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Viability of Existing INL Facilities for Dry Storage Cask Handling

Description: 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.
Date: April 1, 2013
Creator: Bohachek, Randy; Park, Charles; Wallace, Bruce; Winston, Phil & Marschman, Steve
Partner: UNT Libraries Government Documents Department

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

Releases of UF{sub 6} to the atmosphere after a potential fire in a cylinder storage yard

Description: Uranium hexafluoride (UF{sub 6}), a toxic material, is stored in just over 6200 cylinders at the K-25 site in Oak Ridge, Tennessee. The safety analysis report (SAR) for cylinder yard storage operations at the plant required the development of accident scenarios for the potential release of UF{sub 6} to the atmosphere. In accordance with DOE standards and guidance, the general approach taken in this SAR was to examine the functions and contents of the cylinder storage yards to determine whether safety-significant hazards were present for workers in the immediate vicinity, workers on-site, the general public off-site, or the environment. and to evaluate the significance of any hazards that were found. A detailed accident analysis was performed to determine a set of limiting accidents that have potential for off-site consequences. One of the limiting accidents identified in the SAR was the rupture of a cylinder engulfed in a fire.
Date: June 1, 1997
Creator: Lombardi, D.A.; Williams, W.R. & Anderson, J.C.
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

Shipment and Storage Containers for Tritium Production Transportation Casks

Description: The need for a shipping and storage container for the Tritium production transportation casks is addressed in this report. It is concluded that a shipping and storage container is not required. A recommendation is made to eliminate the requirement for this container because structural support and inerting requirements can be satisfied completely by the cask with a removable basket.
Date: April 1998
Creator: Massey, W. M.
Partner: UNT Libraries Government Documents Department

Project W-443 cask/transporation project management plan

Description: This document has been prepared and is being released for Project W-443 participants to use in the performance of project activities. This PMP establishes the organizational responsibilities and baseline controls to be used to manage Spent Nuclear Fuel Subproject W-443.
Date: July 2, 1996
Creator: Byrd, L.C., Westinghouse Hanford
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

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

Description: This test report documents the successful inspection and testing to certify two NuPac PAS-1 casks in accordance with US Department of Energy Certificate of Compliance (CoC) USA/9184/B(U). The primary and secondary containment vessels of each cask met the acceptance criteria defined in the CoC and the test plan.
Date: October 13, 1998
Creator: MERCADO, J.E.
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

Stress analysis of closure bolts for shipping casks

Description: 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.
Date: January 1, 1993
Creator: Mok, G.C.; Fischer, L.E. (Lawrence Livermore National Lab., CA (United States)) & Hsu, S.T. (Kaiser Engineers, Oakland, CA (United States))
Partner: UNT Libraries Government Documents Department

Full scale impact testing for environmental and safety control of energy material shipping container systems

Description: Heavily-shielded energy material shipping systems, similar in size and weight to those presently employed to transport irradiated reactor fuel elements, are being destructively tested under dynamic conditions. In these tests, the outer and inner steel shells interact in a complex manner with the massive biological shielding in the system. Results obtained from these tests provide needed information for new design concepts. Containment failure (and the resulting release of radioactive material to the environment which might occur in an extremely severe accident) is most likely through the seals and other ancillary features of the shipping systems. Analyses and experiments provide engineering data on the behavior of these shipping systems under severe accident conditions and information for predicting potential survivability and environmental control with a rational margin of safety.
Date: January 1, 1978
Creator: Seagren, R.D.
Partner: UNT Libraries Government Documents Department

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

Description: 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).
Date: January 1, 1983
Creator: Bucholz, J.A.
Partner: UNT Libraries Government Documents Department

Transportation Business Plan

Description: The Transportation Business Plan is a step in the process of procuring the transportation system. It sets the context for business strategy decisions by providing pertinent background information, describing the legislation and policies governing transportation under the NWPA, and describing requirements of the transportation system. Included in the document are strategies for procuring shipping casks and transportation support services. In the spirit of the NWPA directive to utilize the private sector to the maximum extent possible, opportunities for business ventures are obvious throughout the system development cycle.
Date: January 1, 1986
Partner: UNT Libraries Government Documents Department

Thermal testing of a dry spent fuel cask. [NL1 /sup 1///sub 2/ legal weight truck cask]

Description: A nuclear spent fuel cask that employs air or an inert gas such as helium as the cooling medium in the fuel cavity is commonly known as a ''dry'' cask. The thermal characteristics of a dry cask differ considerably from a wet cask. A gas is a poorer heat transfer medium than water. As a result, the fuel and cask basket temperatures will be higher for a given fuel heat level. The response characteristics during operating transients will also differ. Detailed thermal tests on a dry cask were performed to verify design parameters and to obtain operational data for shipping and unloading. Results of a number of thermal tests performed on the NL1 /sup 1///sub 2/ legal weight truck cask are presented. The uniqueness of these tests are that they were performed using a full-scale geometric and thermal mockup of a nuclear fuel assembly. This results in more realistic fuel and basket temperatures being measured than those obtained using a simpler, single heater assembly.
Date: January 1, 1978
Creator: Anderson, R.T.
Partner: UNT Libraries Government Documents Department

Realistic assessment of a nuclear cask during a hypothetical railroad accident

Description: The study results indicate that blindly selecting the ''worst-possible-case assumptions'', and then postulating a radioactive material (RAM) release without identifiable mechanisms compounds unrealism and adds confusion. In order for a RAM release to occur, an unlikely series of events must first occur which breach the multiple containment barriers surrounding the fuel. The overall safety margin provided by the packaging equipment increases geometrically beyond the already adequate margin provided by each containment barrier. A case evaluation of the NL 10/24 packaging system illustrates this contention by showing that: (1) the accident events which must occur before a release of RAM, other than gases, is possible are in themselves incredible; and (2) the biological effects of a release of fission gases will in all liklihood be nil.
Date: May 1, 1978
Creator: Anderson, R.T.
Partner: UNT Libraries Government Documents Department

Nuclear cask testing films misleading and misused

Description: 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.
Date: October 1, 1991
Creator: Audin, L. (Audin (Lindsay), Ossining, NY (United States))
Partner: UNT Libraries Government Documents Department

Operational facets of a dry spent fuel cask. [Dryout and cooldown of NLI /sup 1///sub 2/ Legal Weight Truck cask]

Description: There are several operational activities unique to a ''dry'' spent fuel cask. During loading, it is necessary to rapidly displace water and dry out the fuel cavity. During unloading, it is necessary to ''cooldown'' the hot fuel and cask internals prior to placing the cask in a spent fuel pool. Techniques for rapidly and efficiently performing dryout and cooldown were developed for the NLI /sup 1///sub 2/ Legal Weight Truck (LWT) cask at the Barnwell Nuclear Fuel Plant (BNFP). The results of this testing are reported. The techniques developed can be utilized equally well for larger dry casks such as the NLI 10/24 rail cask. The test results indicated that these dry cask operations should not cause problems during loading and unloading in excess of that experienced with a wet cask. In fact, elimination of coolant sampling and the need to meet coolant activity limits is a distinct advantage.
Date: January 1, 1978
Creator: Anderson, R.T.
Partner: UNT Libraries Government Documents Department

Analysis, scale modeling, and full scale tests of a truck spent-nuclear-fuel shipping system in high velocity impacts against a rigid barrier

Description: The report describes analyses conducted to predict the response of a truck tractor-trailer system with a spent-nuclear-fuel shipping cask in very severe (98 to 135 kilometers per hour) head-on crashes into a rigid concrete structure. The analyses include both mathematical and physical scale modeling of the system. The results of the analyses are compared to the results of instrumented full-scale tests conducted as the last step in the research program described in the report.
Date: April 1, 1978
Creator: Huerta, M.
Partner: UNT Libraries Government Documents Department