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At least 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Comparison of German and American fabrication standards for transportation and storage containments of high level radioactive material type B packages - 15055

As the competent authority for the evaluation of German radioactive material (RAM) transport and storage packages, BAM (the German Federal Institute for Materials Research and Testing) has evaluated various international cask manufacturers regarding compliance with the German Type B(U) licensing requirements. The US NRC has recommended paragraphs in the ASME Boiler and Pressure Vessel (B and PV) Code, Section III, Division 3, as the measure to quantify the safety requirements of the US regulations regarding fabrication of RAM transportation and storage containments. The objective of this study is to compare the fabrication criteria of the German Dangerous Goods Procedural Rules (BAM-GGR) and the German Institute for Standardization (DIN) with the US quantitative measures for fabrication of RAM package containment vessels. This comparison will cover major similarities and differences between the requirements of the German standards and the ASME Code requirements for Type B package containment fabrication QC (quality control). While the material most often used for German RAM containment applications is ductile cast iron (DCI), in the US forged austenitic stainless steel is typically used. This study focuses on fabrication QC measures required to ensure reproducible and acceptable properties for DCI containments. (authors)

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DOE OSTI2024

STRUCTURAL MODELING TO SUPPORT POST-YIELD ACCEPTANCE CRITERIA FOR SPENT NUCLEAR FUEL CLADDING

Spent nuclear fuel (SNF) is evaluated for structural failure during storage and transportation scenarios. The U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) program has sponsored significant research in quantifying mechanical loads on SNF during storage and transportation scenarios using experimental and modeling methods. The SFWST program has also performed significant research on measuring the mechanical behavior of irradiated SNF as defueled cladding segments and cladding with fuel pellets to measure composite behavior. This paper considers some of the key material data from the Sibling Pin testing and uses structural modeling and analysis methods that have been informed by testing to consider post-yield acceptance criteria for SNF cladding structural analysis. Test data published by Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL) are the foundation for informing the material behavior of the models developed in this study. In particular, four-point bend (4PB) tests of fueled and defueled cladding segments provide significant information about the bending failure mode of SNF. ORNL’s 4PB test data is on fueled cladding segments, so the composite behavior of SNF is demonstrated. This paper describes PNNL’s coincident beam model that was developed to approximate the composite behavior of SNF. This paper also presents PNNL’s structural dynamic finite element models of a cask tip-over scenario, which is predicted to cause the strongest mechanical loads on SNF of all postulated storage and transportation scenarios. SNF bending loads predicted in the cask tip-over scenario and cladding acceptance criteria beyond yield are considered, with justification based on the Sibling Pin test data. ASME Boiler and Pressure Vessel code stress intensity limits are also considered. The ultimate goal of this work is to aid in the justification of structural acceptance criteria for SNF cladding beyond the cladding’s irradiated yield strength for use in structural analysis of all storage and transportation scenarios.

Klymyshyn, Nicholas A.↗
DOE OSTI2015

A Review of multipurpose canister concepts for standardization in the waste management system - 15563

Since the early 1980's, the federal government and individual utilities have put forth considerable effort to develop a canister for spent nuclear fuel (SNF) that could be used as part of a dry storage system, transportation system, and disposal system without having to be reopened. While the terminology for this type of canister has changed over the years, the underlying concept and purpose have not. These efforts and ongoing activities in this area are highlighted, expanding upon several previous works. During development of dry storage systems in the 1980's, there was growing interest in triple-purpose (storage, transportation, and disposal) canister concepts, and preliminary studies were conducted. That interest evolved into a project to develop a multipurpose canister (MPC) in the 1990's. After the de-funding of that project, the concept lived on in a canister system for US DOE-owned fuel from the mid 1990's to the early 2000's. The commercial spent fuel version of the concept resurfaced in the mid-2000's as the transportation aging and disposal (TAD) canister. After the halting of the Yucca Mountain project, the specific implementation of triple-purpose canisters was reevaluated. This re-evaluation resulted in the standardized transportation aging and disposal (STAD) canister conceptual design and evaluation effort. Each of these efforts is reviewed and summarized. (authors)

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DOE OSTI2019

Unlocking the Vault - Accelerated Delivery Through the Application of Innovation and Collaboration - 19011

The Pile Fuel Cladding Silo (PFCS) at Sellafield was built in 1949/50 to receive and store fuel cladding from reprocessing of Windscale Pile fuel; it has been used to store this material and other waste from the UK nuclear programme ever since. The Silo is unsuitable for indefinite storage; this along with uncertainty in the make-up and condition of the inventory combine to rank the facility alongside the most hazardous in Western Europe. Sellafield Ltd, began working with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS - a joint venture between Bechtel Inc and Cavendish Nuclear Ltd) in 2012, to execute a suite of projects to create an access route to the waste and a capability to retrieve waste from the silo for the first time ever. However, the challenge of providing this capability, while maintaining ongoing safe storage of the waste, and underpinning future performance of both the silo and the waste packages for many hundreds of years, proved to be too much for traditional Sellafield Ltd delivery models. With both cost and schedule forecasts doubling from initial estimates, a new approach was clearly required. The team recognised that success would require all of the experience and expertise of each organisation working together to a common goal. Therefore, significant effort was expended to build a collaborative team environment into which an innovative approach to both the engineering of what was provided and the delivery methods for doing it could be launched. Focussing on collaborative culture in which innovation and fit for purpose thinking were encouraged has completely rebuilt stakeholder confidence. The project is now almost complete; 17 months ahead of schedule and British Pounds 100 m less than the approved cost. (authors)

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DOE OSTI2021

Management of the Three Mile Island, Unit 2, Accident Corium and Severely Damaged Fuel Debris

The Three Mile Island, Unit Two (TMI-2) pressurized water reactor core melted down in 1979 due to an untimely combination of maintenance problems that led to a loss of feedwater, followed by a series of operational misunderstandings and errors. The melted core recovery process required development of a wide array of tools. After approximately three years of water management and other cleanup actions, the first views of the core revealed a much higher degree of damage than previously expected. Approximately 62 metric tons of the core had melted, leaving only 42 of the 177 fuel assemblies standing with a majority of rods intact. The core to be recovered was composed of loose debris and a solidified mass of formerly molten fuel. Robotic tools that had been designed for the task were of limited effectiveness due to the range of material types and phases. Over a period of several years, the central melt was broken up, primarily by use of a core-boring drill originally designed to acquire samples through the depth of the debris field. The broken pieces were loaded into specially designed debris canisters by use of long-handled pick-and-place tools and suctioned into baffled knockout canisters using an airlift vacuum system. A remotely operable underwater plasma arc torch was developed for removal of the lower core support structure to be able to retrieve the fuel pieces and secondary melt that had accumulated on the bottom reactor vessel head. Canister design played a central role in defining the initial retrieval process and later affected transportation, interim wet and its current interim dry storage. Due to concerns about the potential for radiolysis of residual water in the debris and other canister material, the canisters were fitted with hydrogen recombiner catalyst units to prevent a buildup of flammable gas and potential pressurization. To confirm the effectiveness of the recombiners during shipping, eight dewatered canisters were kept sealed for up to 205 days with periodic sampling of the headspace gas. The highest hydrogen value (9 vol%) was observed in a canister held for 147 days while the longest stored canister resulted in a 5% hydrogen concentration. The oxygen concentration never exceeded 0.5 %, and the primary backfill was >80% argon, meaning there was no flammability risk. Radiolytic hydrogen was also observed in wet pool storage, where the vented canisters discharged a portion of the water backfill as a result of gas production. The 344 canisters (268 fuel debris, 62 filter, and 12 knockout type) were dewatered, loaded in groups of seven into a double barrier shipping cask, and transported to the U.S. Department of Energy site in Idaho for 10 years of pool storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermal Modeling of a Loaded Magnastor Storage System at Catawba Nuclear Station

This report represents fulfillment of the M2 Milestone M2FT-16PN080203031, “High Heat Load Thermal Analyses,” under Work Package FT-16PN08020303. The report documents the analysis of the NAC International, Inc. (NAC) Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) module at Duke’s Catawba Nuclear Station. As part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, a high-burnup fuel storage demonstration is planned for a storage module in the North Anna Nuclear Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an AREVA TN-32B High Burnup cask. The main goals of this proposed test are to provide confirmatory data for model validation and potential improvement, support license renewals and new licenses for ISFSIs, and support transportation licensing for high burnup spent nuclear fuel (EPRI 2014). The focus of the Demonstration test is the performance of the high-burnup fuel. To support the High Burnup Spent Fuel Data Project, Pacific Northwest National Laboratory (PNNL) staff performed a detailed thermal analysis (Fort et al. 2016) of the TN-32B cask. It was found that when conservatisms were removed from the thermal models and especially from the calculated decay heats, the predicted peak cladding temperatures (PCTs) were significantly below the 400°C (752°F) regulatory guidance limit, as specified in Interim Staff Guidance 11, Revision 3, (U.S. NRC 2003).

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DOE OSTI2016

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Supplement Analysis for the Proposed Shipment of High Burnup Research Cask to Idaho National Laboratory for Research and Development Purposes

DOE needs to continue with the HBUSFDP and transfer the HBURC to a facility capable of examination and testing of HBU SNF. These types of facilities do not exist at NAPS but are available at DOE national laboratories. The removal and eventual opening of the TN-32B research cask will provide DOE with the opportunity to gather data that will support continued safe storage of SNF at existing nuclear power plants, or a future federal consolidated storage facility. This effort will also help build public trust and confidence by demonstrating DOE’s ability to safely and securely transport commercial SNF using the certified Atlas rail consist.

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DOE OSTI2018

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

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DOE OSTI2019

A probabilistic assessment of PWR SNF cladding pinching failure considering potential transportation accident conditions

This report presents a probabilistic assessment of pressurized water reactor (PWR) spent nuclear fuel (SNF) rod cladding pinching failure triggered by a transportation accident, after 300 years of dry storage. Three thousand pinching scenarios, which were previously evaluated assuming a baseline 9-m cask drop and a horizontal impact on an unyielding surface, are reassessed assuming more realistic train transportation accidents. Each scenario was defined using Latin Hypercube Sampling (LHS) of SNF rod conditions and loading parameters, and it is simulated with a nonlinear finite element (FE) model. Cladding failure is evaluated under consideration of hydride-related embrittlement effects. The reassessment indicates that cladding pinching failure is not significantly affected by the assumptions on transportation accidents, but is rather controlled by other parameters, mainly the pellet-cladding gap width. The reevaluated, conditional failure probability ranges from 1.62 E-7 to 6.47 E-7, for 5,000 km of transportation by train.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Insert Modeling in UNF ST&DARDS

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) is a software tool that integrates a used nuclear fuel (UNF) or spent nuclear fuel (SNF) relational database and key analysis capabilities to simplify and automate numerous UNF management and fuel cycle–related activities. UNF-ST&DARDS is being developed for the US Department of Energy’s Office of Nuclear Energy Spent Fuel and Waste Disposition program. UNF-ST&DARDS provides an integrated framework that uses advanced modeling and simulation to predict the behavior of SNF over the timescales associated with permanent disposal in a geologic repository. After leaving the spent fuel pool, SNF is transferred to dry storage in a dual-purpose canister (DPC). DPCs are considered “dual purpose” because they are designed for both storage and transportation, removing the need to transfer the fuel to a separate transportation cask. However, much research has been conducted investigating the feasibility of directly disposing of DPCs in geologic repositories. Direct disposal of DPCs could reduce worker exposure during repackaging, reducing the amount of low-level waste from the discarded DPCs and potentially saving billions of dollars. Therefore, direct disposal of as-loaded DPCs is desirable if it can be done safely.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Analysis of Gas Samples Taken from the High Burnup Demonstration Cask

The High Burn-Up Demonstration Project was recently initiated by the Department of Energy (DOE) to evaluate the effects of fuel drying and long term dry storage on high burn-up spent nuclear fuel. As part of the project, samples of the He backfill gas were collected 5 hours, 5 days, and 12 days after completion of drying. The samples provide information on the state of the fuel at closure, and on the environment within the cask. At Sandia National Laboratories, the samples were analyzed by gamma-ray spectroscopy to quantify fission product gases and by gas mass spectrometry to quantify bulk and trace gases; water content was measured via humidity probe. Gamma-ray spectroscopy results indicated no detectible 85 Kr, indicating no failed fuel rods were present after drying. Mass spectrometry indicated build-up of CO 2 to 930 ppmv over two weeks, attributed to oxidation of organic compounds (possibly vacuum grease or vacuum pump oil) within the cask. H 2 , generated by either radiolysis or metal corrosion, also increased up to —500 ppmv. Water contents in the cask were higher than anticipated, increasing to —17,400 ppmv ±10% after 12 days. Measuring water content proved challenging, and possible improvements to the method for future analyses are proposed.

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DOE OSTI2019

High-Burnup Demonstration: Thermal Modeling of TN-32B Vacuum Drying and ISFSI Transients

This report describes thermal modeling done for the High-Burnup Spent Fuel Data Project that is being performed by the U.S. Department of Energy’s Spent Fuel and Waste Science and Technology research and development program. The purpose of this project is to investigate the performance of high-burnup spent nuclear fuel in dry storage. Part of this project is a demonstration test being performed with a storage module in the North Anna Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an Orano TN-32B High-Burnup cask which has been loaded with fuel assemblies from the North Anna spent fuel pool.

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DOE OSTI2015

French management model for sustainable transportation of used fuel - 15376

Safe and efficient transportation of radioactive material is key to the success of our industry. Used fuel transports are inherently complex due to the nature of the material and the high visibility. Only a few countries have developed comprehensive used fuel transportation systems that require a wide range of equipment, services and expertise. In the many countries, such as the Unites States, where interim consolidated storage, recycling or geological repository are not available yet, similar systems will have to be implemented. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation system that was initially developed in the 1970's has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous 'lessons learned' have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. Based on AREVA TN extensive expertise, this paper will describe the different phases and milestones that need to be met to set up, license and operate a successful used fuel transportation management system. Transportation of spent fuel in France could not be performed effectively and efficiently without strong collaboration with Electricite de France (EDF) who plays a critical role as the shipper. EDF will provide a nuclear operator's perspective on the most critical aspects of a successful used fuel transportation program. (authors)

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DOE OSTI2018

Post-Irradiation Examinations of High Burnup PWR Fuel Rods - Initial Results

The High Burnup Spent Fuel Data Project (herein referred to as the “project”), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 “sister rods” that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE).

Bevard, Bruce Balkcom↗
DOE OSTI2019

Contamination Control: Challenges and Innovative Design Solutions utilized at Chalk River Laboratories - 19263

CNL has operated research reactors at the Chalk River Laboratories (CRL) site since 1947, generating a variety of spent nuclear fuel wastes produced by irradiating metallic uranium and other prototype fuels. This legacy waste has been stored in below-grade, vertical cylindrical steel and concrete structures known as Tile Holes. Ongoing monitoring of the legacy waste has identified moisture resulting from incomplete drying prior to emplacement, surface water and humidity. The undesirable storage conditions result in corrosion of both the legacy waste and the storage containers. Continued storage in these conditions may result in further degradation of the storage container and contamination of the surrounding environment. In 2004, the Fuel Packaging and Storage (FPS) Project was initiated, with the objective to develop a complete system to safely retrieve legacy waste from these Tile Holes as well as provide a stable storage environment until a final long-term management plan is established. Contamination control during retrieval, transfer and repackaging operations is critical for personnel and environmental safety. The significant challenges with retrieval of the legacy waste from the Tile Holes includes management of loose collateral liquid and solids that will be brought into the system, with the retrieved waste container. During the retrieval, transfer and repackaging operation, the challenge was to: 1) Prevent an environmental release by potentially transferring these contaminants onto the external surface of the Tile Hole Array; 2) Prevent an environmental release and ensure personnel safety while transporting the legacy waste; 3) Control release of contaminants into the new clean storage container while preventing contamination of the packaging equipment. To address the first source of contamination, a debris guide liner capable of interfacing with multiple Tile Hole configurations was integrated within the Retrieval Flask to limit contamination of the flask interior and to direct collateral liquids and solids back into the Tile Hole. The system uses a multi-function sealed Shield Gate and Drip Tray to retain and prevent the contaminants from escaping while transferring the legacy waste. This gate, along with a thin walled, Disposable Debris Guide Liner (DDGL) is an innovative solution developed to control the flow characteristics of the residual waste and to contain the waste to protect the internal components of the repackaging equipment. The development of the complete solution was further challenged as the DDGL must not tear or rip and must compact inside the new storage containers for a range of legacy waste storage containers. (authors)

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DOE OSTI2023

Assessment of ITER radiation environment during the remote-handling operation of In-Vessel components with D1SUNED

During ITER operational life, a remote-handled cask will be used to transfer In-Vessel components to the Hot Cell for maintenance, storage and decommissioning purposes. Due to the distribution of penetrations for system allocation in the facility, the radiation field of each transfer operation presents a high spatial variability; all operations must be studied independently for workers and electronics protection. In this paper, we present a fully representative approach to describe the radiation environment during the complete remote-handling scenario of In-Vessel components in the ITER facility. The impact of all relevant radiation sources during different stages of the operation is addressed. As-built structures and 2020 baseline designs are considered to produce the most detailed neutronics model of the Tokamak Complex, the 400,000-tonne civil structure hosting the tokamak, up to date. Novel capabilities of the D1SUNED code have allowed to compute the integral dose, the dose rate and the photon-induced neutron flux of both moving and static radiation sources. Time bins are included in the simulations to compute the dose rate caused by In-Vessel components at all positions along the transfer. The time evolution of the dose rate is built in video format with a 1-m resolution, especially valuable for hot-spots identification.

42 ENGINEERING↗
DOE OSTI2019

Dry Cask Shipments of ATR Spent Nuclear Fuel at INL

The High Load Charger (HLC) cask will be replacing the Advanced Test Reactor (ATR) spent fuel transfer cask that has been used at Idaho National Laboratory (INL) to wet-transfer spent ATR fuel elements from ATR to Idaho Nuclear Technology and Engineering Center (INTEC) in order to support future plans at both facilities. INTEC intends to discontinue use of the CPP-666 storage pools in compliance with the Idaho Settlement Agreement and the basins will no longer be available to store spent ATR fuel elements. In order to support the core-internals change out in 2021 for ATR, spent ATR fuel elements must be removed from the canal and shipped to INTEC. INL contractors will use the HLC cask for safe, dry transfer of spent Advanced Test Reactor (ATR) fuel elements between the ATR Complex managed by Battelle Energy Alliance, LLC and the Idaho Nuclear Technology and Engineering Center (INTEC) managed by Fluor Idaho, LLC. The HLC cask (also known as the High Load Charging Cask or High Load Charger) is a lead-shielded, stainless steel transfer container originally designed in 1954 by the Idaho Nuclear Corporation to meet the radiation and contamination controls requirements of the time for shipment of spent nuclear fuel. The cask has been modified over the years and is currently used for wet-to-dry transfers at INTEC. It is approved only for out-of-commerce transfers. The cask will arrive at the ATR where eight spent ATR fuel elements will be loaded into two, newly-designed buckets. Each bucket is a quarter section, right-cylinder shape that has four compartments. The cask will be lifted out of the ATR canal allowing most of the water to drain from the cask. Processing the 12,500 kg cask will be completed in ATR and then transported on a tractor trailer to INTEC where the loaded buckets will be removed from the cask. The loaded buckets will undergo a drying process before placing the four buckets into a canister for long-term storage in the Irradiated Fuel Storage Facility (IFSF) at INTEC. A new transportation plan that was submitted to, and approved by, the U.S. Department of Energy to provide the cask safety basis, which describes the hazard analysis and accident analysis for the cask during transport. Supporting accident analysis include criticality, thermal, dose, and drop analyses. Of particular interest for this cask was the drop analysis from which was derived the speed limit for the transport. The drop analysis addressed drops in air onto an unyielding surface and in water onto an energy-absorbing pad for the cask in the ATR facility. These analyses are typically applied as the bounding drop scenarios during transport; however, this resulted in a safety basis speed limit of 12.9 kph. In order to more accurately represent the in-transit accident scenario, LS-DYNA software was procured and the model was updated to more closely represent the potential drop heights and impact surfaces of asphalt and soil for the cask while in transit. This analysis resulted in a safety basis speed limit of 48 kph. This safety basis speed limit is more practical to implement and control.

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