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DOE OSTI2017

Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy specimens of Midland beltline weld have been machined and just arrived to ORNL as part of this international collaboration. The ORNL will initiate tests of the irradiated Linde 80 weld in FY2017 and results of this international program will be reported in FY2018.

42 ENGINEERING↗

DOE OSTI2016

The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Gamma-Ray Raster Imaging with Robotic Data Collection (FY19 Annual Report)

This project has developed an alternative imaging capability for imaging and quantifying radioactive materials in situ. This approach uses a robotic-mounted gamma-ray detector which can move around an area of interest, sampling the space at a high frequency. By rastering across the area, an image can be created with no collimation and a high efficiency. An open-source robotic platform was programed to search an open area, collecting gamma-ray spectra every second. Informative path planning was used to direct the robot to areas of higher dose. Finally, a Bayesian particle filter was used to locate the source with minimal counting statistics. The use of a robotic mount allows data collection for long periods of time unattended, and it will also eliminate uncertainties in positioning typically introduced by personnel. This approach will be particularly relevant for gloveboxes, shielded cells, or process piping which may have complex, non-uniform distributions of material.

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DOE OSTI2016

USE OF SMALL SPECIMENS FOR FRACTURE TOUGHNESS EVALUATION OF RPV STEELS

Sokolov, Mikhail A↗

DOE OSTI2018

Determination of Static and Dynamic Crush Strengths of Stainless Steel Honeycomb Material

Honeycomb structures are ideal energy absorbers due to the high axial strength compared to the low density of the material and consistent mechanical properties. Lighter density materials used in honeycomb structures, like aluminum, have been studied intensively because of their use in aerospace applications. However, relatively denser materials like stainless steel have been less studied. The purpose of this project is to determine if 304 Stainless Steel hexagonal honeycomb can be used in applications related to the field of radioactive material packaging by determining the crush strength of the structure.

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DOE OSTI2018

Contamination control monitoring at the Los Alamos National Laboratory’s Plutonium Facility

Plutonium airborne contamination creates a particular hazard because of the possibility of intake by inhalation. To preclude excursions of radioactive materials into the operator’s breathing zone, controls are used to prevent contamination. Measures to control radioactive contamination include eliminating or minimizing the spread of contamination and preventing cross-contamination. Contamination control philosophy at the plutonium facility is conveyed in one simple statement: Detect the contamination closest to the source as possible! Speed and distance when self-monitoring is critical to detecting plutonium contamination. Responding correctly to an alarm or monitor not working properly minimizes the consequences of plutonium contamination. Performing a full body frisking correctly and using the Peerchecking tool also minimize the consequences of plutonium contamination.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2018

Advanced Characterization of Irradiated Fuels and Materials at INL

Advanced characterization of the irradiated fuels and materials is crucial to help understanding the material and fuel property changes under irradiation. Microstructural defect evolution down to nanometer scale could strongly affect the material and fuel performance on macroscale. Radiation-induced material degradation include dimensional instability, hardening and embrittlement, intergranular cracking, radiation-enhanced chemical interaction at the interfaces and breakaway swelling. To mitigate the radiation damage, advanced characterization on irradiated microstructure, microchemistry, micromechanical property and local thermal property is necessary to identify the controlling mechanisms that are responsible to the degradations. This presentation provides a summary of the advanced characterization capabilities at the Irradiation Materials Characterization Laboratory (IMCL), Materials and Fuels Complex (MFC), Idaho National Laboratory (INL). It highlights the recently established capabilities including shield sample preparation area (SSPA), shield confinement cells for two focused ion beam (FIB) systems, a shielded electron probe microanalysis (EPMA), the thermal property microscopy (TPM) along with non-shielded instruments like scanning electron microscopy (SEM), X-ray diffractometer, and transmission electron microscopy (TEM). Some examples on advanced characterizations on irradiated structural materials and fuels will be presented. It demonstrates the power of combining sample preparation of highly radioactive materials with the advanced characterization down to atomic resolutions within one facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Radioactive Waste Management and Nuclear Facility Decommissioning Progress in Iraq - 13216

Management of Iraq's radioactive wastes and decommissioning of Iraq's former nuclear facilities are the responsibility of Iraq's Ministry of Science and Technology (MoST). The majority of Iraq's former nuclear facilities are in the Al-Tuwaitha Nuclear Research Center located a few kilometers from the edge of Baghdad. These facilities include bombed and partially destroyed research reactors, a fuel fabrication facility and radioisotope production facilities. Within these facilities are large numbers of silos, approximately 30 process or waste storage tanks and thousands of drums of uncharacterised radioactive waste. There are also former nuclear facilities/sites that are outside of Al-Tuwaitha and these include the former uranium processing and waste storage facility at Jesira, the dump site near Adaya, the former centrifuge facility at Rashdiya and the former enrichment plant at Tarmiya. In 2005, Iraq lacked the infrastructure needed to decommission its nuclear facilities and manage its radioactive wastes. The lack of infrastructure included: (1) the lack of an organization responsible for decommissioning and radioactive waste management, (2) the lack of a storage facility for radioactive wastes, (3) the lack of professionals with experience in decommissioning and modern waste management practices, (4) the lack of laws and regulations governing decommissioning or radioactive waste management, (5) ongoing security concerns, and (6) limited availability of electricity and internet. Since its creation eight years ago, the MoST has worked with the international community and developed an organizational structure, trained staff, and made great progress in managing radioactive wastes and decommissioning Iraq's former nuclear facilities. This progress has been made, despite the very difficult implementing conditions in Iraq. Within MoST, the Radioactive Waste Treatment and Management Directorate (RWTMD) is responsible for waste management and the Iraqi Decommissioning Directorate (IDD) is responsible for decommissioning activities. The IDD and the RWTMD work together on decommissioning projects. The IDD has developed plans and has completed decommissioning of the GeoPilot Facility in Baghdad and the Active Metallurgical Testing Laboratory (LAMA) in Al-Tuwaitha. Given this experience, the IDD has initiated work on more dangerous facilities. Plans are being developed to characterize, decontaminate and decommission the Tamuz II Research Reactor. The Tammuz Reactor was destroyed by an Israeli air-strike in 1981 and the Tammuz II Reactor was destroyed during the First Gulf War in 1991. In addition to being responsible for managing the decommissioning wastes, the RWTMD is responsible for more than 950 disused sealed radioactive sources, contaminated debris from the first Gulf War and (approximately 900 tons) of naturally-occurring radioactive materials wastes from oil production in Iraq. The RWTMD has trained staff, rehabilitated the Building 39 Radioactive Waste Storage building, rehabilitated portions of the French-built Radioactive Waste Treatment Station, organized and secured thousands of drums of radioactive waste organized and secured the stores of disused sealed radioactive sources. Currently, the IDD and the RWTMD are finalizing plans for the decommissioning of the Tammuz II Research Reactor. (authors)

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DOE OSTI2018

Review of Exemptions and General Licenses for Fissile Material in 10 CFR 71

Title 10 of the United States (US) Code of Federal Regulations (CFR), Part 71, Packaging and Transportation of Radioactive Material, includes requirements for the transport of fissile material in packages. These regulations provide requirements for packages that are certified by the U.S. Nuclear Regulatory Commission (NRC) to transport fissile material. 10 CFR 71 also includes provisions for exemptions from classification as fissile material (10 CFR 71.15) and permits general licenses for limited quantities of fissile material and plutonium-beryllium special form sources (10 CFR 71.22 and 10 CFR 71.23, respectively). These provisions were extensively revised by the NRC as part of a broad rule-making effort in 2004. The exemption and general license provisions are used frequently for a wide-variety of fissile material shipping operations, and consignors must interpret the regulations for particular shipping situations. The purpose of this document is to provide explanatory information on the background, intent, and anticipated use of the provisions in an effort to assist fissile material licensees in their interpretation and application of the provisions such that criticality safety is ensured during transportation activities. A variety of examples are presented and discussed to illustrate how the regulations should be applied.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2018

sUAS Support for Mars 2020

The release of airborne radioactive material presents a health risk hazard to many individuals, including emergency responders and the public. Characterizing the unknown radioactive dangers produced during such incidents is essential. An advantage to utilizing unmanned aerial systems in this effort is the reduction of risk to personnel otherwise tasked with characterization, as well as more quickly obtaining data points during cloud immersion. By pairing a continuous air monitor (CAM) and detector with a drone, quantification of unknown material in the plume can be determined, and thus the extent of dose to people in the area can be estimated. A small, low-flow air sampler with a Geiger-Mueller counter may be characterized using experimental measurements of point sources representing a cloud of material and modelled using the Monte Carlo N-Particle Transport Code (MCNP), a software package for simulating nuclear processes, to find the usefulness and advantages of the system for detection of nuclides of concern in various nuclear plume scenarios. The flying time of small-unmanned aircraft systems (sUASs) is limited; thus, the minimum detectable activity and concentration rates possible must be determined over the system’s window of operation. While the activity concentrations corresponding to external and internal doses of concern can be detected with certainty for 90 Sr and 137 Cs in a plume, those corresponding to 238 Pu are more difficult to detect. The central principle of employing the sUAS is that the cockpit does not have an operator onboard the aircraft; therefore, control of the aircraft must take place by other means. With a sUAS, command and control can be separated into three distinct methods: (1) ground control or remote piloting, (2) semi-autonomous control, and (3) autonomous control. The isotope 238 Pu is used in the space industry, providing long-lived heat generation for fuel in outer space (NASA 2014). Due to the inherent risks with rocket launches to get objects into space, monitoring for release of these hazardous substances is a major undertaking. The current means is through employment of environmental CAMs. It is conceivable, however, that drone monitoring for release of 238 Pu in the event of a space launch mishap would give a greater chance of detection of plutonium particulate, because the particulate’s diffusion in air upon settling to the ground would make the probability of collecting a particle on a filter in a CAM very low.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

New Radiological Material Detection Technologies for Nuclear Forensics: Remote Optical Imaging and Graphene-Based Sensors

We developed new detector technologies to identify the presence of radioactive materials for nuclear forensics applications. First, we investigated an optical radiation detection technique based on imaging nitrogen fluorescence excited by ionizing radiation. We demonstrated optical detection in air under indoor and outdoor conditions for alpha particles and gamma radiation at distances up to 75 meters. We also contributed to the development of next generation systems and concepts that could enable remote detection at distances greater than 1 km, and originated a concept that could enable daytime operation of the technique. A second area of research was the development of room-temperature graphene-based sensors for radiation detection and measurement. In this project, we observed tunable optical and charged particle detection, and developed improved devices. With further development, the advancements described in this report could enable new capabilities for nuclear forensics applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2021

Liquid Metal Intrusion Testing Final Report

Los Alamos National Lab (LANL) is concerned about potential interactions between a tin-antimony pewter material used for localized radiation shielding and stainless-steel containers used for storage and movement of radioactive materials when containers are subjected to elevated temperatures. The specific concern is whether pewter constituents can diffuse into grain boundaries and lead to embrittlement of the stainless steel, potentially compromising the ability of stainless steel components to maintain containment. This work will provide a qualitative and quantitative understanding of the depth of penetration of the pewter coating constituents into stainless steel when the test coupons are subjected to a range of thermal environments. The results of the present work will allow LANL to determine conditions for safely using the pewter lined containers.

36 MATERIALS SCIENCE↗

DOE OSTI2015

RADIOLOGICAL SEALED SOURCE LIBRARY: A NUCLEAR FORENSICS TOOL

If a terrorist were to obtain and possibly detonate a device that contained radiological material, radiological forensic analysis of the material and source capsule could provide law enforcement with valuable clues about the origin of the radiological material; this information could then provide further leads on where the material and sealed source was obtained, and the loss of control point. This information could potentially be utilized for attribution and prosecution. Analyses of nuclear forensic signatures for radiological materials are generally understood to include isotopic ratios, trace element concentrations, the time since irradiation or purification, and morphology. Radiological forensic signatures for sealed sources provide additional information that leverages information on the physical design and chemical composition of the source capsule and containers, physical markings indicative of an owner or manufacturer. Argonne National Laboratory (Argonne), in collaboration with Idaho National Laboratory (INL), has been working since 2003 to understand signatures that could be used to identify specific source manufacturers. These signatures include the materials from which the capsule is constructed, dimensions, weld details, elemental composition, and isotopic abundances of the radioactive material. These signatures have been compiled in a library known as the Argonne/INL Radiological Sealed Source Library. Data collected for the library has included open-source information from vendor catalogs and web pages; discussions with source manufacturers and touring of production facilities (both protected through non-disclosure agreements); technical publications; and government registries such as the U.S. Nuclear Regulatory Commission’s Sealed Source and Device Registry.

National Nuclear Forensics Library↗

DOE OSTI2019

Laboratory-Scale Melter Systems for Simulated and Real AP-105 Tank Waste - 19301

The Hanford Tank Waste Treatment and Immobilization Plant is being constructed to vitrify the nuclear waste stored in underground tanks on the Hanford Site. Through this conversion process, waste components become partitioned into one of two product streams: glass and off-gas, including condensate. A small-scale melter system is appropriate for analyzing the waste-to-glass process because it does not require the high volumes of input waste or resource commitment of a large-scale melter system while providing dynamic information that can be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM) was sized to collect the relevant process and chemical information from obtainable volumes of Hanford tank waste and designed to operate with a continuous feeding process while periodically pouring glass product and collecting off-gas. 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. Multiple waste simulants were first tested in the non-radioactive CLSM to determine the processability and desired operating window for actual tank waste. About 12 liters of real waste from Hanford tank 241-AP-105 were received, filtered, and ion exchanged for removal of Cs-137. This quantity of waste was vitrified into 9.5 kg of glass product over the course of ∼15 hours of feeding to the CLSM. (authors)

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DOE OSTI2017

Characterization and Remediation of the Hanford 618-10 Burial Ground Vertical Pipe Units - 17121

The 618-10 Burial Ground is approximately 148 m by 143 m (485 feet by 470 feet) and consists of 12 trenches and 94 Vertical Pipe Units (VPU's). It is located approximately 8.1 km (5 miles) north of Richland, Washington in the 600 Area of the Hanford Site. The 618-10 Burial Ground received low- to high-activity waste (e.g., fission products and some plutonium-contaminated waste). Most of the waste originated at the 300 Area laboratory facilities. Lower activity radioactive material and shielded drums were disposed in trenches, while laboratory wastes with higher levels of radiation were disposed in Vertical Pipe Units (VPUs). There are different variants of VPUs. Some VPUs are simply corrugated culverts or standard pipes, 36 cm (14 in.) diameter and 3.1 m to 4.6 m (10- to 15-ft) long, buried vertically. Some VPUs are carbon steel pipes 25 to 61 cm (10 to 24 in.) in diameter and up to 4.6 m (15 ft) in length. The final design of the VPUs is 56 cm (22-in) diameter, 4.4 m (14.4 ft) long waste receptacles constructed by welding five 208 L (55-gallon) bottomless drums together end-to-end and burying them vertically, separated by approximately 3.1 m (10 feet) on center in rows approximately 7.6 m (25 feet) apart. The 618-10 Burial Ground contains approximately 94 VPUs. Historical documents indicate that shielding materials (e.g., concrete or soil) may have been added to the VPUs to reduce dose rates over the openings, and that each VPU was backfilled with sand and capped with concrete at the end of its service life. Historical documents also indicate that the VPUs were covered with approximately 0.6 m (2-ft) of soil when they were closed, and an additional 0.6 m (2-ft) of topsoil was later added for surface stabilization. In general, the process for VPU remediation was as follows: - Installed four cone penetrometers (CPTs) around the exterior of each individual VPU and sent radiation detection probes down the interior of the CPTs to get initial radiation data in order to calculate the overall curie content and create maps showing the dose rate distribution throughout the VPU. - Removed the CPTs and installed a 122 cm (48-in) diameter, 1.3 cm (0.5-in) thick steel over-casing using a vibrating hammer suspended from a crane that encompassed the VPU with some surrounding soil. - Using a Bauer BG-30 drill rig with a specifically designed augur bit and a specially designed confinement system the auger process was used to stabilize and size reduce the contents. - Once the auger reached a depth of 8.5 m (28 feet) the BG-30 drill stem was separated and additional characterization was completed by sending a radiation probe down the center of the drill stem to gather radiological data. - Once auguring and characterization was completed the waste was removed using a specially designed extended clamshell retrieval bucket and placed into individual 2.7 m x 1.5 m x 1.2 m (9-ft x 5-ft x 4-ft) metal boxes. - After the waste was placed into the metal waste boxes a low strength grout was added and the grout and waste material was mixed using a large rake device attached to an excavator. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 19092

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DOE OSTI2020

Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 20466

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating and testing prototype railcars that will comply with the safety standard S-2043. This standard was developed by the Association of American Railroads (AAR) specifically for railcars used to transport High-Level Radioactive Material (HLRM). AAR defines the term HLRM to include both spent nuclear fuel (SNF) and high-level radioactive waste (HLW). DOE is in the process of developing and testing prototype railcars that will satisfy Standard S-2043. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by the Department of Energy (DOE or Department). No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and a lack of Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Insulation Activation and Contamination in Low-Power Experiments

Among the diverse advanced nuclear reactor concepts being developed are designs that expect to operate at much higher temperatures than conventional light-water reactors, and therefore require alternative materials for components capable of withstanding these extreme temperatures. Insulation is one such component where alternative high-temperature industrial composites are being considered. Low-power experiments can provide an opportunity for advancing our understanding of material behavior under irradiation; however, these experiments are not without occupational hazards. The focus of this paper is the potential for activation of insulating composites when exposed to neutron fluence during reactor experiments, and to discuss the subsequent contamination of the testing area. This paper will also discuss practical measures for preventing the creation and inhalation of activated particulate matter (i.e., dust), such as engineering and administrative controls and personal protective equipment. This paper evaluates the suitability of several insulation materials for use in reactor experiments based on their activation from neutron irradiation. The materials considered are Pyrogel XTE by Aspen Aerogel, Cerablanket and Kaowool by Morgan Advanced Materials, Thermo-12 Gold by Johns Manville, and Maxsil CF6-2000 by McAllister Mills. Although these materials are not yet widely used in the nuclear industry, they are common in other industries due to their thermal insulating properties. These materials produce dust while handling that can pose health and safety hazards, which could be further exacerbated by activation. The analysis presented in this paper utilizes a deterministic method developed to predict the activation source term of the provided insulation materials. The method calculates the atomic density of each element within the composites and simulates their exposure to a specified neutron flux density representative of low power experimental conditions. During the irradiation phase and throughout the extensive cooling period, neutron absorptions and nuclear decays are tracked for relevant isotopes. The residual activity within each composite is then assessed to identify the material with the lowest potential radiological hazard to workers. Dust will be created during installation and removal of insulating materials and may also be generated during the experiment. Contamination of surrounding surfaces due to dispersion is highly likely unless preventative measures are taken. The results of this analysis are used to estimate an internal dose from inhalation of activated dust, which is significantly lower than the allowable whole-body dose. However, the risk of internal exposure may increase if the dust becomes a transport vector for nuclear material. The goals of this paper are to enhance the understanding of the potential activation and contamination hazards from the use of these insulating materials in low-power nuclear experiments and to foster safer work environments through informed decision-making and strategic planning. This work ultimately contributes to a broader understanding of how new non-nuclear materials for advanced reactors impact the radiological safety of co-located workers, helping to ensure the responsible management of radioactive materials during installation and decommissioning activities.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗