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Evaluation of impact tests of solid steel billet onto concrete pads, and application to generic ISFSI storage cask for tipover and side drop

Description: Twelve tests were performed at LLNL to assess loading conditions on a spent fuel casts for side drops, end drops and tipover events. The tests were performed with a 1/3-scale model concrete pad to benchmark the structural analysis code DYNA3D. The side drop and tipover test results are discussed in this report. The billet and test pad were modified with DYNA3D using material properties and techniques used in earlier tests. The peak or maximum deceleration test results were compared to the simulated analytical results. It was concluded that an analytical model based on DYNA3D code and has been adequately benchmarked for this type of application. A generic or represented cask was modified with the DYNA3D code and evaluated for ISFSI side drop and tipover events. The analytical method can be applied to similar casks to estimate impact loads on storage casks resulting from low-velocity side or tip impacts onto concrete storage pads.
Date: May 1, 1997
Creator: Witte, M.C.; Chen, T.F.; Murty, S.S.; Tang, D.T.; Mok, G.C.; Fischer, L.E. et al.
Partner: UNT Libraries Government Documents Department

Design analysis report for the TN-WHC cask and transportation system

Description: 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.
Date: February 13, 1997
Creator: Brisbin, S.A., Fluor Daniel Hanford
Partner: UNT Libraries Government Documents Department

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

Description: This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask`s primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives.
Date: April 1, 1995
Creator: Powell, F.P.
Partner: UNT Libraries Government Documents Department

Guide to verification and validation of the SCALE-4 radiation shielding software

Description: Whenever a decision is made to newly install the SCALE radiation shielding software on a computer system, the user should run a set of verification and validation (V&V) test cases to demonstrate that the software is properly installed and functioning correctly. This report is intended to serve as a guide for this V&V in that it specifies test cases to run and gives expected results. The report describes the V&V that has been performed for the radiation shielding software in a version of SCALE-4. This report provides documentation of sample problems which are recommended for use in the V&V of the SCALE-4 system for all releases. The results reported in this document are from the SCALE-4.2P version which was run on an IBM RS/6000 work-station. These results verify that the SCALE-4 radiation shielding software has been correctly installed and is functioning properly. A set of problems for use by other shielding codes (e.g., MCNP, TWOTRAN, MORSE) performing similar V&V are discussed. A validation has been performed for XSDRNPM and MORSE-SGC6 utilizing SASI and SAS4 shielding sequences and the SCALE 27-18 group (27N-18COUPLE) cross-section library for typical nuclear reactor spent fuel sources and a variety of transport package geometries. The experimental models used for the validation were taken from two previous applications of the SASI and SAS4 methods.
Date: December 1, 1996
Creator: Broadhead, B.L.; Emmett, M.B. & Tang, J.S.
Partner: UNT Libraries Government Documents Department

Spent nuclear fuel project detonation phenomena of hydrogen/oxygen in spent fuel containers

Description: Movement of Spent N Reactor fuels from the Hanford K Basins near the Columbia River to Dry interim storage facility on the Hanford plateau will require repackaging the fuel in the basins into multi-canister overpacks (MCOs), drying of the fuel, transporting the contained fuel, hot conditioning, and finally interim storage. Each of these functions will be accomplished while the fuel is contained in the MCOs by several mechanisms. The principal source of hydrogenand oxygen within the MCOs is residual water from the vacuum drying and hot conditioning operations. This document assesses the detonation phenomena of hydrogen and oxygen in the spent fuel containers. Several process scenarios have been identified that could generate detonation pressures that exceed the nominal 10 atmosphere design limit ofthe MCOS. Only 42 grams of radiolized water are required to establish this condition.
Date: September 30, 1996
Creator: Cooper, T. D.
Partner: UNT Libraries Government Documents Department

SNF shipping cask shielding analysis

Description: 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.
Date: January 1, 1996
Creator: Johnson, J.O. & Pace, J.V. III
Partner: UNT Libraries Government Documents Department

Multi-Canister overpack design pressure rating

Description: The SNF project was directed to increase the MCO pressure rating by the U.S. Department of Energy, Richland Operations Office (RL) unless the action was shown to be cost prohibitive. This guidance was driven by RL's assessment that there was a need to improve margin and reduce risks associated with assumptions supporting the bounding pressure calculation for the MCO Sealing Strategy. Although more recent pressure analyses show a bounding MCO pressure of 50 psig, RL still considers it prudent to retain the pressure margin the 450 psig rating provides. This rating creates a real, clearly definable margin and significantly reduces the risk that the safety basis will be challenged.
Date: November 3, 1998
Creator: Smith, K. E.
Partner: UNT Libraries Government Documents Department

MCO Monitoring activity description

Description: Spent Nuclear Fuel remaining from Hanford's N-Reactor operations in the 1970s has been stored under water in the K-Reactor Basins. This fuel will be repackaged, dried and stored in a new facility in the 200E Area. The safety basis for this process of retrieval, drying, and interim storage of the spent fuel has been established. The monitoring of MCOS in dry storage is a currently identified issue in the SNF Project. This plan outlines the key elements of the proposed monitoring activity. Other fuel stored in the K-Reactor Basins, including SPR fuel, will have other monitoring considerations and is not addressed by this activity description.
Date: November 9, 1998
Creator: SEXTON, R.A.
Partner: UNT Libraries Government Documents Department

Design package lazy susan for the fuel retrieval system

Description: This is a design package that contains the details for a Lazy Susan style small tool for the Fuel Retrieval System. The Lazy Susan tool is used to help rotate an MCO Fuel Basket when loading it. This document contains requirements, development design information, tests and test reports that pertain to the production of Lazy Susan small tool.
Date: September 10, 1999
Creator: TEDESCHI, D.J.
Partner: UNT Libraries Government Documents Department

CPP-603 underwater fuel storage facility site integrated stabilization management plan (SISMP). Volume I

Description: The CPP-603 Underwater Fuel Storage Facility (UFSF) Site Integrated Stabilization Management Plan (SISMP) has been developed to describe the activities required for the relocation of spent nuclear fuel (SNF) from the CPP-603 facility. These activities are the only Idaho National Engineering Laboratory (INEL) actions identified in the Implementation Plan developed to meet the requirements of the Defense Nuclear Facilities Safety Board (DNFSB) remediation in the Defense Nuclear Facilities Complex. To date, 622 spent nuclear fuel units have been moved from the CPP-603 north and middle water basins, leaving 743 units in the south basin to be relocated from the facility by December 31, 2000. Besides moving fuels from the CPP-603, in 1993 and 1994 more than 300 fuel storage yokes in the north and middle basins were redundantly rigged because of corrosion problems. More than 200 fuel transfers within the north and middle basins were also made to ensure proper spacing of the fuels, and 104 corroded cans containing spent space reactor fuel were repackaged underwater to prevent potential release of their contents. This document is provided to address the relocation activities for the remaining 743 units in the south basin into wet storage pools at building CPP-666 or into dry storage at the Irradiation Fuel Storage Facility (IFSF).
Date: September 1, 1996
Creator: Wachs, G.W.; Blake, H.M.; Cottam, R.E.; Denney, R.D. & Shiffern, R.A.
Partner: UNT Libraries Government Documents Department