may renew a specific license or a CoC for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and DCSS that address aging effects that

  20. Managing aging effects on dry cask storage systems for extended long-term storage and transportation of used fuel - rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y. (Decision and Information Sciences); ( EVS); ( NE)

    2012-07-06

    The cancellation of the Yucca Mountain repository program in the United States raises the prospect of extended long-term storage (i.e., >120 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Application for ISFSI license renewals must include the following: (1) Time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation; and (2) a description of the aging management program (AMP) for management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must also include design bases information as documented in the most recent updated final safety analysis report as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference provided that those references are clear and specific. The NRC has recently issued the Standard Review Plan (SRP) for renewal of used-fuel dry cask storage system (DCSS) licenses and Certificates of Compliance (CoCs), NUREG-1927, under which NRC may renew a specific license or a CoC for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and DCSS that address aging effects that

  1. Direct final disposal of transport and storage casks. A realizable technical concept; Direkte Endlagerung von Transport- und Lagerbehaeltern. Ein umsetzbares technisches Konzept

    Energy Technology Data Exchange (ETDEWEB)

    Graf, Reinhold; Brammer, Klaus-Juergen [GNS Gesellschaft fuer Nuklear-Service mbH (Germany); Filbert, Wolfgang [DBE Technology GmbH (Germany)

    2012-11-01

    GNS and DBE TEC developed possible alternatives and supplementary concepts to the existing German reference concept POLLUX and the concept of direct final disposal in boreholes (BSK3) the concept of direct final disposal of transport and storage casks (DIREGT). Advantages of this include the avoidance of necessary elaborate segmentation of fuel elements and core structures, the reduction of waste package transfers and standardized technical equipment for the final disposal engineering. The tasks to be studied include the adaptation of the shaft lifting to the high workload, the adaptation of the underground hauling to the high loads and the development of an appropriate storage technology, considerations concerning the safety with respect to criticality for the demonstration of long-term safety. The basic feasibility of the concept has been demonstrated, the work to be done concerns the demonstration of approvability of the concept for licensing purposes.

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

    International Nuclear Information System (INIS)

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

  3. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowen, Douglas G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  4. Conceptual design of a high-integrity impact limiter for use in shipment of dual-purpose spent-fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. (Applied Science and Technology, Inc., Poway, CA (United States)); Haelsig, R.T.; Hansen, L.J. (Hansen Haelsig Associates, Bellevue, Washington (USA))

    1991-09-01

    A conceptual design for a high-integrity impact limiting system to protect dry metallic spent fuel storage casks during rail transport is proposed. The system is intended to limit the deceleration of the cask during severe rail accidents through three layers of energy-absorbing polyurethane foam material. The crush strengths of the foam is chosen such that the lowest crush strength foam forms the most exterior layer, with the crush strengths increasingly progressively in the two inner layers. The design basis for the external layer of foam is the hypothetical 30-foot free drop impact event prescribed in 10 CFR 71, with a peak steady deceleration limit of about 75 g. The two interior layers absorb up to five times the impact energy of the 30-foot free drop while limiting the decelerations to first 125 g and then to 175 g. The former is felt to be a nominal fuel rod failure threshold, while the latter is at or near the failure level for bolted closure assemblies. These deceleration targets, if met, provide a means for substantially reducing the risk of radioactive material transport. The conceptual design incorporates features for maintaining the integrity of the impact limiter attachment system during severe accidents and enhancing heat dissipation through the impact limiter for short-cooled fuel, through the use of radial aluminum fins. An alternative impact-limiting material -- aluminum honeycomb -- is included in the economic assessment. Both the polyurethane foam and aluminum honeycomb designs appear to meet a cost target of $1.0M, with the polyurethane foam limiter cost estimated at somewhat less than $400K and the aluminum honeycomb cost at somewhat less than $700K. 28 refs., 17 figs., 5 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Martin

    2009-08-20

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

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

    International Nuclear Information System (INIS)

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

  7. The use of albedo neutron dosemeters for the measurement of low doses in mixed photon neutron radiation fields at transport casks for high active waste

    International Nuclear Information System (INIS)

    Radiation exposure of the police forces accompanying transports of spent fuel elements and high-active waste form reprocessing (HAW) is determined by means of albedo dosemeters. The official dosimetry services use this type of dosemeter to mesure the personal dose in mixed gamma/neutron radiation fields above all for routine monitoring of workers occupationally exposed to radiation. The present report describes the detailed set-up and functioning of the albedo dosemeter, the process of obtaining the photon and neutron personal dose from the detector indications as well as the determination of the detection limit of the total personal dose of the albedo dosemeter according to the methods specified in the valid standards. Determination of the detection limit is based on the experience gained during previous transports, on measurements performed at transport casks, on results of type tests at PTB (Federal Physical and Technical Authority), on the measurement uncertainties obtained from the annual intercomparison measurements of the PTB as well as on the test irradiation specially performed in the range of small neutron and photon doses under laboratory conditions. For the dosimetry systems of the dosimetry services and the specific transport conditions, a reference level of 100 μSv was specified with regard to the dose detection limit. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Martin

    2009-08-20

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

  11. Perfil físico-químico de aguardente durante envelhecimento em tonéis de carvalho Chemical profile of aguardente - Brazilian sugar cane alcoholic drink - aged in oak casks

    Directory of Open Access Journals (Sweden)

    Mariana Branco de Miranda

    2008-12-01

    Full Text Available Avaliou-se por um período de 390 dias o perfil da composição química da aguardente sob envelhecimento em tonéis de carvalho de 20 L. O envelhecimento da aguardente em tonéis de madeira melhora a qualidade sensorial do destilado. As aguardentes envelhecidas foram analisadas aos 0, 76, 147, 228, 314 e 390 dias de armazenamento quanto às concentrações de etanol, acidez volátil, ésteres, aldeídos, furfural, álcoois superiores (n-propílico, isobutílico e isoamílicos, metanol, cobre, extrato seco, taninos e cor. Após os 390 dias de armazenamento, a aguardente apresentou maiores concentrações de acidez volátil, ésteres, aldeídos, furfural, álcoois superiores, congêneres, extrato seco e tanino. Sua coloração tornou-se amarelada. As concentrações de etanol e de metanol não se alteraram, e o teor de cobre apresentou ligeiro declínio. O envelhecimento da aguardente por 390 dias em tonéis de carvalho alterou a sua composição química, porém ela se manteve dentro de todos os padrões de qualidade estabelecidos pela legislação nacional em vigor.The chemical composition of aguardente - Brazilian sugar cane alcoholic drink - under aging during in 20 L oak casks was evaluated for 390 days. Aging sugar cane aguardente in wood casks improves the sensorial quality of the distillate. The concentrations of ethanol, volatile acidity, esters, aldehydes, furfural, higher alcohols (n-propylic, isobutylic and isoamylics, methanol, copper, dry extract, tannins, and color of the aged sugar cane aguardente were analysed at 0, 76, 147, 228, 314, and 390 days of storage. After 390 days of aging the sugar cane aguardente presented higher concentrations of volatile acidity, esters, aldehydes, furfural, higher alcohols, congeners, dry extract, and tannin. Its color became golden. The concentrations of ethanol and methanol did not change and the copper content decreased slightly. The aging of the sugar cane aguardente in oak casks for 390 days

  12. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage; Estudo de um casco nacional e sua instalacao para armazenagem seca de combustivel nuclear queimado gerado em reatores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio

    2009-07-01

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  13. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  14. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  15. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  16. The Fusion of Horror and Aesthetics:Gothic Narration in The Cask of Amontillado%恐怖与美学的融合:《阿芒提拉多酒桶》的哥特式叙事

    Institute of Scientific and Technical Information of China (English)

    孙峰

    2014-01-01

    As an important way of narration, Gothic style is vividly and incisively applied in many of Edgar Allan Poe’s short stories, showing quite a lot of horrible and scary scenes for the readers. While reaching to an extreme, the horror may turn to its opposite side and presents the story with a strange beauty of literature, which is bestowed with unconventional aesthetic characteristics. As one of Edgar Allan Poe’s masterpieces, the short story The Cask of Amontillado deals with the theme of revenge and death. It contains only a few characters and its plot is not too complex. However, the story has a rich atmosphere of horror, which conveys exactly the Gothic style and achieves an unusual artistic effect.%作为一种重要的叙事方式,哥特手法在爱伦·坡的众多短篇小说中得到了淋漓尽致的运用,为读者展现了一幕幕令人惊悚恐惧的场景。这种恐怖达到一个极致,便向它的相反面转化,使作为载体的故事情节呈现了一种异样的文学之美,从而被赋予了反传统的美学艺术特征。短篇小说《阿芒提拉多酒桶》作为爱伦·坡的短篇代表作之一,以复仇和死亡为主题,人物不多,情节也不太复杂,但却有着浓郁的恐怖气氛,恰到好处地传达出这种创作特征,实现了异乎寻常的艺术效果。

  17. Aspectos da composição química e aceitação sensorial da aguardente de cana-de-açúcar envelhecida em tonéis de diferentes madeiras Aspects of the chemical composition and sensorial acceptance of sugar cane spirit aged in casks of different types of woods

    Directory of Open Access Journals (Sweden)

    André Ricardo Alcarde

    2010-05-01

    sugar cane spirit aged for 3 years in casks of different types of wood (peanut wood, araruva or striped wood, red cabreuva, oak, cherrywood, Brazilian gold wood, purple tabebuia, cariniana legalis, and pear tree. The simple alcoholic distillate which originated the sugar cane spirit was produced at the Distillery of ESALQ/USP. After aging, the sugar cane spirits were analyzed in terms of ethanol concentrations o, volatile acidity, furfural, aldehydes, esters, higher alcohols, methanol, copper, total phenolic compounds, color, and sensorial acceptance. Regardless the type of wood the casks were made of, the aged sugar cane spirits became darker and presented higher concentrations of volatile acidity, furfural, esters, higher alcohols, congeners, and total phenolic compounds than the simple alcoholic distillate. On the other hand, the aged sugar cane spirits presented lower concentrations of aldehydes, methanol, and copper than the simple alcoholic distillate. The statistical analysis, considering the global physicochemical composition of the sugar cane spirits aged in the casks made of different types of wood, showed similarities among the sugar cane spirits aged in the casks of peanut wood, araruva or striped wood, and cariniana legalis. It also indicates similarities among the sugar cane spirits aged in the casks of red cabreuva and pear tree and among the sugar cane spirits aged in the casks of oak, cherrywood, Brazilian gold wood, and purple tabebuia. The sugar cane spirits aged in the casks of the different types of wood were in accordance with the composition and quality standards established by the Brazilian laws. The sugar cane spirit aged in oak presented the best sensorial acceptance. Among the Brazilian woods, purple tabebuia, peanut wood, red cabreuva, cherrywood and pear tree were those that produced sugar cane spirits with better sensorial qualities.

  18. Effects of temperature on concrete cask in a dry storage facility for spent nuclear fuels%退役核燃料干式贮存混泥土护箱受高温作用之效应研究

    Institute of Scientific and Technical Information of China (English)

    黄伟庆; 吴瑞贤; 郑裕宽

    2011-01-01

    In the dry storage of spent nuclear fuels, concrete cask serves both as a shielding and a structural containment. The concrete in the storage facility is expected to endure the decay heat of the spent nuclear fuel during its service life. Thus, effects of the sustaining high temperature on concrete material need be evaluated for safety of the dry storage facility. In this paper, we report an experimental program aimed at investigating possible high temperature effects on properties of concrete, with emphasis on the mechanical stability, porosity, and crack-resisting ability of concrete mixes prepared using various amounts of Portland cement, fly ash, and blast furnace slag. The experimental results obtained from concrete specimens exposed to a temperature of 94℃ for 90 days indicate that: (1)compressive strength of the concrete remains practically unchanged; (2) the ultrasonic pulse velocity, and dynamic modulus of elasticity of the concrete decrease in early stage of the high-temperature exposure, and gradually become stable with continuing exposure; (3) shrinkage of concrete mixes exhibits an increase in early stage of the exposure and does not decrease further with time; (4) concrete mixes containing pozzolanic materials, including fly ash and blast furnace slag, show better temperature-resisting characteristics than those using only Portland cement.%退役核燃料干式贮存设施主体由混凝土构成,混凝土得在长时期内承受残余核燃料释出的衰变热,加上台湾地区特殊的环境气候条件,混凝土材料可能产生劣化.依据核能安全混凝土结构物的材料规定的配比,我们制作了混凝土试样,用实验室模拟法研究干式贮存混凝土护箱在高温环境作用下可能出现的损害或劣化,甚至耐久性变差等.利用非破坏性检测方法(超音波试验、动弹性模数试验及反弹锤试验等),观察混凝土受持续高温作用下的结构致密性及内部是否产生裂缝,而影响

  19. Cask systems development program seal technology

    Energy Technology Data Exchange (ETDEWEB)

    Madsen, M.M.; Edwards, K.R.; Humphreys, D.L.

    1991-01-01

    General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the US Code of Federal Regulations Part 71 (10 CFR 71). Seals that provide the containment system interface between the packaging body and the closure must function in both high- and low-temperature environments under dynamic and static conditions. Experiments were performed to characterize the performance of several seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fuorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Results show that the seal materials tested, with the exception of silicone S613-60, are not leak tight at manufacturer low-temperature ratings. This paper documents the initial series of experiments developed to characterize the performance of several static seals under conditions representative of RAM transport container environments. Helium leak rates of face seals were measured at low and ambient temperatures to compare seal materials. As scaling laws have not been developed for seals, the leakage rates measured in this program are intended to be used in a qualitative rather than quantitative manner. 5 refs., 7 figs., 2 tabs.

  20. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... fall within the scope of the definition of small entities set forth in the Regulatory Flexibility Act... will be considered if it is practical to do so, but the NRC staff is able to ensure consideration only... Document Room (PDR) reference staff at 1-800-397-4209, 301-415- 4737, or by email to...

  1. Licensing of LLW final storage casks as transport casks of type IP for fissile radioactive materials

    International Nuclear Information System (INIS)

    In 1995 Siemens AG decided to stop the manufacture of fuel assemblies in Hanau and to decommission the Uranium and MOX plant. Since this time about 70.000 tons of radioactive contaminated materials have either been recycled or prepared for final storage in the planned repository Konrad. In the case of Siemens AG Hanau essentially two types of containers were used, which are called Steel Container Type IV and Type VI

  2. Property enhancement of cast iron used for nuclear casks

    Science.gov (United States)

    Behera, R. K.; Mahto, B. P.; Dubey, J. S.; Mishra, S. C.; Sen, S.

    2016-01-01

    Ductile iron (DI) is a preferred material for use in various structural, automotive, and engineering fields because of its excellent combination of strength, toughness, and ductility. In the current investigation, we elucidate the relationship between the morphological and mechanical properties of DI intended for use in safety applications in the nuclear industry. DI specimens with various alloying elements were subjected to annealing and austempering heat treatment processes. A faster cooling rate appeared to increase the nodule count in austempered specimens, compensating for their nodularity value and subsequently decreasing their ductility and impact strength. The ductility and impact energy values of annealed specimens increased with increasing ferrite area fraction and nodularity, whereas an increase in the amounts of Ni and Cr resulted in an increase of hardness via solid solution strengthening. Austempered specimens were observed to be stronger than annealed specimens and failed in a somewhat brittle manner characterized by a river pattern, whereas the ductile failure mode was characterized by the presence of dimples.

  3. Testing and modelling of shock absorbing materials in transportation casks

    International Nuclear Information System (INIS)

    Soft Impact Limiters, such as polyurethane foams and aluminum honeycombs are being studied to develop a broad base of information on the mechanical behavior of these materials. Static and dynamic tests under different load configurations were carried out and the results are presented and discussed. Types of material tested included aluminum honeycombs (hexagonal cell structure) and closed cell rigid polyurethane foams. Four different densities of each material were tested. (author)

  4. Discussion on the Gothic features in A Cask of Amontillado

    Institute of Scientific and Technical Information of China (English)

    刘亚平

    2011-01-01

    Edgar Allan Poe is a writer who establishes a new school in the American history of literature.He firstly and consciously takes the short story as an independent literature style,advances his own composing theory of short stories and applies it to his com

  5. Neutron absorber plate and radioactive material transportation cask

    International Nuclear Information System (INIS)

    Aluminum alloy flame-coating layers are formed at the outer surface of a neutron absorber plate in order to prevent corrosion due to potential difference. However, pin holes of micron order are sometimes formed on the flame-coating membranes, which are hard to be found by usual inspection. Then, ferrous flame-coating membranes are formed at the outer surface of boron carbide and aluminum alloy flame-coating membranes are formed at the outer surface thereof. The outer surface of a boron carbide plate is coated with the ferrous flame-coating membranes instead of being coated with an external plate made of neutron cells, and an aluminum alloy flame-coating membranes or mixed flame-coating layers of aluminum oxide and titania are coated thereover in order to prevent rusts. Whether the pin holes are present or not can be confirmed easily by a ferroxyl test. If there are pin holes, flame-coating is applied again to form complete membranes. Then, since it is no more necessary to fix a neutron absorbing cell at the outer surface of a fuel cell by means of welding, production cost can be reduced. (N.H.)

  6. Comparison of cask and drywell storage concepts for a monitored retrievable storage/interim storage system

    International Nuclear Information System (INIS)

    The Department of Energy, through its Richland Operations Office is evaluating the feasibility, timing, and cost of providing a federal capability for storing the spent fuel, high-level wastes, and transuranic wastes that DOE may be obligated by law to manage until permanent waste disposal facilities are available. Three concepts utilizing a monitored retrievable storage/interim storage (MRS/IS) facility have been developed and analyzed. The first concept, co-location with a reprocessing plant, has been developed by staff of Allied General Nuclear Services. the second concept, a stand-alone facility, has been developed by staff of the General Atomic Company. The third concept, co-location with a deep geologic repository, has been developed by the Pacific Northwest Laboratory with the assistance of the Westinghouse Hanford Company and Kaiser Engineers. The objectives of this study are: to develop preconceptual designs for MRS/IS facilities: to examine various issues such as transportation of wastes, licensing of the facilities, and environmental concerns associated with operation of such facilities; and to estimate the life-cycle costs of the facilities when operated in response to a set of scenarios that define the quantities and types of waste requiring storage in specific time periods, generally spanning the years 1989 to 2037. Three scenarios are examined to develop estimates of life-cycle costs for the MRS/IS facilities. In the first scenario, the reprocessing plant is placed in service in 1989 and HLW canisters are stored until a repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at intervals as needed to meet the demand. In the second scenario, the reprocessing plants are delayed in starting operations by 10 years, but the repositories open on schedule. In the third scenario, the repositories are delayed 10 years, but the reprocessing plants open on schedule

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

    International Nuclear Information System (INIS)

    COBRA-SFS is a general thermal-hydraulic analysis computer code for prediction of material temperatures and fluid conditions in a wide variety of systems. The code has been validated for analysis of spent fuel storage systems, as part of the Commercial Spent Fuel Management Program of the US Department of Energy. The code solves finite volume equations representing the conservation equations for mass, moment, and energy for an incompressible single-phase heat transfer fluid. The fluid solution is coupled to a finite volume solution of the conduction equation in the solid structure of the system. This document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part 1 describes the conservation equations, constitutive models, and solution methods used in the code. Part 2 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part 3 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available

  8. 75 FR 41404 - List of Approved Spent Fuel Storage Casks: NUHOMS®

    Science.gov (United States)

    2010-07-16

    ... average size of the boron carbide ] particles in the finished product is approximately 50 microns after....3 of the SAR, are not precisely quantified in that it requires that ``the average size of the boron carbide particles in the finished product is approximately 50 microns after rolling.'' Use of...

  9. 75 FR 41369 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD Revision 1; Withdrawal

    Science.gov (United States)

    2010-07-16

    ... boron carbide particles in the finished product is approximately 50 microns after rolling.'' Use of... average size of the boron carbide particles in the finished product is approximately 50 microns after... FR 24786), is withdrawn. FOR FURTHER INFORMATION CONTACT: Jayne M. McCausland, Office of Federal...

  10. 76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... Federal Register. Procedural Background On May 6 and 7, 2010, respectively, a direct final rule (75 FR 24786) and companion proposed rule (75 FR 25120) were published in the Federal Register to revise the... implemented. On July 16, 2010, the NRC withdrew the direct final rule (75 FR 41369) and the companion...

  11. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

  12. 78 FR 8050 - Spent Fuel Cask Certificate of Compliance Format and Content

    Science.gov (United States)

    2013-02-05

    ... Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory... safety and is cost-justified.'' 3. Delete the requirement in 10 CFR 72.212(b)(6) for general licensees to... direct and indirect costs of implementation are outweighed by the increased protection.'' The...

  13. A proposal for an international brittle fracture acceptance criterion for nuclear material transport cask applications

    Energy Technology Data Exchange (ETDEWEB)

    Sorenson, K.B.; Salzbrenner, R.J.; Nickell, R.E.

    1989-01-01

    This paper presents a fundamental basis for a brittle fracture acceptance criterion, examine several existing criteria and propose examples for consideration as international brittle fracture acceptance criteria. The proposed criteria are intended to stimulate discussion in order to advance the development of a consensus approach. 8 refs., 1 fig., 1 tab.

  14. German physical protection concept for the storage of spent fuel elements in transport and storage casks

    International Nuclear Information System (INIS)

    Full text: In Germany, the legal regulations and requirements derived from rules and guidelines for the protection of storage facilities for spent fuel elements from disruptive action or other inference by third parties are structured hierarchically. The Atomic Energy Act constitutes the top level. It is supported by federal ordinances. The next level down is formed by the rules and guidelines. The storage of nuclear fuels may only be authorized, according to the provisions of the Atomic Energy Act, if the required protection from disruptive action or other interference by third parties can be guaranteed as following: it must be possible to prevent any danger to life and health due to a substantial amount of direct radiation or due to the release of a substantial amount of radioactive material; it must be possible to prevent singular or repeated acts of stealing nuclear fuels in such amounts that a critical accumulation can be produced directly without reprocessing and enrichment. Knowing that nuclear installations cannot be protected from every possible interference, physical protection is focused on basic security standards, the so-called design basic threat (DBT), departing from the assumed interference. DBT is regularly reviewed by the competent federal authorities and authorities of the states and are revised on the basis of newly gained knowledge, if necessary, such as in the wake of the terrorist attacks in the U.S. on September 11, 2001. The operator must guarantee and give proof of a sufficient level of physical protection of the plant. The sole physical protection measures implemented by the operator cannot ensure the required protection from other interference by third parties for an unlimited time span. The concept therefore requires additional physical protection measures by the police. (author)

  15. The Structural Design and Analysis of Pallet in ITER Transfer Cask for Remote Handling Operations

    Science.gov (United States)

    Zhou, Zibo; Yao, Damao; Cao, Lei; Li, Ge

    2009-06-01

    Necessary adjustment ranges and accuracies of the pallet for ITER are presented. Detailed structural designs and structural finite element analyses for pallet components are made to determine whether the results satisfy the requirements of the pallet structure to be used in ITER.

  16. The Structural Design and Analysis of Pallet in ITER Transfer Cask for Remote Handling Operations

    Institute of Scientific and Technical Information of China (English)

    ZHOU Zibo; YAO Damao; CAO Lei; LI Ge

    2009-01-01

    Necessary adjustment ranges and accuracies of the pallet for ITER are presented. Detailed structural designs and structural finite element analyses for pallet components are made to determine whether the results satisfy the requirements of the pallet structure to be used in ITER.

  17. Improvement of input parameters for the estimation of fuel rod temperature in dry transport cask

    International Nuclear Information System (INIS)

    A typical PWR spent fuel bundle has a 17 x 17 rod array, and an analysis requires a very long computation time and a vast amount of memory. Therefore, we applied the lumped fuel bundle analysis approach with the homogenized method to estimate the fuel cladding temperature efficiency. Thermal analysis results for lumped fuel bundles showed an excessive radiative heat transfer, and we applied an emissivity modification factor to compensate for this radiation effect. The value of the factor decreased as the number of the rods in the homogenized array decreased.. For the lumped 8 x 8 array, the best emissivity modification factor was shown to be 0.40. The rod emissivity of 0.8 is generally recommended to be used in COBRA-SFS[D. R. Rector et al.] calculations. Therefore, we can use the modified rod emissivity of 0.32 for lumped 8 x 8 array. There are good agreements between the results from lumped 8 x 8 array bundle and the results from real 17 x 17 array bundle. By homogenization, we can increase the computational speed substantially, as well as reduce the requirements on computer memory and space. (authors)

  18. Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks

    International Nuclear Information System (INIS)

    This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs

  19. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ... National Technology Transfer and Advancement Act of 1995 (Pub. L. 104-113) requires that Federal agencies... spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced Generation... explains why the rule would be inappropriate, including challenges to the rule's underlying premise...

  20. Validation Experiments for Spent-Fuel Dry-Cask In-Basket Convection

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Barton L. [Utah State Univ., Logan, UT (United States). Dept. of Mechanical and Aerospace Engineering

    2016-08-16

    This work consisted of the following major efforts; 1. Literature survey on validation of external natural convection; 2. Design the experiment; 3. Build the experiment; 4. Run the experiment; 5. Collect results; 6. Disseminate results; and 7. Perform a CFD validation study using the results. We note that while all tasks are complete, some deviations from the original plan were made. Specifically, geometrical changes in the parameter space were skipped in favor of flow condition changes, which were found to be much more practical to implement. Changing the geometry required new as-built measurements, which proved extremely costly and impractical given the time and funds available

  1. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Science.gov (United States)

    2010-01-01

    ... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR... spent fuel (i.e., intact assembly or consolidated fuel rods), the inerting atmosphere requirements. (b... removal of the stored spent fuel from a reactor site, transportation, and ultimate disposition by...

  2. Validation Experiments for Spent- Fuel Dry-Cask In-Basket Convection

    Energy Technology Data Exchange (ETDEWEB)

    Smith, barton [Utah State Univ., Logan, UT (United States)

    2016-08-16

    This work consisted of the following major efforts; 1. Literature survey on validation of external natural convection; 2. Design the experiment; 3. Build the experiment; 4. Run the experiment; 5. Collect results; 6. Disseminate results; and 7. Perform a CFD validation study using the results. We note that while all tasks are complete, some deviations from the original plan were made. Specifically, geometrical changes in the parameter space were skipped in favor of flow condition changes, which were found to be much more practical to implement. Changing the geometry required new as-built measurements, which proved extremely costly and impractical given the time and funds available

  3. 76 FR 70374 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... ID NRC- 2011-0008. Address questions about NRC dockets to Carol Gallagher, telephone: (301) 492-3668, email: Carol.Gallagher@nrc.gov . Mail comments to: Secretary, U.S. Nuclear Regulatory Commission..., 2011. For the Nuclear Regulatory Commission. Michael F. Weber, Acting Executive Director for...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. 75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-07

    ... HD System to include pressurized water reactor fuel assemblies with control components, reduce the... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR- RELATED GREATER THAN CLASS C...; #0; #0;#0;Federal Register / Vol. 75, No. 88 / Friday, May 7, 2010 / Proposed Rules#0;#0; ]...

  7. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Science.gov (United States)

    2010-01-01

    ... Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C... adversely affected structures, systems, and components important to safety....

  8. 76 FR 2243 - List of Approved Spent Fuel Storage Casks: NUHOMS ® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... pressurized water reactor fuel assemblies with control components; reduce the minimum initial enrichment of... sector, for the dry storage of spent nuclear fuel at civilian nuclear power reactor sites, with the... sites of civilian nuclear power reactors without, to the maximum extent practicable, the need...

  9. 75 FR 24786 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-06

    ... include pressurized water reactor fuel assemblies with control components, reduce the minimum initial..., for the dry storage of spent nuclear fuel at civilian nuclear power reactor sites, with the objective... sites of civilian nuclear power reactors without, to the maximum extent practicable, the need...

  10. Dry storage of the BR3 spent fuel in the Castor BR3 cask

    Energy Technology Data Exchange (ETDEWEB)

    Ooms, L.; Massaut, V.; Noynaert, L. [SCK/CEN, B-2400 Mol (Belgium); Braeckeveldt, M. [Niras/Ondraf, B-1210 Brussels (Belgium)

    2001-07-01

    Twenty-five years of operation has resulted in an inventory of spent fuel with a wide variety in the BR3 nuclear pilot power plant. Studies were launched to evaluate all possible solutions for the BR3 experimental and 'exotic' spent fuel, i.e. reprocessing, dry storage in containers and dry storage in canisters. For the BR3 spent fuel the interim dry storage in Castor BR3 containers was chosen. The present paper describes in a first part the history and characteristics of the spent fuel. A second part handles with the different options, which were studied for the spent fuel evacuation. The last part focuses on the spent fuel preparation and the production of the Castor BR3. This project allowed the SCK-CEN to build up an important know-how in the field of spent fuel management, especially the management of research reactor fuel, which is very specific and not comparable with spent fuel of commercial nuclear power plants. (author)

  11. 78 FR 37927 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-06-25

    ... Federal Regulations is sold by the Superintendent of Documents. #0;Prices of new books are listed in the... revises authorized contents to include: pressurized water reactor (PWR) damaged fuel contained in damaged.... Tanious, Office of Federal and State Materials and Environmental Management Programs, U.S....

  12. Comparison of cask and drywell storage concepts for a monitored retrievable storage/interim storage system

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, D.E.

    1982-12-01

    The Department of Energy, through its Richland Operations Office is evaluating the feasibility, timing, and cost of providing a federal capability for storing the spent fuel, high-level wastes, and transuranic wastes that DOE may be obligated by law to manage until permanent waste disposal facilities are available. Three concepts utilizing a monitored retrievable storage/interim storage (MRS/IS) facility have been developed and analyzed. The first concept, co-location with a reprocessing plant, has been developed by staff of Allied General Nuclear Services. the second concept, a stand-alone facility, has been developed by staff of the General Atomic Company. The third concept, co-location with a deep geologic repository, has been developed by the Pacific Northwest Laboratory with the assistance of the Westinghouse Hanford Company and Kaiser Engineers. The objectives of this study are: to develop preconceptual designs for MRS/IS facilities: to examine various issues such as transportation of wastes, licensing of the facilities, and environmental concerns associated with operation of such facilities; and to estimate the life-cycle costs of the facilities when operated in response to a set of scenarios that define the quantities and types of waste requiring storage in specific time periods, generally spanning the years 1989 to 2037. Three scenarios are examined to develop estimates of life-cycle costs for the MRS/IS facilities. In the first scenario, the reprocessing plant is placed in service in 1989 and HLW canisters are stored until a repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at intervals as needed to meet the demand. In the second scenario, the reprocessing plants are delayed in starting operations by 10 years, but the repositories open on schedule. In the third scenario, the repositories are delayed 10 years, but the reprocessing plants open on schedule.

  13. Binding of Y-P30 to syndecan 2/3 regulates the nuclear localization of CASK

    NARCIS (Netherlands)

    Landgraf, P.; Mikhaylova, Marina; Macharadze, T.; Borutzki, C.; Zenclussen, A.C.; Wahle, P.; Kreutz, M.R.

    2014-01-01

    The survival promoting peptide Y-P30 has documented neuroprotective effects as well as cell survival and neurite outgrowth promoting activity in vitro and in vivo. Previous work has shown that multimerization of the peptide with pleiotrophin (PTN) and subsequent binding to syndecan (SDC) -2 and -3 i

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  15. Liprin-α2 promotes the presynaptic recruitment and turnover of RIM1/CASK to facilitate synaptic transmission

    NARCIS (Netherlands)

    S.A. Spangler (Samantha); S.K. Schmitz (Sabine); J.T. Kevenaar (Josta); E. de Graaff (Esther); M. De Wit (Meike); J.A.A. Demmers (Jeroen); P.W. Toonen (Pim ); C.C. Hoogenraad (Casper)

    2013-01-01

    textabstractThe presynaptic active zone mediates synaptic vesicle exocytosis, and modulation of its molecular composition is important for many types of synaptic plasticity. Here, we identify synaptic scaffold protein liprin-α2 as a key organizer in this process. We show that liprin-α2 levels were r

  16. Modelling of pool fire environments using experimental results of a two-hour test of a railcar/cask system

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, J.E.; Klein, D.E.; Pope, R.B.; Yoshimura, H.R.

    1980-01-01

    It was demonstrated that time and spatial variations in the local source temperatures, the radiant shielding of intervening structure and the effects of wind can significantly affect the amount of heat input to a large package in a simulated accidental fire. The pool fire provided a significantly non-uniform heat source to the package. Despite these effects, however, the amount of heat input to the package was generally equivalent to that which would be received from a regulatory 800/sup 0/C uniform thermal source. 7 firegures.

  17. 76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ... Reactor (BWR) fuel with high initial enrichment (up to 4.8 weight percent uranium-235 planer average...) The ability to store and transport BWR fuel with high initial enrichment (up to 4.8 weight percent... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR...

  18. Gamma irradiation tests of concrete material recommended for storage casks of spent nuclear fuel arising from Cernavoda NPP

    International Nuclear Information System (INIS)

    Considerable effort is being devoted in the Romania's Nuclear Spent Fuel and Waste Management R and D Program to develop engineered barriers for the containment of nuclear fuel waste under conditions of deep geological disposal. Engineering practice suggests that the concrete should fulfill the requirements of long term physical stability and resistance to radiation damage. With an appropriate system of metal reinforcement, it should be possible to obtain the tensile and impact strength required avoiding the risk of mechanical damage during handling and emplacement. In accordance with the concept developed by SITON-Bucharest, presently, the dry storage of spent nuclear fuel is thought by two choices: the alternative of dry storage type MMB3 and the alternative of dry storage type TRANSTOR. By using ORIGEN and PELSHIE computer codes, we evaluated the absorbed gamma radiation dose absorbed by the concrete walls of the storage vault both in MMB3 and in TRANSTOR designing choice.The irradiation tests were performed at the Gamma Irradiation Facility from the Institute for Nuclear Research. (authors)

  19. Gamma irradiation tests of concrete material recommended for storage casks of spent nuclear fuel arising from Cernavoda NPP

    International Nuclear Information System (INIS)

    Considerable effort is being devoted to the Romania's Nuclear Spent Fuel and Waste Management R and D Program to develop engineered barriers for the containment of nuclear fuel waste under conditions of deep geological disposal. Engineering practice suggests that the concrete should fulfil the requirements of long term physical stability and resistance to radiation. With an appropriate system of metal reinforcement, it should be possible to obtain the tensile and impact strength required, avoiding the risk of mechanical damage during handling and emplacement. In accordance with the concept developed by CITON-Bucharest, presently, the dry storage of spent nuclear fuel is thought by two choices: - The alternative of dry storage type MMB3; - The alternative of dry storage type TRANSTOR. By using ORIGEN and PELSHIE computer codes, we evaluated the gamma radiation dose absorbed by the concrete walls of the storage vault both in MMB3 and in TRANSTOR designing variants. The irradiation tests were performed at the Gamma Irradiation Facility of the Institute for Nuclear Research. (authors)

  20. Actinide Partitioning-Transmutation Program Final Report. V. Preconceptual designs and costs of partitioning facilities and shipping casks (appendix 3)

    International Nuclear Information System (INIS)

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