Results for “certification testing” — Page 13

DOE OSTI2022

An Introduction to Los Alamos National Laboratory [Slides]

The Weapon Systems Engineering Division (W) provides the system engineering and program management necessary to sustain the safety, reliability, and security of the Los Alamos National Laboratory’s assets in the active United States nuclear stockpile - the B61, W76, W78 and W88. The Division generates key certification data for Annual Assessment supporting the Laboratory Director's letter to the President on the health of those warheads. This role demands ongoing surveillance of the active stockpile and evaluation of the potential impact of any issues through design, engineering, fabrication, testing using state-of-the-art computational simulation tools and engineering test facilities. The Division works in close liaison with the several production facilities across the nuclear security complex as well as with the customers in the US Navy and Air Force.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2021

Comprehensive Performance Calibration Guidance for Perovskites and Other Emerging Solar Cells

Emerging photovoltaic (PV) technologies (e.g., organic, perovskite, and solution processed quantum dot) have attracted remarkable attention with the rapid growth of their efficiencies, and their transition toward commercialization. Accurate and reliable efficiency measurements of these PV technologies are crucial, yet much more complicated than for conventional PV technologies due to the former's pronounced dynamic responses to changes in measurement conditions (e.g., current–voltage (I–V) scan rate and preconditioning) and their susceptibility to degradation. Adjustments to the measurement procedures are therefore necessary so that a reproducible “steady state” is reached during measurement. Furthermore, given the small size of many emerging cells, inappropriate device area definition and solar simulator setup can lead to measurement errors. Here, comprehensive efficiency calibration guidance is offered for emerging solar cells, including area measurement; spectral irradiance translation to standard test conditions; and steady-state electrical performance. The necessity of reporting steady-state efficiency is justified with a statistical performance comparison between conventional and steady-state I–V scans over hundreds of cells the group has received globally for efficiency certifications. The procedures described here do not require specialized measurement instrumentation; what matters most are changes to the measurement protocols. These described changes aim to enable better comparisons between reported efficiencies.

14 SOLAR ENERGY↗

DOE OSTI2017

FATIGUE BEHAVIOR OF SPENT NUCLEAR FUEL RODS IN SIMULATED TRANSPORTATION ENVIRONMENT

Nuclear fuel rod is composed of cladding tube and a specified number of fuel pellets contained. In the United States, spent nuclear fuel (SNF) is expected to be transported to at least one storage facility before permanent disposal. The fatigue behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending must be understood in order to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup SNFs (>45 GWd/MTU). This report presents the experimental results related to Zircaloy (Zry)-4 - based surrogate rods and high-burnup SNFs, based on recent work performed at Oak Ridge National Laboratory (ORNL). The surrogate rod was made of Zry-4 cladding and alumina pellets, and high-burnup fuel rods were discharged from H.B. Robinson pressurized water reactor. The reversed cyclic bending testing was conducted at 5 Hz under loading control. The effect of pre-hydriding and burnup or irradiation on the flexural rigidity and fatigue life of cladding-pellet system were discussed. The fatigue data obtained are extremely useful to the future certification of SNF storage and transportation cask.

Wang, Jy-An John↗

DOE OSTI2015

Model-based approaches to calibration and certification of nondestructive assay systems - 15615

Increasingly, model-based methods founded in physics are employed to calibrate, or extend the calibration of, nondestructive assay (NDA) systems used to characterize materials contaminated with plutonium, uranium, and other radioactive isotopes. In addition, modeling and mathematical methods are used to establish system parameters, correction factors, and measurement uncertainty components in NDA measurements. The methods combine the characteristics of the measurement configuration (detector structure, object shape and dimensions, source-detector distance and orientation, and collimation), generator knowledge of the waste composition, and known physical properties of the system in order to establish these parameters, factors, and components. These methods offer several important benefits over traditional methods used to test, calibrate, operate, and certify radiation measurement systems. First, instruments can be set up and put into service more quickly and correctly using a model-based approach. Second, calibrations can be performed, tested, and modified with reduced use of plutonium and uranium standards when modeling and mathematical methods are employed. Also, modeled approaches avoid calibration error resulting from heterogeneous source distribution. Moreover, the parameters can be installed quickly and simply in an office setting without the need for the safety, security, and safeguards considerations that are present when standards are used. Calibrations can also be tailored to fit unique or nonstandard container configurations (for example, over-packed drums), new container types, or specialized matrices. Finally, when modeling methods are applied, correction factors and measurement uncertainty elements can be developed, tested, and validated relatively quickly and without the need for surrogate matrix materials and a large number of test measurements. A combined gamma and neutron NDA system has recently been moved from Idaho National Laboratory to the TA 55 facility at Los Alamos National Laboratory (LANL). The system will be used to assay transuranic waste packaged in drums. After the move, the system was modified to replace degraded components and improve system performance. Afterwards, it was set up to assay different container types and to extend the calibration range. Model-based approaches were used to adapt the system to its new purpose at LANL. This report will describe the setup, calibration, and certification of the combined system at LANL Included in this discussion will be the measurement applications and details of the measurement methods. The steps in these processes that involved the use of modeling and mathematical calculations to replace or supplement more traditional methods will be identified and discussed. Finally, an overarching discussion of the methods and advantages of model-based methods to establish the properties of NDA systems will be presented. We conclude that validated mathematical methods offer a robust, flexible, and mature approach which is technically defensible from first principle physics. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Simple Acceptance Decisions and Acceptance Risk

Acceptance risk may be estimated using cumulative probability distribution functions applied to populations of measurands and test measurement values. “Simple acceptance” is a decision rule that sets the acceptance range of a test result equal to the tolerance range specification. While acceptance risk is comprised of both consumer risk and producer risk, this paper compares the effects of simple acceptance decision rules and guardbanding decision rules on consumer risk. Consumer risk is also known as the probability of false acceptance. The terms describe the risk of accepting test results as passing when the actual values exceed specification limits. False acceptance is only possible when the true value of a measurand is out of tolerance and the test result indicates that the measurand is within tolerance. Metrologists generally have some information about the test uncertainty regarding a specific acceptance test result. Along with a general lack of knowledge regarding the measurand population parameters, the complicated interplay between risk, dispersion, and central values generally prevents calibration laboratories from fully characterizing acceptance probabilities. Organizations that model the measurand populations can reduce consumer risk by avoiding certification of items whose measurand populations are not well centered. The models in this paper present a recurring trend: guardbanding reduces nonnegligible risks of false acceptance when compared to simple acceptance. However, guardbanding is not the most effective means of acceptance risk mitigation. Systematic characterization of measurand populations can provide the information a calibration laboratory needs to reliably control acceptance risk.

42 ENGINEERING↗

DOE OSTI2022

Aeroelastic Modeling for Distributed Wind Turbines: March 11, 2021 - November 10, 2021

Aeroelastic modeling (AM) is the primary methodology for structural and performance assessment of any wind turbine whereby providing an understanding of the impact of design parameters on its loading and power response before witnessing it in the field. Despite these advantages, the use of AM in the Distributed Wind Technology (DWT) sector is limited, especially within the less established manufacturers. This project represents an in-depth assessment of the status of AM and its role within the Standards for the DWT industry. The study gathered input and feedback from a large number of national and international stakeholders, reviewed technical strengths and weaknesses of the current edition of the design standards, analyzed recent industry workshops' and meetings' minutes, collected publicly available AM templates, and provided an evaluation of the existing AM codes. The study achieved several goals including providing strategies for the load assessment categorization of turbines based on rotor swept area and archetype, and guidance for AM verification and validation (V&V), which includes discussions of measurement requirements and a sample test-plan useful for future V&V campaigns and design standard development. This document summarizes the different tasks conducted in the course of the project and highlights the steps required to improve the AM adoption based on a multifaceted approach that encompasses: 1) augmenting AM software capabilities, 2) publishing AM best-practices and design-basis, 3) creating new model templates, 4) providing guidance for V&V of codes and specific turbine models leveraging field testing best-practice, and 5) addressing weaknesses in the current standards. Many of the future objectives identified in this study could leverage NREL's upcoming testing campaigns of three modern distributed wind turbines. Recommendations within this study will advance the value and the ease-of-use of AM, thereby allowing the industry to better capitalize this underutilized tool resulting in a more efficient design process, an easier path to certification, and overall better and more distributed reliable wind turbine products.

17 WIND ENERGY↗

DOE OSTI2016

MO-F-201-00: PANEL DISCUSSION: Preparing for Parts 2 and 3 of the ABR Therapy Medical Physics Exam

The goal of this session is to provide guidance to medical physicists undergoing the American Board of Radiology certification process in therapeutic medical physics. This panel discussion will focus on parts 2 (computer-based) and 3 (oral) of the examination. Unlike the latter portions of the exam which are specialty-driven, part 1 is universal for all medical physics fields and will not be addressed. This session is structured into different topics that aim to guide the participants on how to successfully prepare for the board exams. The subjects of discussion will include timing and strategies for exam preparation, crucial differences in preparing for the clinical computer-based exam versus the oral exam, what study tools are currently available for each, etc. The panel discussion format will allow the speakers to collectively present their experience and advice relating to each topic and foment audience participation. Learning Objectives: Know the main differences between what is expected for parts 2 and 3 Know different resources for test preparation Know how to formulate a plan to best study for each part based on the specific skill set the two parts require Know how to best present his/her answers during the oral examination – demeanor, answer structure, etc.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2022

Radiochronometric analysis of an historic Cs-137 activity standard

In this work, a 137 Cs activity standard discovered during routine inventory was analyzed to determine its model age using radiochronometry. The aqueous activity standard was separated using an established separation method that employs commercially available Sr resin. The method was also tested against a nuclear forensics reference material developed specifically for benchmarking 137 Cs radiochronometry methods. Results of the analyses showed good agreement between the results and the certified values of the reference material. Analysis of the discovered activity standard were also in good agreement with the activity certification date, though uncertainty was higher due to natural Ba contaminating the sample and the stable 133 Cs used as a carrier during the standard’s production.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

Report from the Inaugural Meeting on the Opportunities at the SEEMS Facility

An Inaugural Meeting on the Opportunities at the SEEMS Facility was held at the Spallation Neutron Source (SNS) on June 11 - 12, 2019. It focused on the second of the user communities - industrial and academic researchers. Indeed, the SEEMS design concept started with a Spallation Neutron Source (SNS) study conducted for the Federal Aviation Administration (FAA) that examines test facility options for providing neutron radiation simulating that in the atmosphere originating from cosmic rays and solar particles. The study was stimulated by earlier work calling for neutron radiation facilities with increased capacity and capabilities to meet current and future needs. Regulators, aircraft producers and avionics equipment suppliers recognize the growing urgent need for SEE testing in prototypical radiation fields. Industry standards formalizing procedures to assure that critical systems are robust against SEE phenomena have been issued. Regulatory bodies including the Federal Aviation Administration (FAA) and the European Union Aviation Safety Agency (EASA) now requiring that updated standards are met for aircraft certification.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Implementation and Demonstration of the Digital Twin Certification System Remote Operations Framework

Microreactors are one promising advanced-reactor concept being pursued by the nuclear industry. They are distinguished by a relatively low power output of 20 MWth or less. These microreactors are intended for deployment in applications where conventional small-capacity power solutions, such as diesel generators, are either economically unfeasible or logistically challenging. Such applications include providing electric power and/or heat for remote communities, mining sites, defense installations, and humanitarian and disaster-relief missions. An important feature for the successful deployment of microreactors is their capability to be operated remotely. This capability can significantly reduce staffing costs by eliminating the need for licensed operators to be physically present at each reactor site. Instead, operators can be centralized in a single remote operations center placed in an economically advantageous location, thereby optimizing resources by consolidating expertise and enhancing operational efficiency. However, the implementation of a remote operation system for nuclear reactors raises new concerns regarding the security, reliability, and resilience of such a system. One way in which remote operations can be supported in a manner that maintains system security, reliability, and resilience is through the use of digital twins in a novel framework designed to verify and validate sensor data and commands communicated between the remote operations center and reactor. This framework, known as the Digital Twin Certification System (DTCS), has previously been proposed as an operations architecture that can bring security and resiliency levels of remote nuclear-reactor operations to a level acceptable for commercial deployment. This paper moves the proposed DTCS architecture from concept to reality by presenting the implementation and testing of the system. The rationale and implementation of the DTCS using tools such as DeepLynx and Apache Airflow, is covered in-depth. This is followed by a demonstration of the DTCS by applying the implemented system architecture to the Single Primary Heat Extraction and Removal Emulator, a small-scale non-nuclear test bed that emulates thermal behavior of a microreactor. The demonstration includes both normal and abnormal operating scenarios to highlight how the DTCS can increase the security, reliability, and resilience of a remote operations system.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Implementation and Demonstration of the Digital Twin Certification System Remote Operations Framework

Microreactors are one promising advanced-reactor concept being pursued by the nuclear industry. They are distinguished by a relatively low power output of 20 MWth or less. These microreactors are intended for deployment in applications where conventional small-capacity power solutions, such as diesel generators, are either economically unfeasible or logistically challenging. Such applications include providing electric power and/or heat for remote communities, mining sites, defense installations, and humanitarian and disaster-relief missions. An important feature for the successful deployment of microreactors is their capability to be operated remotely. This capability can significantly reduce staffing costs by eliminating the need for licensed operators to be physically present at each reactor site. Instead, operators can be centralized in a single remote operations center placed in an economically advantageous location, thereby optimizing resources by consolidating expertise and enhancing operational efficiency. However, the implementation of a remote operation system for nuclear reactors raises new concerns regarding the security, reliability, and resilience of such a system. One way in which remote operations can be supported in a manner that maintains system security, reliability, and resilience is through the use of digital twins in a novel framework designed to verify and validate sensor data and commands communicated between the remote operations center and reactor. This framework, known as the Digital Twin Certification System (DTCS), has previously been proposed as an operations architecture that can bring security and resiliency levels of remote nuclear-reactor operations to a level acceptable for commercial deployment. This paper moves the proposed DTCS architecture from concept to reality by presenting the implementation and testing of the system. The rationale and implementation of the DTCS using tools such as DeepLynx and Apache Airflow, is covered in-depth. This is followed by a demonstration of the DTCS by applying the implemented system architecture to the Single Primary Heat Extraction and Removal Emulator, a small-scale non-nuclear test bed that emulates thermal behavior of a microreactor. The demonstration includes both normal and abnormal operating scenarios to highlight how the DTCS can increase the security, reliability, and resilience of a remote operations system.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

MO-F-201-01: Panel Member

The goal of this session is to provide guidance to medical physicists undergoing the American Board of Radiology certification process in therapeutic medical physics. This panel discussion will focus on parts 2 (computer-based) and 3 (oral) of the examination. Unlike the latter portions of the exam which are specialty-driven, part 1 is universal for all medical physics fields and will not be addressed. This session is structured into different topics that aim to guide the participants on how to successfully prepare for the board exams. The subjects of discussion will include timing and strategies for exam preparation, crucial differences in preparing for the clinical computer-based exam versus the oral exam, what study tools are currently available for each, etc. The panel discussion format will allow the speakers to collectively present their experience and advice relating to each topic and foment audience participation. Learning Objectives: Know the main differences between what is expected for parts 2 and 3 Know different resources for test preparation Know how to formulate a plan to best study for each part based on the specific skill set the two parts require Know how to best present his/her answers during the oral examination – demeanor, answer structure, etc.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

MO-F-201-02: Panel Member

The goal of this session is to provide guidance to medical physicists undergoing the American Board of Radiology certification process in therapeutic medical physics. This panel discussion will focus on parts 2 (computer-based) and 3 (oral) of the examination. Unlike the latter portions of the exam which are specialty-driven, part 1 is universal for all medical physics fields and will not be addressed. This session is structured into different topics that aim to guide the participants on how to successfully prepare for the board exams. The subjects of discussion will include timing and strategies for exam preparation, crucial differences in preparing for the clinical computer-based exam versus the oral exam, what study tools are currently available for each, etc. The panel discussion format will allow the speakers to collectively present their experience and advice relating to each topic and foment audience participation. Learning Objectives: Know the main differences between what is expected for parts 2 and 3 Know different resources for test preparation Know how to formulate a plan to best study for each part based on the specific skill set the two parts require Know how to best present his/her answers during the oral examination – demeanor, answer structure, etc.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

MO-F-201-03: Panel Member

The goal of this session is to provide guidance to medical physicists undergoing the American Board of Radiology certification process in therapeutic medical physics. This panel discussion will focus on parts 2 (computer-based) and 3 (oral) of the examination. Unlike the latter portions of the exam which are specialty-driven, part 1 is universal for all medical physics fields and will not be addressed. This session is structured into different topics that aim to guide the participants on how to successfully prepare for the board exams. The subjects of discussion will include timing and strategies for exam preparation, crucial differences in preparing for the clinical computer-based exam versus the oral exam, what study tools are currently available for each, etc. The panel discussion format will allow the speakers to collectively present their experience and advice relating to each topic and foment audience participation. Learning Objectives: Know the main differences between what is expected for parts 2 and 3 Know different resources for test preparation Know how to formulate a plan to best study for each part based on the specific skill set the two parts require Know how to best present his/her answers during the oral examination – demeanor, answer structure, etc.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2021

Mitigating Phase Unbalance for Distribution Systems with High Penetrations of Solar PV (Final Technical Report)

Distribution system operators have traditionally limited unbalance among phases by maintaining similar loadings on each phase. High penetrations of distributed solar PV continually change the net loading on each phase, resulting in time-varying phase unbalances that can damage three-phase devices such as three-phase motors, violate grid codes, and increase technical losses. This project has developed several control strategies for the reactive power outputs of solar PV inverters in order to mitigate power quality issues related to phase unbalance. These control strategies include a decentralized approach that is solely based on local measurements, distributed and grouped approaches that consider subsets of loads and PV generators, and a centralized approach that leverages measurements from a variety of locations in order to compute optimal reactive power setpoints for each inverter. Variants of the controllers handle challenges relevant to practical implementations, including noisy measurements, delayed communications, and reactive power limits. Moreover, the project developed theory that provides convergence guarantees for systems with multiple interacting controllers as well as “balanceability” certificates that ensure satisfaction of phase unbalance requirements with variable loading. The controllers were integrated with NRECA’s Open Modeling Framework (omf.coop) and evaluated using actual distribution system models obtained from several NRECA member utilities. Application of the controllers results in significant improvements to phase unbalance in these test cases with decreases from base case levels of over 3% to under 0.5%, which is within the 2% IEC phase unbalance standard.

14 SOLAR ENERGY↗

DOE OSTI2016

Advanced Simulation and Computing FY17 Implementation Plan, Version 0

The Stockpile Stewardship Program (SSP) is an integrated technical program for maintaining the safety, surety, and reliability of the U.S. nuclear stockpile. The SSP uses nuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. It includes stockpile surveillance, experimental research, development and engineering programs, and an appropriately scaled production capability to support stockpile requirements. This integrated national program requires the continued use of experimental facilities and programs, and the computational capabilities to support these programs. The Advanced Simulation and Computing Program (ASC) is a cornerstone of the SSP, providing simulation capabilities and computational resources that support annual stockpile assessment and certification, study advanced nuclear weapons design and manufacturing processes, analyze accident scenarios and weapons aging, and provide the tools to enable stockpile Life Extension Programs (LEPs) and the resolution of Significant Finding Investigations (SFIs). This requires a balance of resource, including technical staff, hardware, simulation software, and computer science solutions. ASC is now focused on increasing predictive capabilities in a three-dimensional (3D) simulation environment while maintaining support to the SSP. The program continues to improve its unique tools for solving progressively more difficult stockpile problems (sufficient resolution, dimensionality, and scientific details), and quantifying critical margins and uncertainties. Resolving each issue requires increasingly difficult analyses because the aging process has progressively moved the stockpile further away from the original test base. Where possible, the program also enables the use of high performance computing (HPC) and simulation tools to address broader national security needs, such as foreign nuclear weapon assessments and counter nuclear terrorism.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Design and Development of the R79: A new Type B and IP-1-2-3 Transportation Cask for Historic Waste - 18570

In October 2014, Robatel began the design and development of a new Type B and IP-1-2-3 transportation cask, the R79. This cask is designed to transport 'historical' waste from nuclear power plants, nuclear research labs, and nuclear medical isotope production facilities. This 'historical' waste may consist of activated and/or contaminated components including irradiated metal from decommissioning activities, fuel assembly containers, consumables, dried residues, etc. It may also include dewatered ion exchange resins, dried concentrates (salt, metal oxide, sludge, etc.), and hydrogenous materials. The main isotopes in this waste are Co-60 and Cs-137. The cask is not designed to transport fissile material but a trace quantity of up to 22.5 g is allowed. For ease of loading and unloading, the cask is designed to handle waste prepackaged into standard drums. With a loading capacity of 300 kg (661 lb) and a volume of 350 L (12.3 ft{sup 3}), the R79 cask is well suited to carry small to medium-sized waste shipments. The cask body is constructed of 304L stainless steel, its impact limiters are filled with Robatel's FENOSOL{sup TM} foam, and the shielding material includes Robatel's PNT7{sup TM} compound for its thermal and neutron absorbing properties. The cask is loaded in a dry environment and is designed for road transportation. Some novel features were integrated into the design of this cask. With its impact limiters installed, the cask may be treated as a Type B container. Depending on the result of the gas generation analysis, the cask can be treated as either Type B(U) (no limit of the transport duration) or Type B(M) (transport duration limited to 2 weeks). With the impact limiters removed, the R79 cask functions as an industrial-type package (either IP-1, IP-2, or IP-3 depending on the content classification). This can be advantageous for a number of reasons. First, it takes less time for loading and unloading operations when the impact limiters are not installed. Secondly and more importantly, three industrial-type packages can be transported on the same conveyance compared to two Type B packages. These options allow the shipper more flexibility: if the content meets the lower radiation limits for an industrial type package, more waste can be transported on one shipment. If the content must be classified as Type B, the user only must install the impact limiters prior to shipment. The design and development of this cask was performed in 2015, with the first drop testing beginning in July 2015. The complete design documents were first submitted September 2015. Robatel completed the foam for the impact limiters in October 2016, with the completed impact limiters finished in March 2017. The end of manufacturing, delivery of the cask to our customer, and receipt of the regulatory certificate of approval is scheduled before the end of 2017. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Cask RT-100 Type B(U) Self-Certification for DOT 7A, Type A Nuclear Waste Transportation - 20098

The purpose of this Type A certification is to diversify the use of the RT-100 cask by allowing the end user to transport either Type A or Type B radioactive waste in accordance with the handling procedures for each waste category. The project's scope included the development of a matrix that contains all applicable requirements for a Type A package from the Code of Federal Regulations (CFR), Title 49 for the Department of Transportation (DOT) with a justification for each requirement. Moreover, the scope included the writing of conditions and restrictions list for the transportation of Type A nuclear waste, the writing of a compliance statement, and the design and fabrication of a Type A placard to be permanently placed on the RT-100 cask. Type A packages are self-certified by the end user using one of three methods as permitted by 49 CFR 173.461, either by performance of tests with prototypes or samples of the specimens representing the packaging, or reference to a previous satisfactory demonstration of compliance of a sufficiently similar nature, or calculations or reasoned evaluation using reliable and conservative procedures and parameters. Since the RT-100 cask was licensed by the Nuclear Regulatory Commission (NRC) as a Type B cask, Robatel relied on the method of reference to a previous satisfactory demonstration of compliance of a sufficiently similar nature for Type A certification. This method was determined to be acceptable, since Type B packages are evaluated for more demanding transport conditions than those required for Type A packages. This paper presents Robatel's process for certifying the RT-100 cask for Type A shipments, an overview of the DOT regulations related to Type A packages, and the documents that were generated for the end user. This paper is concluded with a discussion on the advantages of having the option to transport both Type A and Type B nuclear waste, and possible future projects to further improve the operation of the cask when making Type A shipments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗