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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Metallic Low-Level Waste (LLW) Handling Options: An Optimal Life cycle-Based Approach for North American Operations - 19413

Daily activities associated with nuclear power plant operations produce large volumes of low-level contaminated waste. During operation, a small percentage of this low-level waste is metallic (< 1% by volume). However, this figure increases materially when a facility is finally decommissioned, with approximately 1000 m{sup 3} of contaminated metal waste generated for every 1 GW decommissioned (13% by volume). With global nuclear decommissioning efforts expected to increase until 2050, the amount of low-level metallic waste generated is expected to increase the global inventory by over an order of magnitude. The high cost of conventional surface storage and burial of radioactive metal, coupled with the projected increase in low-level metallic waste volumes provides strong incentive to identify cost-effective waste management solutions. Surface decontamination and thermal methods are proven options for decontaminating low-level metallic waste. Surface decontamination processes include chemical, electrochemical, or mechanical methods, which are used to remove contamination from exposed layers of metal. Surface decontamination is most effective when metal has low levels of activity localized on the outer layers. When the activity is more uniformly distributed in the metal (or the geometry is complex), alternative methods such as thermal decontamination are required. In the thermal decontamination process, contaminated scrap metal is melted in a furnace and held in a molten state for a period of time while the radioactive material is separated out through gravity. Radioactive material reports to slag which is handled separately, or to the exhaust which is handled using conventional gas handling technologies. Thermal methods of decontamination involving metal melting present an opportunity to reduce long-term storage cost with a reduction in waste volume, along with the opportunity to reuse the decontaminated metal. Methods for dealing with metallic low-level waste from nuclear facilities vary extensively across the globe. The approach employed at any given site may be set by a host of factors ranging from regulatory influence, public perception, local economics, available geology for underground storage and often, precedent established long ago by early operations. This paper examines several low-level metallic waste handling options including techniques for surface decontamination and metal melting, as well as a combination of both. The various methods are categorized according to their suitability for waste characterization and local/global installations. The effectiveness of each option is benchmarked against a range of criteria, including capital and operating costs, environmental impact, overall sustainability and risk management issues. The overall objective of this paper is to develop a methodology for determining the most beneficial low-level metallic waste management strategy for a facility once the quantity, type and nature of the waste is established on a long-term basis. It is the authors' hope that the methodology can be used as a tool to test a facility's current plans (which are primarily focused on long-term storage) and identify areas for optimization with an emphasis on overall cost and environmental footprint reduction. The trend towards long term planning to handle large amounts of contaminated metal, along with obtaining a social license to implement a sustainable waste management solution, are key decision-making variables that will be discussed in relation to the LLW handling options presented in this paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Assessment of Quality Management for Transport Packages Not Requiring Authority Design Approval - 18194

The majority of transports of radioactive materials are carried out in packages which don't need a package design approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. According to IAEA regulations the 'non-competent authority approved package types' are the excepted packages and the industrial packages of Type IP-1, IP-2 and IP-3 and of Type A. For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the regulatory density for packages requiring competent authority design approval. The countries apply different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. In order to help manufactures, which measurements and documents are needed for the assessment procedure, BAM developed a guideline. The guideline BAM-GGR 016 [5] describes the needed quality management for the design, manufacture, testing, documentation, use, maintenance and inspection of a packaging and the package. The poster to be presented will explain the different requirements of the quality assurance measures and give examples. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Application of the PHDS Germanium Gamma-ray Imager (GeGI) in Facility Decontamination Assessment - 19471

As part of a site wide investigation, U.S. EPA Region 5 had to assess the nature and extent of contamination in a former radioactive source manufacturing facility. This work involved verification of the radioisotope expected, ensuring no other radioisotopes were present, and quantification of the activity and areal extent of contamination throughout the facility. The device used to support this assessment work was the Germanium Gamma-ray Imager (GeGI) manufactured by PHDS Co. of Knoxville, Tennessee. GeGI is a germanium-semiconductor imaging detector system that provides identification, location, and quantification of radioactive materials within its field of view. GeGI provides the highest fidelity radiological assay information including the best gamma-ray energy resolution and spatial localization of any existing technology, including gamma-ray energy resolution of ∼0.3% full-width half-max (FWHM) at 662 keV and angular location resolution ∼ +/-1 degree. The GeGI is a mechanically cooled, portable system well-suited to field operations. PHDS Co. also provided on-site support during assessment operations and performed subsequent data analysis. Gamma-ray spectral data and associated images were collected with GeGI in six primary areas of the facility having significant residual contamination. A total of 61 gamma-ray spectrum/image sets were collected over the course of three working days (22 Compton images and 39 Pinhole/Compton images). During data collection, Compton imaging was used as the means by which to identify and prioritize hotspots. After contaminated regions were identified via Compton imaging, short-distance Pinhole spectrum/image sets were collected for additional analysis. All of the spectrum/image sets were analyzed to determine the radioisotope distribution within the GeGI field of view, as well as to provide an estimate of the activity associated with each hotspot. It was determined that the nature of residual contamination present was distributed across specific areas of elevated activity, where in some areas there was significant accumulation of contaminated material. It was also determined that no discrete sources of high-activity were present, as would be expected in the case of radioactive material product intended for sealed source manufacture. Results included the activity of residual contamination for each of the areas assessed, and an estimate of the physical area over which that contamination was distributed. It was anticipated by EPA that if decontamination were to be performed that it would involve some type of concrete scabbling or other means of physical removal of contaminated material. The GeGI data could be used to estimate the areal extent (square footage) of concrete removal that would be required for decontamination. Using the activity estimate data from the GeGI along with an estimated depth of concrete removal, planning could be performed for decontamination work (equipment and personnel), waste collection, packaging, transportation, and disposal, along with an associated cost estimate. With regard to health and safety, use of the GeGI reduced the radiation dose to the EPA R5 staff performing the assessment work. The GeGI is a non-destructive device that supports assessment of radioactively contaminated areas and sources from a distance, reducing the need for close contact to radiation, and the time necessary for personnel to operate near such hazards. Use of the GeGI in this facility assessment was our first attempt to apply this technology in what proved to be a quick and efficient decontamination assessment method. The GeGI can assess the nature and extent of residual contamination in a facility, first by quickly identifying the areas of elevated activity, followed by refined measurement and assessment of those elevated areas. The GeGI data can then be used to develop planning and a cost estimate of required decontamination and waste management activities. Since it's a non-destructive gamma identification and imaging device, it can reduce the distance of workers from radioactive sources and the dose incurred from assessment work. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Nuclear Forensics National Frameworks [Slides]

Nuclear forensics is the examination of nuclear or other radioactive material, or of evidence contaminated with radionuclides, in the context of legal proceedings under international or national law related to nuclear security. The goal of forensic science is to discover linkages among people, places, materials, and events. Nuclear forensics is an essential component of national response plans to events involving nuclear or other radioactive materials out of regulatory control and informs prevention, detection, and response.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Investigation Support Scenario-based Policy Discussion Development [Slides]

Nuclear forensics is the examination of nuclear or other radioactive material, or of evidence contaminated with radionuclides, in the context of legal proceedings under international or national law related to nuclear security. The goal of forensic science is to discover linkages among people, places, materials and events. Nuclear forensics is an essential component of national response plans to events involving nuclear or other radioactive materials out of regulatory control and informs prevention, detection and response.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Deciding Between an X-Ray and 137Cs Irradiator – It’s Not Just About Energy Spectra

Irradiators utilizing radioactive cesium-137 or cobalt-60 have been used for medical applications for many decades. This includes the irradiation of much of the nation’s blood supply, and radiation biology research. In 2005, the U.S. Nuclear Regulatory Commission was assigned to evaluate and prevent the misuse of radioactive materials with malicious intentions, with gamma-ray sources being high on the list. Licensees who possessed higher quantities of radioactive material of concern were required to set up increased security controls such as constant surveillance, controlled access, personnel background checks, and other security controls. As a result of introducing such regulations, and problems with disposing of high activity sources, many users of gamma-ray sources started to look into alternative technologies to replace their gamma-ray irradiators – the main alternative being an X-ray irradiator. To make an educated decision on whether a particular type of X-ray irradiator is equivalent to a particular type of cesium-137 irradiator for specific applications, one has to rely on relevant published comparison studies from other researchers, or perform the comparison studies on their own. This publication focuses on the authors’ technical performance comparison of two cesium-137 irradiator models and two X-ray irradiator models, for the purpose of determining whether the X-ray irradiator models could validly replace the cesium-137 irradiator models for certain applications. Utilizing 47 gram and 120 gram tissue-equivalent rodent dosimetry phantoms, our results indicate that, depending upon the user’s dose uncertainty budget and maximum size of specimens to be irradiated, the Model RS2000 160 kVp X-ray irradiator, Model XRAD-160 X-ray irradiator, or Model XRAD-320 X-ray irradiator, could successfully replace a cesium-137 irradiator. It is expected such equivalency could be extrapolated to other irradiated specimens of known sizes and density. Results also reveal that differences in inherent source geometry, field geometry and irradiation geometry can counter some of the influence due to differences in energy spectrum. It is hoped that this publication can be used as a guide for other similar studies, providing investigators guidance on important details that can make the difference between strong and weak comparison conclusions.

Cs-137 irradiator, X-ray Irradiator, self-containe↗
DOE OSTI2024

INL Alpha Background Radiation Study

Radiological emergencies that result in the dispersion of radioactive materials to the environment require the deployment of field monitoring teams to determine the quantity and location of the dispersed radioactive material. This information is utilized by the emergency operations center to aid in determining the appropriate protective action recommendations. The background alpha emission rates from normal environmental surfaces of soil, concrete and asphalt as measured by portable instrumentation used by field monitoring teams are not well documented. Idaho National Laboratory conducted a study of the alpha radiation background levels of the environmental surfaces in eastern Idaho to determine the normal background rates. Results for normal dry surfaces and those wetted by precipitation were measured and reported.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Decontamination and Decommissioning Activities of a Radiopharmaceuticals Production Facility - 18067

On May 2012, Best Medical Belgium S.A. (BMB), a manufacturer of pharmaceutical products located on the nuclear site of Fleurus, was declared bankrupt. According to the Belgian law, ONDRAF/NIRAS (Belgian Radioactive Waste Management Agency) was entrusted with the clean-up and decommissioning operations of the former BMB facilities, nowadays known as 'ONDRAF Site Fleurus' (ONSF). In June 2015, ONSF launched a Decontamination and Decommissioning (D and D) program, for which Tractebel: - Supports the definition of the clean-up and decommissioning strategies; - Drafts the Final Decommissioning Plan (FDP) and Safety Analysis Report (SAR) that will be submitted to FANC/AFCN (Belgian Federal Agency for Nuclear Control) in order to get the License for decommissioning; - Drafts the technical specifications of the public tendering process related to the D and D program; - Ensures the day to-day follow-up of the D and D activities on site. Four nuclear buildings are part of the D and D activities. One of them, building 14 (B14), represents the most challenging and extensive task with the clean-up of highly contaminated areas and its further decommissioning up to the final state. This building features two cyclotrons (CGR and IBA cyclotrons) successively used to produce radioisotopes, as well as hot cells and glove boxes used for their purification and the conditioning of the final pharmaceutical product. The present paper aims at presenting the most recent development steps of the clean-up phase regarding the adaptations of existing facilities and the upcoming activities in preparation of the decommissioning activities. To allow the sorting and the treatment of the radioactive material identified in the physical and radiological inventory, several workshops have been implemented within the B14 building in order to deal with radioactive materials arising from the clean-up and D and D activities. Following the conceptual design phase, Tractebel has actively monitored the on-site activities for the construction of those workshops such as: - The modification of the civil works; - The revamping of the ventilation system; - The installation of dedicated lifting equipment and workstations equipped with cutting devices; - The assembly of a docking station for filling the radioactive packages. The removal of the activated concrete surrounding the IBA cyclotron requires the erection of a containment structure and an extension structure. The containment structure, composed of light metallic beams, is located above the cyclotron and its irradiation zones. For operational security reasons and working efficiency, the cutting process of the concrete ceilings and walls will be performed from the top inside the containment structure by means of an eight tons overhead bridge crane. Concrete blocks resulting from the cutting process will then be lifted, without any loss of containment, towards the extension zone of the nuclear zone of the IBA cyclotron for further resizing, sorting and evacuation towards the identified waste routes. To prepare this task which will be performed during the decommissioning phase, Tractebel sets up the necessary technical specifications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Milestone Report - M3FT-15OR03120213 - A Literature Survey to Identify Potentially Problematic Volatile Iodine-Bearing Species Present in Off-Gas Streams

Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet U.S. regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3 H, 14 C, 85 Kr, and 129 I. Of these, 129 I has the longest half-life and potentially highest biological impact. Accordingly, control of the release of 129 I is most critical with respect to U.S. regulations for the release of radioactive material in stack emissions. Current U.S. Environmental Protection Agency regulation governing nuclear facilities (40 CFR 190) states that the total quantity of radioactive materials entering the general environment from the entire uranium fuel cycle, per gigawatt-year of electrical energy produced by the fuel cycle, must contain less than 5 mCi of 129 I. The study of inorganic iodide in off-gas systems has been almost exclusively limited to I 2 , and the focus of organic iodide studies has been CH 3 I.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

LANL Environmental ALARA Program Status Report for CY 2014

Los Alamos National Laboratory is committed to ensuring that radiation exposures of members of the public and the environment from LANL operations, past and present, are as low as reasonably achievable (ALARA). A Finding (RL.2-F-1) and Observation (RL.2-0-1) in the NNSA/ LASO report, September 2007, Release of Property (Land) Containing Residual Radioactive Material Self-Assessment Report, indicated that LANL had no policy or documented process in place for the release of property containing residual radioactive material. In response, LANL developed PD410, Los Alamos National Laboratory Environmental ALARA Program. This document was officially published as a Laboratory-wide policy and became effective on August 8, 2008 and was updated in 2011. The document provides program authorities, responsibilities, descriptions, processes, and thresholds for conducting qualitative and quantitative ALARA analyses for prospective and actual radiation exposures to the public and the environment from DOE activities conducted on site.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

LANL Environmental ALARA Program Status Report for CY 2015

Los Alamos National Laboratory (LANL) ensures that radiation exposures to members of the public and the environment from LANL operations, past and present, are below regulatory thresholds and are as low as reasonably achievable (ALARA) through compliance with DOE Order 458.1 Radiation Protection for the Public and the Environment, and LANL Policy 412 Environmental Radiation Protection. In 2007, a finding (RL.2-F-1) and observation (RL.2-0-1) in the NNSA/ LASO report, September 2007, Release of Property (Land) Containing Residual Radioactive Material Self-Assessment Report, indicated that LANL had no policy or documented process in place for the release of property containing residual radioactive material. In response, LANL developed PD410, Los Alamos National Laboratory Environmental ALARA Program. The most recent version of this document became effective on September 28, 2011. The document provides program authorities, responsibilities, descriptions, processes, and thresholds for conducting qualitative and quantitative ALARA analyses for prospective and actual radiation exposures to the public and t o the environment resulting from DOE activities conducted on the LANL site.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

LANL Environmental ALARA Program Status Report for CY 2016

Los Alamos National Laboratory (LANL) ensures that radiation exposures to members of the public and the environment from LANL operations, past and present, are below regulatory thresholds and are as low as reasonably achievable (ALARA) through compliance with DOE Order 458.1 Radiation Protection for the Public and the Environment, and LANL Policy 412 Environmental Radiation Protection (LANL2016a). In 2007, a finding (RL.2-F-1) and observation (RL.2-0-1) in the NNSA/ LASO report, September 2007, Release of Property (Land) Containing Residual Radioactive Material Self-Assessment Report, indicated that LANL had no policy or documented process in place for the release of property containing residual radioactive material. In response, LANL developed PD410, Los Alamos National Laboratory Environmental ALARA Program. The most recent version of this document became effective in 2014 (LANL 2014a). The document provides program authorities, responsibilities, descriptions, processes, and thresholds for conducting qualitative and quantitative ALARA analyses for prospective and actual radiation exposures to the public and t o the environment resulting from DOE activities conducted on the LANL site.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

Waste Encapsulation and Storage Facility Stabilization and Ventilation Project-18121

Department of Energy contractor CH2M HILL Plateau Remediation Company (CH2M) is responsible for the safe and compliant operation of the Waste Encapsulation and Storage Facility (WESF), a nuclear materials storage facility at Hanford containing 1,936 highly radioactive capsules containing cesium and strontium. The capsules are currently stored under approximately 4 m (13 ft) of water. The water shields workers from radiation and helps cool the capsules. WESF is a large, concrete structure nominally 48 m (160 ft) long, 30 m (98 ft) wide and 12 m (40 ft) high with the pool cells being approximately 4 m (13 ft) below grade level. It is a robust facility with walls ranging from 0.3 m to 1 m (1 ft to 3 ft) thick. While the capsules are currently in a safe configuration, WESF is an aging facility, and DOE and CH2M are modifying the facility to ensure continued safe and compliant operations and to support the eventual transfer of capsules to dry storage. Dry storage would eliminate the possibility of a release of radioactive material in the unlikely event of a major earthquake that might result in loss of pool storage water and overheating and breach of one or more capsules. The project was divided into two major deliverables: 1) Ventilation: replacement of the existing ventilation system which included a below-grade high efficiency particulate air (HEPA) filter pit with an above grade HEPA filter skid and new fans and ducting. 2) Stabilization: grouting deactivated facility hot cells and other void spaces to the maximum extent practical to reduce material at risk from legacy contamination. The aging ventilation system relied on HEPA filters that exceeded their operational life. Due to the high levels of radionuclide contamination, replacing the filters was impractical and highly hazardous. CH2M installed a new ventilation system to provide safe and compliant ventilation for future operations. The project used an existing, unused ventilation skid that was designed for another project. Repurposing the unused skid presented additional technical challenges to modify the system for the project specific requirements. The new system is above ground, easier to maintain and is built to current standards. New connections inside the facility were installed to allow the new ventilation to operate effectively. The stabilization project is unique in that core drilling and grouting was performed in an operating Hazard Category 2 nuclear facility, which brought forth safety basis and environmental regulatory challenges. The hot cells were isolated from all existing support utilities including electrical, process chemicals, manipulator ports, instrument air, etc. to seal the hot cells and prevent highly contaminated supernate or grout from leaking into adjacent galleries. A high flow grout was used to maximize the distance between injection ports. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Recent Advances in Nuclear Forensic Chemistry

Nuclear forensics is a rapidly growing field with a widespread impact on global nonproliferation. From the beginning, nuclear forensic capabilities have been inextricably bound to existing limitations within chemistry and materials science. As nuclear forensics has continued to develop as a unique field of research, new advances in chemistry have revolutionized our ability to rapidly and quantitatively analyze radioactive materials for attribution. In particular, microanalytical techniques originally developed for nanoscience, such as micro-XRF, SIMS, and SEM-EDX, have enabled spatially resolved chemical analyses to be performed with only nanograms of material. Such techniques have redefined the paradigm for pre- and postdetonation nuclear forensics, as large quantities of material are no longer necessary for full characterization and attribution. This Review explains how modern nuclear forensic techniques have developed from their origins in chemistry and provides real-world examples of how these techniques have been used to characterize radioactive materials, including fallout from nuclear detonations such as the 1945 Trinity test.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Management of Radioactive Wastes (RWs) and Disused Sealed Radioactive Sources (DSRSs) in Bangladesh: Recent Activities and Challenges Ahead - 16039

The use of radioactive sources in different applications and practices gives rise to various types of RWs that may be detrimental to human health and the environment if not properly managed. Since early sixties of the last century, radioisotopes have found diversified and incremental uses in Bangladesh in the field of medicine, industry, agriculture, food irradiation, research and education and so on. Currently, the country is operating nuclear facilities like 3 MW TRIGA research reactor (TRR), radioisotope production (RIP) facility as well as particle accelerators. Recently, the country has decided to establish nuclear power plants in order to cope with its increasing demand of electric power. The first nuclear power plant is anticipated to come into operation by 2021-2022. Consequently, a notable amount and range of RWs including DSRSs are being expected to be generated from these diversified applications of radioactive materials in near future. Bangladesh has limited facilities and capabilities to manage the RWs which are being currently generated from the conventional activities and practices with radioactive materials. However, the country is also working to improve its necessary capabilities and infrastructures for the management of RWs expected to be generated from both conventional and nuclear activities in future in a comprehensive, integrated and safer manner, ensuring that no undue burden is imposed on present and future generations and on the environment. This paper illustrates the current activities as well as its future plans on the management of RWs and DSRSs in the country. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Graphite Specimen Mass Measurements for NESHAP Calculations

The Advanced Graphite Creep (AGC) experiments were designed to irradiate various types of graphite specimens to gain an understanding of how the physical and thermal properties of the graphite change. The graphite specimens are machined from nuclear grade graphite that is very pure relative to other graphite types. However, it still contains parts per million (PPM) levels of metal contaminants that are dispersed throughout it. During irradiation in the Advanced Test Reactor (ATR) these metal contaminants become activated resulting in the graphite specimens being classified as a radioactive material. Following their irradiation, a variety of measurements and experiments are performed on these specimens that require them to be heated above 100°C.Whenever a radioactive material is heated to above 100°C the National Emission Standards for Hazardous Air Pollutants (NESHAP) calculations require the activity content, in Curies, of the isotopes that could potentially be released to the environment. The purpose of this ECAR is to document the method by which the mass loss of each sample is determined and how the isotopic content of this mass loss is determined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Characterisation of Specimens from a Former Soviet Thorium Processing Facility to Support Site Remediation

The former Prydniprovski Chemical Plant (PChP) is a large ex-soviet nuclear materials processing site of roughly 260 hectares, located in central Ukraine. PcHP was in operation between 1948 and 1991 for the purposes of processing uranium and thorium ore into ore concentrate. During its operational lifetime, tens of millions of tonnes of radioactive waste was generated on the site. A consequence of these activities, coupled with the mismanagement of the resulting radioactive materials and by-products, is that approximately 45% of the southern part of the site has become radioactively contaminated. This area comprises some 100-200 buildings and structures, many of which are in a state of disrepair and dereliction. Around 20 of these buildings are heavily radioactively contaminated, with radiation levels from a few μSv/hr to a maximum of 4 mSv/hr reported. The European Commission is currently working with the Ukraine Ministry of Energy and Coal Use and the Nuclear Regulator to improve radiological safety and security at the site. In order for these remediation works to proceed, the site management needs to characterise the radioactive materials that are held there. As part of this activity, the team from the University of Bristol have undertaken a survey of one of the most heavily contaminated buildings, collecting a selection of in situ material samples. A suite of complimentary materials characterisation techniques (e.g. X-ray Fluorescence, Raman Spectroscopy, etc) have been deployed in order to determine the radionuclides and composition of the compounds present. Characterisation of these samples will allow the operators of the chemical plant to go forward with remediation works within the building.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗