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Title: TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

Abstract

The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 [1]. Both the International Atomic Energy Agency regulations for transporting radioactive materials [2, paragraph 727], and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 [3], for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask [3] for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS® [4] and LS DYNA™ [5] finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFRmore » 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event.« less

Authors:
; ; ;
Publication Date:
Research Org.:
Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
951568
Report Number(s):
PNNL-SA-54285
401001050; TRN: US0902220
DOE Contract Number:
AC05-76RL01830
Resource Type:
Journal Article
Resource Relation:
Journal Name: Packaging, Transport, Storage and Security of Radioactive Material, 18:11-18; Journal Volume: 18
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 36 MATERIALS SCIENCE; CASKS; CLOSURES; CONTAINMENT; FASTENERS; FLANGES; HYPOTHETICAL ACCIDENTS; IAEA; LICENSING; NUCLEAR FUELS; ORIENTATION; RADIOACTIVE MATERIALS; REGULATIONS; SPENT FUELS; STEELS; TRANSPORT

Citation Formats

Shah, M. J., Klymyshyn, Nicholas A., Adkins, Harold E., and Koeppel, Brian J. TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events. United States: N. p., 2007. Web. doi:10.1179/174651007X191125.
Shah, M. J., Klymyshyn, Nicholas A., Adkins, Harold E., & Koeppel, Brian J. TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events. United States. doi:10.1179/174651007X191125.
Shah, M. J., Klymyshyn, Nicholas A., Adkins, Harold E., and Koeppel, Brian J. 2007. "TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events". United States. doi:10.1179/174651007X191125.
@article{osti_951568,
title = {TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events},
author = {Shah, M. J. and Klymyshyn, Nicholas A. and Adkins, Harold E. and Koeppel, Brian J.},
abstractNote = {The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 [1]. Both the International Atomic Energy Agency regulations for transporting radioactive materials [2, paragraph 727], and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 [3], for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask [3] for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS® [4] and LS DYNA™ [5] finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event.},
doi = {10.1179/174651007X191125},
journal = {Packaging, Transport, Storage and Security of Radioactive Material, 18:11-18},
number = ,
volume = 18,
place = {United States},
year = 2007,
month = 3
}
  • The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (CFR), Title 10, Part 71 [1]. Both the International Atomic Energy Agency (IAEA) regulations for transporting radioactive materials [2, paragraph 727], and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop impact event on a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC-certified transportation casks, the HI-STAR 100 [3], for drop impactmore » events. The specific area examined is the behavior of the bolted connections in the “overpack” top flange and the closure plate, which are significantly loaded during the hypothetical drop impact event. The term “overpack” refers to the cask that receives and contains a sealed multi-purpose canister (MPC) containing spent nuclear fuel. The analytical work to evaluate the NRC-certified HI-STAR 100 spent fuel transport cask [3] for a 9-m (30-ft) drop impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS® [4] and LS DYNA™ [5] finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free-drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the closure plate surfaces may occur, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event.« less
  • Interim storage plants for spent fuel elements, based on dry storage technology in transport casks, are planned in the Federal Republic of Germany. The casks are arranged in storage buildings. The decay heat is removed from the outer cask surfaces by natural convection of air entering the building through openings in the walls, and leaving through outlets in the roof. As the differential equations describing the complex three-dimensional flow and temperature field can only be solved for simple boundary conditions, experiments were conducted using scaled-down models of the casks and the building. The relevant similarity conditions have been investigated andmore » used for design and operation of the 1:5 scale test setup. The cask models were heated electrically. Cask temperatures, air temperatures, as well as air flow and velocities, were measured. It was found that the cooling conditions at the different cask positions show very small differences and that the cask surface temperatures inside the building are a maximum of 10/sup 0/C higher than on a free-standing cask.« less
  • The three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, is applied to the analysis of a series of simple geometry benchmark experiments and prototypic spent-fuel storage cask measurements. The simple geometry experiments were performed in Japan and at the General Electric-Morris Operation facility; the cask measurements were performed at the Idaho National Engineering Laboratory. A total of five storage cask problems and two simple geometry problems were analyzed to determine the expected accuracies of computational analyses using well-established source-generation and Monte Carlo codes. The general trends seen in this work are in agreement withinmore » 30% or better with the measurements for neutron dose rates along the ask side, lid, and bottom. The gamma-ray dose rates with substantial contributions from the top endfitting, plenum, and bottom endfitting regions also are in good agreement. based on the latest results, gamma-ray dose rate calculations with major contributions due to the active fuel region show a consistent factor of 1.6 overprediction of the measured quantities for casks with iron and concrete shields. Major uncertainties exist in the quantification of {sup 59}Co concentrations in endfitting hardware materials. The results presented support the accuracy of source generation methods and dose estimation methods in these regions given accurate impurity characterizations. Thus, it is felt that the practice of using upper bounds for {sup 59}Co initial concentrations should ensure conservative cask designs.« less
  • The Department of Energy (DOE) has developed a design concept for a set of standard canisters for the handling, interim storage, transportation, and disposal in the national repository, of DOE spent nuclear fuel (SNF). The standardized DOE SNF canister has to be capable of handling virtually all of the DOE SNF in a variety of potential storage and transportation systems. It must also be acceptable to the repository, based on current and anticipated future requirements. This expected usage mandates a robust design. The canister design has four unique geometries, with lengths of approximately 10 feet or 15 feet, and anmore » outside nominal diameter of 18 inches or 24 inches. The canister has been developed to withstand a drop from 30 feet onto a rigid (flat) surface, sustaining only minor damage - but no rupture - to the pressure (containment) boundary. The majority of the end drop-induced damage is confined to the skirt and lifting/stiffening ring components, which can be removed if de sired after an accidental drop. A canister, with its skirt and stiffening ring removed after an accidental drop, can continue to be used in service with appropriate operational steps being taken. Features of the design concept have been proven through drop testing and finite element analyses of smaller test specimens. Finite element analyses also validated the canister design for drops onto a rigid (flat) surface for a variety of canister orientations at impact, from vertical to 45 degrees off vertical. Actual 30-foot drop testing has also been performed to verify the final design, though limited to just two full-scale test canister drops. In each case, the analytical models accurately predicted the canister response.« less