Mechanical Shock and Vibration Analysis of Spent Nuclear Fuel Carried by the Atlas Railcar

Author(s):  
Nicholas Klymyshyn ◽  
Kevin Kadooka ◽  
Pavlo Ivanusa ◽  
Casey Spitz

Abstract Researchers at Pacific Northwest National Laboratory have completed a structural-dynamic analysis of spent nuclear fuel subjected to the mechanical shock and vibration environment that is anticipated during normal conditions of transport in casks carried by the Atlas railcar. The Atlas railcar is a new railcar design that is being developed specifically for the purpose of carrying spent nuclear fuel casks. The analysis used best-estimate railcar dynamics models of the Atlas railcar and considered 17 different spent nuclear fuel transportation cask systems, representing the current fleet of cask options. This work used NUCARS, a specialized railcar dynamics explicit finite element code to calculate railcar dynamic response to prescribed speeds and track configurations. The railcar dynamics models provided cask transient motion for a wide range of speeds and track conditions, generating a relatively large database of potential cask motion. All of the cask motion transients were then applied as loading conditions to LS-DYNA structural-dynamic models of a single fuel rod. The analyses predict that the Equipos Nucleares S.A./U.S. Department of Energy (ENSA/DOE) multimodal transportation test of 2017 provided a relatively stronger vibration environment than is expected from the Atlas railcar. This paper describes the analysis methods, the analysis results, and compares the results of the Atlas transportation analysis to the test results and analyses of the ENSA/DOE multimodal transportation test of 2017.

Author(s):  
Kevin Kadooka ◽  
Nicholas Klymyshyn

Abstract The primary mode of spent nuclear fuel transportation within the United States will be by railcar. One such system is the Atlas railcar, which is designed to transport 17 different spent nuclear fuel cask systems, including bare fuel systems and canister fuel systems. In the latter configuration, multipurpose canisters containing spent nuclear fuel may be placed within an overpack for storage, or within a cask for transportation. Compared to bare fuel systems, canister fuel systems have additional degrees of freedom for motion during transportation, because clearance between the cask and canister allows for some motion of the canister to occur relative to the cask. This work investigates the effect of canister motion on the shock and vibration imparted to the spent nuclear fuel within. Structural dynamic analyses have been conducted to identify the effects of canister to cask clearance, presence and type of dunnage, and loading direction and frequency. This modeling study calculates anticipated cask motion, canister motion, and spent nuclear fuel structural dynamic response to normal conditions of transportation railcar motion using finite element analysis methods that were developed to model the rail segment of the ENSA/DOE (Equipos Nucleares S.A., U.S. Department of Energy) multimodal transportation test of 2017.


2021 ◽  
Author(s):  
Nicholas Klymyshyn ◽  
Kevin Kadooka ◽  
Pavlo Ivanusa ◽  
Casey Spitz

Author(s):  
Mikal A. McKinnon ◽  
Leroy Stewart

Abstract Research studies by the Electric Power Research Institute (EPRI) established the technical and operational requirements necessary to enable the onsite cask-to-cask dry transfer of spent nuclear fuel. Use of the dry transfer system has the potential to permit shutdown reactor sites to decommission pools and provide the capability of transferring assemblies from storage casks or small transportation casks to sealed transportable canisters. Following an evaluation by the Department of Energy (DOE) and the National Academy of Sciences, a cooperative program was established between DOE and EPRI, which led to the cost-shared design of a dry transfer system (DTS). EPRI used Transnuclear, Inc., of Hawthorne, New York, to design the DTS in accordance with the technical and quality assurance requirements of the code of Federal Regulations, Title 10, Part 72 (10CFR72). EPRI delivered the final design report to DOE in 1995 and the DTS topical safety analysis report (TSAR) in 1996. DOE submitted the TSAR to the United States Nuclear Regulatory Commission (NRC) for review under 10CFR72 and requested that the NRC staff evaluate the TSAR and issue a Safety Evaluation Report (SER) that could be used and referenced by an applicant seeking a site-specific license for the construction and operation of a DTS. DOE also initiated a cold demonstration of major subsystem prototypes in 1996. After careful assessment, the NRC agreed that the DTS concept has merit. However, because the TSAR was not site-specific and was lacking some detailed information required for a complete review, the NRC decided to issue an Assessment Report (AR) rather than a SER. This was issued in November 2000. Additional information that must be included in a future site-specific Safety Analysis Report for the DTS is identified in the AR. The DTS consists of three major sections: a Preparation Area, a Lower Access Area, and a Transfer Confinement Area. The Preparation Area is a sheet metal building where casks are prepared for loading, unloading, or shipment. The Preparation Area adjoins the Lower Access Area and is separated from the Lower Access Area by a large shielded door. The Lower Access Area and Transfer Confinement Area are contained within concrete walls approximately three feet thick. These are the areas where the casks are located and where the fuel is moved during transfer operations. A floor containing two portals separates the Lower Access Area and the Transfer Confinement Area. The casks are located below the floor, and the fuel transfer operation occurs above the floor. The cold demonstration of the DTS was successfully conducted at the Idaho National Engineering and Environmental Laboratory (INEEL) as a cooperative effort between the DOE and EPRI. The cold demonstration was limited to the fuel handling equipment, the cask lid handling equipment, and the cask interface system. The demonstration included recovery operations associated with loss of power or off-normal events. The demonstration did not include cask receiving and lid handling; cask transport and lifting; vacuum/inerting/leak test; canister welding; decontamination; heating, ventilation, and air conditioning; and radiation monitoring. The demonstration test was designed to deliberately challenge the system and determine whether any specific system operation could adversely impact or jeopardize the operation or safety of any other function or system. All known interlocks were challenged. As in all new systems, there were lessons learned during the operation of the system and a few minor modifications made to ease operations. System modifications were subsequently demonstrated. The demonstration showed that the system operated as expected and provided times for normal fuel transfer operations. The demonstration also showed that recovery could be made from off-normal events.


Author(s):  
William H. Lake ◽  
Nancy Slater-Thompson ◽  
Ned Larson ◽  
Franchone Oshinowo

Technology development activities are being conducted by the Department of Energy, Office of Civilian Radioactive Waste Management to support spent nuclear fuel and high-level radioactive waste transport to the federal repository at Yucca Mountain, Nevada in 2010. The paper discusses the motivation for pursuing transport technologies for a private sector operated transportation program, and describes some of the current technologies being pursued.


Author(s):  
Nicholas Klymyshyn ◽  
Pavlo Ivanusa ◽  
Kevin Kadooka ◽  
Casey Spitz

Abstract In 2017, the United States Department of Energy (DOE) collaborated with Spanish and Korean organizations to perform a multimodal transportation test to measure shock and vibration loads imparted to used nuclear fuel (UNF) assemblies. This test used real fuel assembly components containing surrogate fuel mass to approximate the response characteristics of real, irradiated used nuclear fuel. Pacific Northwest National Laboratory was part of the test team and used the data collected during this test to validate numerical models needed to predict the response of real used nuclear fuel in other transportation configurations. This paper summarizes the modeling work and identifies lessons learned related to the modeling and analysis methodology. The modeling includes railcar dynamics using the NUCARS software code and explicit dynamic finite element modeling of used nuclear fuel cladding in LS-DYNA. The NUCARS models were validated against railcar dynamics data collected during captive track testing at the Federal Railroad Administration’s Transportation Technology Center in Pueblo, CO. The LS-DYNA models of the fuel cladding were validated against strain gage data collected throughout the test campaign. One of the key results of this work was an assessment of fuel cladding fatigue, and the methods used to calculate fatigue are detailed in this paper. The validated models and analysis methodologies described in this paper will be applied to evaluate future UNF transportation systems.


Author(s):  
Spencer D. Snow ◽  
D. Keith Morton ◽  
Tommy E. Rahl ◽  
Robert K. Blandford ◽  
Thomas J. Hill

The National Spent Nuclear Fuel Program (NSNFP) at the Idaho National Engineering and Environmental Laboratory (INEEL) prepared four representative Department of Energy (DOE) spent nuclear fuel (SNF) canisters for the purpose of drop testing. The first two canisters represented a modified 24-inch diameter standardized DOE SNF canister and the second two canisters represented the Hanford Multi-Canister Overpack (MCO). The modified canisters and internals were constructed and assembled at the INEEL. The MCO internal weights were fabricated at the INEEL and assembled into two MCOs at Hanford and later shipped to the INEEL for drop test preparation. Drop testing of these four canisters was completed in August 2004 at Sandia National Laboratories. The modified canisters were dropped from 30 feet onto a flat, essentially unyielding surface, with the canisters oriented at 45 degrees and 70 degrees off-vertical at impact. One representative MCO was dropped from 23 feet onto the same flat surface, oriented vertically at impact. The second representative MCO was dropped onto the flat surface from 2 feet oriented at 60 degrees off-vertical. These drop heights and orientations were chosen to meet or exceed the Yucca Mountain repository drop criteria. This paper discusses the comparison of deformations between the actual dropped canisters and those predicted by pre-drop and limited post-drop finite element evaluations performed using ABAQUS/Explicit. Post-drop containment of all four canisters, demonstrated by way of helium leak testing, is also discussed.


Author(s):  
D. Keith Morton ◽  
Spencer D. Snow ◽  
Tom E. Rahl ◽  
Tom J. Hill ◽  
Richard P. Morissette

The Department of Energy (DOE) has developed a set of containers for the handling, interim storage, transportation, and disposal in the national repository of DOE spent nuclear fuel (SNF). This container design, referred to as the standardized DOE SNF canister or standardized canister, was developed by the Department’s National Spent Nuclear Fuel Program (NSNFP) working in conjunction with the Office of Civilian Radioactive Waste Management (OCRWM) and the DOE spent fuel sites. This canister had to have a standardized design yet be capable of accepting virtually all of the DOE SNF, be placed in a variety of storage and transportation systems, and still be acceptable to the repository. Since specific design details regarding the storage, transportation, and repository disposal of DOE SNF were not finalized, the NSNFP recognized the necessity to specify a complete DOE SNF canister design. This allowed other evaluations of canister performance and design to proceed as well as providing standardized canister users adequate information to proceed with their work. This paper is an update of a paper [1] presented to the 1999 American Society of Mechanical Engineers (ASME) Pressure Vessels and Piping (PVP) Conference. It discusses recent progress achieved in various areas to enhance acceptance of this canister not only by the DOE complex but also fabricators and regulatory agencies.


Author(s):  
Johan Andersson ◽  
Kristina Skagius ◽  
Anders Winberg ◽  
Anders Stro¨m ◽  
Tobias Lindborg

The Swedish Nuclear Fuel and Waste Management Co., SKB, is currently finalizing its surface based site investigations for the final repository for spent nuclear fuel in the municipalities of O¨sthammar (the Forsmark area) and Oskarshamn (the Simpevar/Laxemar area). The investigation data are assessed into a Site Descriptive Model, constituting a synthesis of geology, rock mechanics, thermal properties, hydrogeology, hydrogeochemistry, transport properties and a surface system description. Site data constitute a wide range of different measurement results. These data both need to be checked for consistency and to be interpreted into a format more amenable for three-dimensional modeling. The three-dimensional modeling (i.e. estimating the distribution of parameter values in space) is made in a sequence where the geometrical framework is taken from the geological models and in turn used by the rock mechanics, thermal and hydrogeological modeling. These disciplines in turn are partly interrelated, and also provide feedback to the geological modeling, especially if the geological description appears unreasonable when assessed together with the other data. Procedures for assessing the uncertainties and the confidence in the modeling have been developed during the course of the site modeling. These assessments also provide key input to the completion of the site investigation program.


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