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DOE OSTI2019

SAFETY ANALYSIS REPORT FOR PACKAGING SHIELDING & NUCLEAR CRITICALITY SAFETY COURSES DEVELOPED AND CONDUCTED BY OAK RIDGE NATIONAL LABORATORY

The US Department of Energy (DOE) Packaging Certification Program, Office of Packaging and Transportation, is offering Safety Analysis Report for Packaging (SARP) Shielding and Nuclear Criticality Safety (NCS) courses for SARP generalists and analysts. The SARP generalist course is designed for project managers, supervisors, NCS/shielding subject matter experts (SME), and SMEs in non-NCS/shielding technical areas (e.g., structural, thermal, package design) who need to improve their understanding of how NCS/shielding analyses fit into the broader spectrum of SARP documentation. The generalist course provides an overview of the regulations and guidelines for the criticality and shielding analyses included in a SARP, and it illustrates how the NCS/shielding chapters integrate with the other parts of the SARP. After course material is presented, students in the generalist course will review an actual SARP document to examine key elements of the shielding and criticality analyses. The analyst course provides detailed instructions on the radioactive material package shielding analyses and NCS evaluation fundamentals needed by analysts/practitioners, including safety analysts and/or technical reviewers, to prepare and/or review technical analyses for SARP documentation. The analyst course also provides an overview of regulations and guidelines, in addition to detailed in-class exercises associated with package shielding and NCS analyses. Analysis teams will be faced with staged SARP examples in which several important decision processes in the generation of a SARP will be demonstrated and discussed. Both courses are offered at Oak Ridge National Laboratory, and for interested students, the analyst course is offered for one hour of graduate credit (course ME-690) from the University of Nevada at Reno. An overview of the course will be provided to ensure that the NCS community is aware of the course and its intent to support the shielding and NCS aspects of packaging and transportation analyses for SARP activities.

Bowen, Douglas G.↗

DOE OSTI2016

Simulating Regional Groundwater Flow in the Vicinity of the Nevada National Security Site, Nye County, Nevada - 16104

The Death Valley Regional Groundwater Flow System (DVRFS) occupies an area of about 70,000 km{sup 2} in south-central Nevada and southeastern California. Interest in this system stems from the need to (1) understand the groundwater flow paths and travel times associated with potential migration of radioactive material from the Nevada National Security Site (NNSS) where underground nuclear testing was conducted from 1956 to 1992; (2) characterize the groundwater system in the vicinity of the proposed high-level radioactive waste repository at Yucca Mountain, Nev.; and (3) address a variety of potential effects from NNSS activities on users down gradient from the NNSS and Yucca Mountain. A single groundwater flow model of the DVRFS has been developed that integrates data and results of two previous regional-scale models and supports Department of Energy (DOE) programs at the NNSS. The numerical model was constrained by organizing hydrogeologic data and interpretations in three-dimensional (3D) geographic information systems and constructing a digital three-dimensional (3D) hydrogeologic framework model (HFM) to represent 27 hydrogeologic units (HGUs) and major structures in the DVRFS region. The 3D hydrogeologic data sets were discretized to 1,500 meters (m) grid cell resolution for input arrays required for the model. The DVRFS was simulated using the USGS 3D groundwater-flow modeling code MODFLOW-2000. Two examples showing how the DVRFS regional model is used by the DOE as part of waste management activities at the NNSS: (1) use of the regional model to provide the boundary conditions for site-scale models, and (2) use of the regional model to assess changes in hydraulic gradient and flow paths in the regional flow system resulting from changes in pumping in various areas of interest. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Selection of Vadose Zone Flow and Transport Properties with Gravel Fraction Corrections for the Hanford Site Composite Analysis and Cumulative Impact Evaluation

This environmental calculation file (ECF) is a compilation of hydrologic and radiological/chemical properties to be used for the updated Hanford Site Composite Analysis (CA) and Cumulative Impact Evaluation (CIE) vadose zone (VZ) modeling. The CA will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1 Chg 1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. The CIE evaluates the effects of cleanup decisions regarding groundwater quality in the Hanford Site Central Plateau (DOE/RL-2018-69, Cumulative Impact Evaluation Technical Approach Document). Due to the complexity and large number of waste sites in source operable units (OUs), the computational tools used for the CIE must be capable of representing a range of site conditions and source terms in the VZ while also efficiently computing the impact that cleanup decisions have on the underlying aquifer. Waste-site proximity between and within source OUs has resulted in contaminants commingling in the vadose and saturated zones in complex ways. Plume commingling requires cleanup decisions to be evaluated considering the surrounding waste sites and existing groundwater contamination, therefore demonstrating the need to evaluate cumulative impacts from the VZ to groundwater.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2011

Age Dating of Mixed SNM--Preliminary Investigations

Recently we investigated the nuclear forensics problem of age determination for mixed special nuclear material (SNM). Through limited computational mixing experiments and interactive age analysis, it was observed that age dating results are generally affected by the mixing of samples with different assays or even by small radioactive material contamination. The mixing and contamination can be detected through interactive age analysis, a function provided by the Decay Interaction, Visualization and Analysis (DIVA) software developed by NSTec. It is observed that for mixed SNM with two components, the age estimators typically fall into two distinct clusters on the time axis. This suggests that averaging or other simple statistical methods may not always be suitable for age dating SNM mixtures. Instead, an interactive age analysis would be more suitable for age determination of material components of such SNM mixtures. This work was supported by the National Center for Nuclear Security (NCNS).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2014

International Training Course on Implementing State Systems of Accounting for and Control (SSAC) of Nuclear Material for States with Small Quantity Protocols (SQP): Course Overview and Objectives [Slides]

This course is hosted by Los Alamos National Laboratory on behalf of the U.S. National Nuclear Security Administration (NNSA) in collaboration with the IAEA to support Member States in managing small quantities of nuclear and radioactive materials in a manner that is compliant with their international obligations.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2019

Panel Session 57: Nuclear Power Plant Waste Management - LLW Processor Issues

This panel resulted in senior utility managers discussing issues with commercial LLW processors. LLW processors updated utilities/NPPs on new or different services or technologies being offered to assist in their radioactive waste management goals, changes to their facility license, revisions to their acceptance guidelines, and emerging compliance issues. Presentations proposed include: 1 - EnergySolutions Waste Processing Overview (Joe Heckman, EnergySolutions (USA)); 2 - Perma-Fix: A Nuclear Services and Waste Management Company (John Kelly, Perma-Fix Environmental Services (USA)); 3 - UniTech - 'Expanding to Reduce Costs' LLW Processor Update (Gregg Johnstone, UniTech Services Group, Inc. (USA)); 4 - Low - Level Radioactive Materials Processing (Michael Miller, Veolia Nuclear Solutions (USA))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

High temperature mechanical properties of fluorite crystal structured materials (CeO 2 , ThO 2 , and UO 2 ) and advanced accident tolerant fuels (U 3 Si 2 , UN, and UB 2 )

The mechanical interaction between the fuel and cladding that occurs during operation of a nuclear reactor is important to understand as it can lead to cladding failures and release of radioactive material into the coolant. Additionally, in order to develop better models of the pellet-clad mechanical interactions, the mechanical properties of the fuel at relevant operating temperatures, like the elastic moduli, are needed for current and advanced accident tolerant fuels (ATFs). In this work, elevated temperature nanoindentation and resonant ultrasound spectroscopy were used to measure the moduli and hardness of several fluorite materials (CeO 2 , ThO 2 , UO 2 ) and several ATF candidates (ATF) (U 3 Si 2 , UN, UB 2 ). In addition, a comparison of the two techniques was performed in this study to independently validate the mechanical properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

MPF X-Ray Module Setup for 2018

The Mobile Plutonium Facility (MPF) is a portable module design facility that can be deployed anywhere in the world for the safe recovery, packaging, and shipment of a nations legacy, or excess plutonium stockpile. All of the modules of MPF have specific purposes for operating in a mission if one were to ever come up. For my past four internships here at SRNL I have been working on the support of the X-Ray module for MPF. This year I installed new equipment I had purchased last year for the X-Ray Module and made sure the module was mission ready for the chance that the team deploys in the future for plutonium recovery. Using digital radiography (DR) and computed tomography (CT) provides the MPF team the ability to verify the contents of containers that would be holding material. This also ensures the safety of the team members in the glovebox facilities by verifying that there are no sharps, liquids, or other items that may harm the workers or glovebox operations when working with radioactive material. X-Ray imaging can also pick up low density objects like plastic bags and tape. Note how the DR image hides the puncture in the inner can while the CT image shows the damage. New DR and CT computers with upgraded software, New imaging panel, NOVO Portable X-Ray System. Because of the new upgrades to the X-Ray module system containers can be scanned and evaluated quicker than before. With the last system the 3D model reconstruction alone would take anywhere from 10-12 minutes, now with the new system that time has been brought down to just shy of 1 minute 30 seconds. The new portable system also can 1,2 or 4 images of an object and stitch them together into a mosaic for scanning of larger containers.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

CONSTRAINTS ON THE ORIGIN OF THE FIRST LIGHT FROM SN 2014J

We study the very early light curve of supernova 2014J (SN 2014J) using the high-cadence broad-band imaging data obtained by the Kilodegree Extremely Little Telescope, which fortuitously observed M 82 around the time of the explosion, starting more than 2 months prior to detection, with up to 20 observations per night. These observations are complemented by observations in two narrow-band filters used in an Hα survey of nearby galaxies by the intermediate Palomar Transient Factory that also captured the first days of the brightening of the supernova. The evolution of the light curves is consistent with the expected signal from the cooling of shock heated material of large scale dimensions, ≳1R {sub ☉}. This could be due to heated material of the progenitor, a companion star or pre-existing circumstellar environment, e.g., in the form of an accretion disk. Structure seen in the light curves during the first days after explosion could also originate from radioactive material in the outer parts of an exploding white dwarf, as suggested from the early detection of gamma-rays. The model degeneracy translates into a systematic uncertainty of ±0.3 days on the estimate of the first light from SN 2014J.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2018

Vitrification of Hanford Tank Waste 241-AP-105 in a Continuous Laboratory-Scale Melter

Low-activity waste (LAW) stored in underground tanks on the Hanford site in Washington State is planned to be filtered for solids removal and processed through ion exchange columns for cesium removal at the Low-Activity Waste Pretreatment System facility. These pretreatment steps will allow the waste to be transferred to the Hanford Tank Waste Treatment and Immobilization Plant LAW facility for immobilization into glass. The liquid waste will be combined with glass-forming chemicals (GFCs) to form a waste feed slurry that can be fed to electric melters for vitrification. The process of continuously converting the aqueous feed slurry into a melt is dynamic and includes multiple reactions, degassing, and dissolution processes that depend on heat from the melt below. In this conversion process, waste components are partitioned into one of two streams: glass and off-gas. Processing information and chemical information associated with these waste product streams has been requested by Washington River Protection Solutions for actual tank waste from tank 241-AP-105 (referred to as AP-105). To acquire this type of information, a small-scale melter system was desired that would not require high volumes of input waste or the large resource commitment of a full-scale melter system, while also providing dynamic information that would be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM), has been designed to operate with a continuous feeding process, while periodically pouring glass product and collecting off-gas. The CLSM vessel has been sized to collect the relevant process and chemical information from obtainable volumes of AP-105 waste samples. Two CLSMs were constructed: one in a non-radioactive environment for processing waste simulants and another in a fume hood capable of handling radioactive material for processing actual Hanford tank waste. This dual-system setup allowed for comparison between simulated and real tank waste with essentially identical test equipment. AP-105 waste simulant was first tested in the non-radioactive CLSM to determine the processability and desired operating window for the actual tank waste. Rhenium (Re) was added to the simulant to act as a surrogate for technetium-99 ( 99 Tc), a key waste component found in the actual tank waste. The partitioning of Re into the glass and off-gas products during the continuous feeding of simulant was determined, which allowed the Re retention, the mass flow rate of Re out of the CLSM via the glass product with respect to the mass flow rate of Re into the CLSM via the feed slurry, to be calculated. Key processing results and the average Re retention, from the CLSM test run with AP-105 simulant feed slurry are shown in Table ES.1.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Management of Documents and Information on Radioactive Waste - 19455

The assurance of the safety of a repository for low- and intermediate-level radioactive waste, during the 300 years of institutional control after closure, depends on assuming that the responsible institution will be capable to control the documents with information regarding the facility, the disposed of waste and the established measures for counteracting events that could breach the isolation of the radioactive material. Management of documents or information does not represent a major challenge in itself in view of consolidated practices and international standards, such as ISO 15489. However, considering the time scale involved in the disposal of the radioactive wastes, efforts should be made to ensure the preservation and proper access to documents and information. In this matter, the technological aspect of document production can be a complicating factor. In fact, it is common to be unable to access electronic documents produced 20 or 30 years ago, due either to problems related to the media where the information is stored, or to a lack of compatible systems for reading the files. This is just one example of complication when thinking about production, organization, representation, preservation and retrieval of recorded information that is to last for centuries. In this way, this work seeks to present the situation of the management of documents and information on radioactive waste in the scope of the Institute of Energy and Nuclear Research (IPEN) and with national regulation in Brazil. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Recommendations for Secure Transport: Material Conveyance Physical Protection Technology

Nuclear material is at higher risk of theft and sabotage during transport than during any other phase of the nuclear fuel lifecycle. Nuclear materials are transported in the public domain where adversaries have an upper hand by taking advantage of the time and location of the theft or sabotage attempt. As such, even modest threat profiles for transport of nuclear and radioactive material can require substantial detection and delay measures to support timely response. Conveyance tracking augmented with technology that improves on-the-scene situational awareness at a remote monitoring center has been adopted as a de-facto standard approach for transportation security. At present, the extended tracking and situational awareness capabilities needed for a nuclear material shipment, as is provided by the purpose designed, Transportation – Security, Tracking and Reporting (T-STAR) System, do not exist in a single commercial off-the-shelf (COTS) solution. Typically, the COTS systems that excel in one area are deficient in other areas, presenting challenges to designing well-rounded, robust systems. Still, COTS solutions can offer the basic set of tracking and situational awareness capabilities by indicating last update time, current location, and route taken. By incorporating vehicle and driver performance measures via the vehicle’s controller area network (CAN bus) and video alongside other data streams allows telemetry and other shipment information to be assessed in novel ways.This paper describes the operation, capabilities and features of various conveyance protection systems and approaches, assesses emerging technologies and how they could be used through a unified interface, and enumerates additional ways to provide early detection using vehicle, onboard technologies, effective delay technologies and approaches and simple equipment to improve protection during transport.

Shannon, Michael↗

DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

This study focuses on requirements and challenges pertaining to advanced transportable nuclear energy systems, which are currently in development and feature varying degrees of mobility. These reactors mostly use high-assay low-enriched uranium (HALEU) fuel, and only limited refueling is required during their lifecycles. Due to the smaller source terms, higher burnup fuel, and potentially higher thermal efficiency involved, the waste inventory will be less than that generated by conventional light-water reactors. However, the relevant regulatory requirements must be met when transporting HALEU fuel (whether fresh or used) inside the reactor vessel—mainly the fuel material transportation, siting, and physical protection requirements outlined in US Nuclear Regulatory Commission (NRC) regulations 10 CFR 71 and 73. Spent nuclear fuel (SNF), as well as irradiated fuel (used during any interval of reactor operation), contains highly radioactive materials that generate heat and necessitate special precautions during transportation. This study addresses the challenges that this incurs, and outlines potential solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

A Novel and Cost Effective Approach to the Decommissioning and Decontamination of Legacy Glove Boxes - Minimizing TRU Waste and Maximizing LLW Waste - 13634

This paper describes the process of decommissioning two gloveboxes at the Argonne National Laboratory (ANL) that were employed for work with plutonium and other radioactive materials. The decommissioning process involved an initial phase of clearing tools and materials from the glove boxes and disconnecting them from the laboratory infrastructure. The removed materials, assessed as Transuranic (TRU) waste, were packaged into 55 gallon (200 litre) drums and prepared for ultimate disposal at the Waste Isolation Pilot Plant (WIPP) at Carlsbad New Mexico. The boxes were then sampled to determine the radioactive contents by means of smears that were counted with alpha and beta detectors to determine the residual surface contamination, especially in terms of alpha particle emitters that are an indicator of TRU activity. Paint chip samples were also collected and sent for laboratory analysis in order to ascertain the radioactive contamination contributing to the TRU activity as a fixed contamination. The investigations predicted that it may be feasible to reduce the residual surface contamination and render the glovebox structure low level waste (LLW) for disposal. In order to reduce the TRU activity a comprehensive decontamination process was initiated using chemical compounds that are particularly effective for lifting and dissolving radionuclides that adhere to the inner surfaces of the gloveboxes. The result of the decontamination process was a reduction in the TRU surface activity on the inner surfaces of the gloveboxes by four orders of magnitude in terms of disintegrations per unit area (DPA). The next phase of the process involved a comprehensive assay of the gloveboxes using a combination of passive neutron and gamma ray scintillation detectors and a shielded and collimated high purity Germanium (HPGe) gamma ray detector. The HPGe detector was used to obtain gamma ray spectra for a variety of measurement positions within the glovebox. The spectra were used to determine the TRU content of the boxes by assessing the activity of Am-241 (59 keV) and Pu-241 (414 keV). Using the data generated it was possible for qualified subject matter experts (SME) to assess that the gloveboxes could be consigned for disposition as LLW and not as TRU. Once this determination was assessed and accepted the gloveboxes were prepared for final disposition to the Nevada National Security Site (NNSS) - formerly the Nevada Test Site (NTS). This preparation involved fixing any remaining radioactive contamination within the gloveboxes by filling them with a foam compound, prior to transportation. Once the remaining contamination was fixed the gloveboxes were removed from the laboratory and prepared for transported by road to NNSS. This successful glovebox decontamination and decommissioning process illustrates the means by which TRU waste generation has been minimized, LLW generation has been maximized, and risk has been effectively managed. The process minimizes the volume of TRU waste and reduced the decommissioning time with significant cost savings as the result. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Data Quality Objectives Supporting Radiological Air Emissions Monitoring for the PNNL Site

Pacific Northwest National Laboratory (PNNL) is in the process of developing a radiological air monitoring program for the PNNL Site that is distinct from that of the nearby Hanford Site. The original DQO (PNNL-19427) considered radiological emissions at the PNNL Site from Physical Sciences Facility (PSF) major emissions units. This first revision considers PNNL Site changes subsequent to the implementation of the original DQO. A team was established to determine how the PNNL Site changes would continue to meet federal regulations and address guidelines developed to monitor air emissions and estimate offsite impacts of radioactive material operations.

Compliance↗

DOE OSTI2025

Microscopic Measurement: Need for Nuclear Material Sample Specimen Characterization Capability

The Surveillance, Fabrication, and Off-Site Operations team at Lawrence Livermore National Laboratory is beginning an update to the nuclear material machining lab in Superblock. Plutonium and other nuclear materials tend to accelerate the aging process for equipment and machines that would otherwise be maintainable for long lives. They have their lifespans shortened from deteriorated seals, embrittled polymer components, accelerated corrosion, and damage to electronics from exposure to radioactive materials. The cost of installation of equipment into a glove box (GB) tends to outstrip the cost of the equipment itself. Due to the magnification of cost, the historical approach to equipment installs has been to purchase and install the highest-quality and most robust equipment available to maximize time between failures and updates. To make room for the installation of a new lathe, the inspection GB is being moved to a different location within the machining lab. As part of the overall effort, the equipment within the inspection GB is being revisited. Current measurement capabilities are antiquated and starting to become problematic due to age. Advances in metrology are sought for more comprehensive characterization of scientific samples. This document aims to identify several options for the replacement of current measurement equipment and identify a system for purchase and eventual installation. The intention is not to select the cheapest system, but rather to identify a system that best meets the needs of both the operators and principal investigators (PI). Following the convention of the SYSM-5620 Design Thinking and System Engineering, it was determined that a laser confocal measurement system from Keyence is the most suitable option for our programmatic need. Form factor, capability, and maintainability were considered. Along with stakeholder input, to ensure the best decision for the Nuclear Material Technology Program.

36 MATERIALS SCIENCE↗