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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Using FLAC{sup R} Modeling Software to Analyze the Stability of a LLRW Disposal Facility - 20031

Waste Control Specialists (WCS) wished to amend its existing LLRW disposal license such that it might use gravel aggregate instead of concrete as backfill around large component (LC) parts from decommissioned nuclear reactors. A classical mechanics evaluation of the strength of the LCs was presented to the Texas Commission on Environmental Quality (TCEQ), demonstrating that the LCs had adequate capacity to withstand the loads within the facilities, and that concrete was unnecessary for structural support. TCEQ requested that additional analyses of the LCs within the facilities be performed using FLAC{sup R} finite difference modeling software. Because FLAC{sup R} has the capability of modeling the support provided by the lateral earth pressures, the factors of safety against failure for the LCs increased. TCEQ found the FLAC{sup R} analyses presented in this report to be satisfactory, and granted WCS' license amendment request. This work demonstrates that FLAC{sup R} may be used advantageously to more accurately represent the complexities of the physical situation being modeled. (authors)

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

Sorbmatech{sup R}Cs, an Innovative Highly Selective Cs Sorbent for Wastewater Decontamination - 20065

The treatment of radioactive wastewater from the operating nuclear industry (energy production, reprocessing of spent fuel, decommissioning of shut-down facilities, etc.) or in case of incidental or accidental situation (such as Fukushima Daichii NPP accident) is an important issue requiring constant improvement of the technologies and techniques used to prevent radioactive releases into the environment and/or to reduce the production of secondary waste. CEA, Orano and CTI are collaborating since 2014, with the support of the French government's program 'investments for the future', to develop Sorbmatech{sup R}Cs, an innovative and patented selective sorbent (Grandjean and Barre 2014) to deal with the removal of radioactive {sup 137}Cs, one of the most hazardous isotope to be managed in nuclear waste, highly efficient either in agitated reactor or in column process. Sorbmatech{sup R}Cs sorbent consists of mesoporous silica particles loaded with potassium-copper hexacyanoferrate nanoparticles (KCu-HCF). This peculiar structure enables much faster sorption kinetics compared to denser crystalline structures, such as Silicotitanate or Zeolite sorbents, whose drawback consists in hindering the diffusion process through the material, and whose larger sorbent particles slow down the ion exchange process. The equilibrium of Sorbmatech{sup R}Cs is reached in few minutes, even at high flow rates in a fixed-bed column process (up to 10 m.h{sup -1} Darcy velocity) with low drop pressure. Moreover, Sorbmatech{sup R}Cs can be used over a wide pH range from 2 to 10 and has a very high selectivity towards {sup 137}Cs even in brine solution (saline, mineral or sea water). This new sorbent also has a K{sub d} higher than 10{sup 5} mL.g{sup -1} depending on the salinity of the wastewater, enabling a highly efficient removal of Cs trace in radioactive wastewater compared to Cs selective sorbent such as Zeolites and Silicotitanates. Its radiological resistance enables the treatment of high, medium and low activity wastewater. The CEA-Orano-CTI/ALSYS partnership leads to the development of a simple and low-cost synthetic route, which demonstrated large-scale manufacturing of the product on a batch of 100 kg. Two products have been developed and are ready for short run market introduction so far with grain size well adapted either for column processes or agitated reactors.. Finally, the industrial implementation has been studied and largely benefits from the properties of the sorbent. Indeed, its high sorption kinetics enable high flow rates resulting in a decrease of operation times, unit sizes, and volumes of secondary waste produced. Moreover, Sorbmatech{sup R}Cs is designed to be combined into a full and integrated solution, ranging from preliminary optimization of operating conditions through numerical dimensioning model, industrial implementation and treatment of the wastewater (through cartridge or preconditioned columns systems developed thanks to Orano's extensive expertise), to long-term stabilization of generated waste (under progress). (authors)

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

Test for Reconditioning RA Waste with Simulated Bitumen and Concrete in a 1,2 MW Plasma Test Facility - 20092

The operation and maintenance of nuclear power plants, the non nuclear fuel cycle, etc generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000 deg. C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified, oxidized and purified in an off-gas cleaning system. First the paper describes the new full-scale Plasma Melting Facility (PMF) at the Kozloduy Nuclear Power Plant in Bulgaria which was taken in nuclear operation In May 2018. The plant has a capacity of 250 tons per year and the maximum contact dose rates of the incoming waste is 2 mSv/h. Different mixtures of radioactive waste packed in 200 l drums were successfully treated resulting in a glassy slag free from liquids and organic material with an important volume reduction factor (VRF). The Project was co-financed through a grant by Kozloduy International Decommissioning fund (KIDSF) administrated by the EBRD through Bulgarian national funding. Plasma is a suitable technology for treatment of problematic waste or even reconditioning waste so Belgoprocess was contracted to do plasma tests with simulated conditioned waste types. One can do tests on a laboratory scale on smaller samples and torch capacities of e.g. 50 kW but Belgoprocess wanted to do more realistic and reliable tests. So Belgoprocess contracted Phoenix Solutions Co who has a full-scope test facility equipped with a 1200 kW plasma torch for full-scope treatment of simulated conditioned waste. For a first confidential contract simulated 200 l (55 gallon) bitumen drums were treated. The drums contained different pucks of compacted waste such as rags, used filters, granulates, etc. The pucks were stacked in the 200 l drums and subsequently embedded with bitumen. A total of 6 drums were treated in the plasma facility. For a second contract simulated homogeneous 200 l (55 gallon) concrete drums with on the one hand concentrates and on the other hand spent resins were selected. A total of 6 drums with concrete and spent resins were treated and melted in the plasma testing facility. The paper describe the test facility, volume reduction factor (VRF) of different waste streams and most important parameters. (authors)

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

Use of Unmanned Aerial Systems for Post-closure Waste Site Maintenance and Surveillance - 20143

The P- and R-Area Reactor buildings, located at SRS near Aiken, SC, were in-situ decommissioned by grouting below-grade portions of the buildings and demolishing some above-grade structures to grade level. Other reactor building structures were left above-grade and sealed to prevent human or animal access. Because the above grade-structures are expected to continue in their present state for hundreds of years, the condition of the building roofs is critical to mitigate rainwater intrusion. For this reason, building roof areas were strengthened with high strength concrete to ensure the long-term integrity of the roof structure. Periodic inspections of the roof structures are required to ensure that the roofs are functioning properly and to identify damage areas. The traditional method of inspecting closed reactor buildings requires the use of a helicopter and photographer to generate photographs and videos for review. This method proved sufficient but lacked the resolution and clarity that is required for a thorough inspection of the building structure. Fortunately, the Savannah River National Laboratory (SRNL) has established a Small Unmanned Aircraft System (sUAS) program using commercially available and custom-built remote-controlled aircraft. EC and ACP contacted the SRNL program to evaluate whether the sUAS technology could be employed for periodic reactor building inspections. The sUAS can fly within a few meters of the reactor buildings and hover, allowing for a more thorough aerial inspection. The initial sUAS inspection of the P-Area Reactor building was completed in February 2018 and at the R-Area Reactor building in August 2018 and were successful in providing higher resolution photos and videos. The sUAS inspections also revealed that vegetation had begun to grow on the roofs that could potentially damage the structure integrity. SRNL partnered with Virginia Polytechnic Institute and State University to build a custom heavy lift sUAS that could be equipped with herbicides to remotely treat the vegetative growth. Herbicides were successfully applied to the R-Area Reactor building roof using the custom sUAS in September 2018. The use of unmanned aircraft systems at SRS to perform aerial inspections and herbicide treatment in otherwise inaccessible areas has proven to be an efficient and cost-effective technology that provides high value, increases knowledge of facility conditions, and provides for early detection of damage. Periodic inspections of the reactor building roofs using sUAS technology are performed safely, efficiently, and at a significant cost and schedule savings while reducing emissions, noise, and fossil fuel use. The use of sUAS equipment for building inspections and herbicide application at SRS is unique within the DOE complex. The goal of SRS is to apply this remote technology to other waste unit operations and maintenance activities in the future. (authors)

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

Influence of Borate on Sorption of Strontium and Barium to Calcium Silicate Hydrate - 20148

Radioactive waste with the decommissioning process at the Fukushima Daiichi Nuclear Power Plant (FDNPP), contains borate with relatively high concentration. As with general low-level wastes, such waste is also assumed the solidification with cementitious materials which calcium silicate hydrate (C-S-H) is one of the main components. In this study, the sorption behavior of strontium (Sr) ions and barium (Ba) ions to C-S-H in the presence of borate was examined by the sorption experiments without drying process. In the sorption experiments, C-S-H was synthesized by the mixture of calcium oxide, fumed silica and a solution with setting the Ca/Si molar ratios to 0.4, 0.8, 1.2 and 1.6. The total amount of a solid phase was 1.5 g, and the total volume of the solution containing borate and adsorbed ions (Sr or Ba) was 30 mL. The concentration of borate was set to 0 M or 0.6 M and the sodium/boron (Na/B) ratios was set to 0.2 or 1.0 by mixing boric acid and sodium hydroxide. The concentration of Sr ions or Ba ions was adjusted to 1 mM. The sorption period was set to 7 days. In the results, the sorption ratios of Sr ions and Ba ions for C-S-H exceeded 80% under all experimental conditions, although the sorption competition of cations with Na ions and the inhibition of the C-S-H formation with generating borate compounds observed by Raman spectroscopy might slightly reduce the sorption of them. These suggest that cationic radionuclides will be immobilized by C-S-H contained in the cements even in the presence of borate. (authors)

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

Amphiphilic Block Copolymers for Flocculation and Hydrophobization of Legacy Waste Suspensions in Flotation Driven Dewatering Operations - 20162

Flotation has been effectively demonstrated as a low footprint, high efficiency separation process when utilising simple anionic surface modifying agents such as sodium dodecyl sulphate (SDS). These agents increase hydrophobicity to remove suspended cationic surface charged particulates such as Mg(OH){sub 2} from waste suspensions similar to those at British nuclear fuel management sites- where corrosion processes have degraded fuel cladding. The technology is advantageous as it could effectively be retrofitted into existing waste management facilities at nuclear waste management sites such Sellafield and Hanford as the simplicity (no moving parts) and size of this technology coupled with low cost of construction could be rapidly deployed to aid in risk and hazard reduction decommissioning operations with minimum impact to secondary waste generation. Previously, sodium dodecyl sulphate has been deployed as a collector agent to increase the hydrophobicity of Mg(OH){sub 2} suspended particles, achieving a maximum particulate extraction percentage of 93%. To achieve further particulate extraction, flotation performance has been determined to be a function of the hydrodynamic limitations extracting fine material (>20 μm). Fine particulates have been shown in the minerals industry to have an inability to overcome slipstreams created from rising bubbles in flotation cells, likely due to the low mass and thus momentum of the colloidal material, preventing interaction between the hydrophobized particles and air-water interfaces (bubbles). To combat this mass transfer resistance poly(acrylic acid)-block-poly(n-butyl acrylate) amphiphilic diblock copolymers were synthesised via reversible addition-fragmentation chain-transfer (RAFT) polymerisation. The anticipated mode of action of these polymers involves the hydrophilic poly(acrylic acid) block polymer acting as a flocculation agent to increase the mass of the aggregates, and the hydrophobic poly(n-butyl acrylate) modifies the hydrophobicity of the resulting flocs. The concentration of this polymer (termed a macro-collector) was varied and the change in particle size distribution of the particulate suspension was analysed using static light scattering. The flotation performance was then characterised by several methods including percentage particulate recovery, volume reduction factor and residual bulk particulate concentration. This was then benchmarked against the traditional surfactant collector agent SDS. (authors)

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

Application of FREMES to Characterize and Sort Soil During Site Remediation - 20182

During decommissioning of buildings and the related release of nuclear sites, large amounts of bulk materials such as soil must be measured for clearance. These measurements are required to demonstrate that for example potentially contaminated soil is below the respective radiological limits. Doing so also helps to minimize the amount of bulk material which has to be disposed of as nuclear waste. NUKEM Technologies Engineering Services GmbH (NUKEM) applies their FREMES technology in a project at the former fuel element factory of FBFC International at Dessel, Belgium. The purpose is to measure and check for potential contamination within the excavated soil. The project started in 2017 with design and procurement of the necessary equipment. In October 2017 the FREMES system was installed at the site. After installation and an extensive test program the FREMES facility went fully into operation in July 2018. Original foreseen to measure between 4.000 and 8.000 tons of material, it has currently (mid 2019) measured 23.600 tons. It is expected that another 5.000 tons still have to be measured before the project successfully ends. (authors)

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

Co-Disposal Waste Package Loading Options for DOE SNF and HLW - 20331

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Modeling Cost of Offshore Carbon Storage in Saline Reservoirs

Poster presentation to the AAPG-SEG-SPE CCUS 2024 Conference. Offshore saline reservoirs provide a significant and accessible resource for geologic carbon storage (CS). The offshore environment requires distinct approaches to site selection, operations, monitoring, and risk that affect the technoeconomic assessment of offshore CS projects. The National Energy Technology Laboratory (NETL) has developed a CS cost model for offshore saline reservoirs known as CO2_S_COM_Offshore. Based on NETL’s widely used CO2_S_COM cost model for onshore saline CS, CO2_S_COM_Offshore enables technoeconomic analysis of CS in offshore areas. This model comprehensively incorporates multiple facets of offshore CS projects, from regional evaluation and site selection to permitting, transport, operations, monitoring, site closure, and decommissioning. Developed to model cost for offshore United States (US) Exclusive Economic Zones, aspects of this model can be adapted to international projects. Presented at the SPE/AAPG/SEG Carbon Capture Utilization and Storage Conference in Houston, TX, March 11-13, 2024.

Mark-Moser, Mackenzie K.↗
DOE OSTI2024

Energy and Carbon Payback Times for Modern U.S. Utility Photovoltaic Systems

Manufacturing and operating a PV system consumes non-renewable energy and produces carbon emissions, as does end-of-life handling when PV systems are eventually decommissioned. To fully account for PV's contribution toward decarbonization, these life cycle impacts must be quantified. A 2023 NREL LCA of utility PV systems in the United States Study show energy payback times between 0.5 and 1.2 years and carbon payback times between 0.8 to 20 years, depending on the system install location. Payback times are affected by the amount of local solar radiation, and carbon payback time is significantly affected by the carbon-intensity of the local grid it offsets, as well as the future projected grid mix for the location it is installed. In an average U.S. location, carbon payback times are less than 2 years.

carbon payback time↗
DOE OSTI2024

01-07 Reactor Supplemental Shielding For the NRIC DOME Test Bed 01-07 Reactor Supplemental Shielding For the NRIC DOME Test Bed 2

A radiation shielding system is needed to protect personnel and equipment from harsh neutron and photon fluxes during: Reactor operations Post shut down operations Disassembly Decommissioning Examples of RSS driving requirements Radiological Controls Operating dose rate: the dose rate during normal reactor operations shall be <0.5 mrem/hr at the exterior wall of DOME containment Outdoor radiation limit: shielding shall provide the dose to a facility worker outside of DOME to <0.05 mrem/hr at 50 ft from the exterior wall of the DOME Temperature Reactors may have operating temperatures up to 800C The RSS shall be capable of removing up to 100 KW of heat during normal operating conditions the concrete structure has an upper temperature limit of 100°C

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Criticality Safety Support for Deactivated Facilities at Savannah River Site

The Savannah River Site (SRS) has a number of nuclear facilities. Some date back to the mid-1950s while others are actively being built today. As facilities have aged and missions changed, a number of the older nuclear facilities have gone through various stages of deactivation, decontamination, decommissioning, and dismantlement (D4). Due to the nature of the site’s historic missions a number of these facilities processed and utilized significant quantities of fissionable materials. D4 activities have now spanned multiple decades and as such have spanned multiple regulatory and industry standards have variety of documentation levels associated with them.

Stover, Tracy E.↗
DOE OSTI2024

Leveraging NREL's ResStock & ComStock Dataset to Evaluate Building Stock Electrification: Preprint

Residential and commercial buildings accounted for 40% of U.S. energy consumption in 2022 and represent a significant opportunity for decarbonization through energy efficiency and electrification, and for grid planning. Building stock energy modeling is a powerful tool that can evaluate what-if scenarios as utilities, municipalities, policymakers, building owners and others work towards equitable building decarbonization and climate goals. This presentation will highlight several high-impact use cases of the National Renewable Energy Laboratory (NREL)'s highly granular, bottom-up building stock energy modeling tools, ResStock and ComStock. These use cases cover a wide range of project scale, from neighborhood electrification analysis and municipality long-term energy planning, to state energy code development and national policy evaluation. This presentation will showcase specific real-world applications for which ResStock and ComStock have been utilized across the country, including California codes and standards cost-effectiveness analysis, New York City affordable housing electrification cost gap analysis, and California targeted electrification and gas decommissioning analysis. For each use case, this presentation will illustrate how ResStock and ComStock played a crucial role in accurately characterizing regional building stocks, providing discrete and aggregated end-use load shapes, and calculating lifecycle consumption, emissions, and costs for a variety of building electrification strategies and scenarios. Finally, this presentation will demonstrate how the data provided by ResStock and ComStock can help unlock significant outcomes for these use cases, including but not limited to, customer bill impact, incentive and program design, and energy equity analyses.

building stock modeling↗
DOE OSTI2024

"Source Term Modeling for Advanced Gas Micro-Reactors"

Maintaining the safety of the public, environment, and operating personnel is the most important factor in designing, operating, maintaining, and decommissioning nuclear reactors. In recent years, there has been a growing interest in the development of micro-reactors employing TRi-structural ISOtropic (TRISO)-coated particle fuel. In gas reactors, TRISO fuel plays an important role in the safety case for high temperature reactors because of the fission product retention properties of the fuel. This ability enables the use of a functional containment strategy for the reactor where multiple barriers are used to prevent fission product release to the environment. Part of the safety analysis of these advanced reactors is the assessment of radionuclide releases under normal and accident conditions through the multiple credited safety barriers. Using conservative assumptions, a mechanistic analysis can be performed to quantify these releases that combines the probabilistic assessment of failure with analytic solutions to radionuclide transport equations. Source term modeling for TRISO fuel has been performed for previous reactor designs; however, these models are outdated, in many cases proprietary, and need updates to be applied to the current state of TRISO fuel technology and alternative gas reactor core configurations [1]. Currently, the only publicly available source term assessment for gas reactors is an expert-based Monte Carlo simulation based on the effectiveness of the fuel kernel, coating layers, and graphite block in a modular high temperature gas reactor [2]. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that could be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. The release is calculated by the diffusion of the key safety important fission products through the kernel, silicon carbide (SiC), graphite for both intact and defective TRISO particles based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system to remove fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. The model then can calculate the fission product release for any transient temperature profile and fission product releases can then be used to assess radiological dose to the workers and the public using conventional dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy (DOE) Advanced Gas Reactor (AGR) TRISO fuel development program. The model is coded in python with inputs and outputs in excel spreadsheets, as well as python plotting utilities to aid in the interpretation of the results. References: [1] INL, NGNP Mechanistic Source Term White Paper, INL-10-17997, July 2010. [2] David A. Petti, Richard R. Hobbins, Peter Lowry, Hans Gougar, “Representative Source Terms and The Influence of Reactor Attributes on Functional Containment in Modular High Temperature Gas-cooled Reactors,” Nuclear Technology, Vol. 184, p. 181-197, Nov. 2013.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Part IV: Remedial Action

This part of the book focuses on the remedial action where the remedy is selected and approved with associated regulatory documentation, detailed designs are prepared, and construction and/or implementation of the remedy is completed. The application may include remediation of environmental media (e.g., subsurface sediments or surface water sources), waste processing, and/or deactivation and decommissioning of historical facilities. Once a preferred remedy is implemented, the final steps include long-term stewardship and site closure (discussed in Part V). Refer to Section 3.2 of Chapter 1, Remediation Strategy for Complex Waste Sites, for context on how these steps fit into the broader remediation process.

Johnson, Christian D.↗
DOE OSTI2025

Development of a Real-Time Neutron Noise Analysis System for Fuel Debris Removal at Fukushima Daiichi

The decommissioning of Units 1, 2, and 3 at Fukushima Daiichi presents unique challenges, particularly in mitigating the risk of re-criticality during fuel debris removal. Disturbing previously stable debris configurations has the potential to cause changes in the multiplication of the system, necessitating real-time monitoring to ensure operational safety. Current neutron detection systems, primarily passive, are not optimized for continuous real-time analysis and are limited in their ability to detect rapid changes in system reactivity.

Neutron Detection Systems↗
DOE OSTI2024

Initial Progress on Tank Bottom Repair for the DOE EM Tank Waste R&D Program - 25089

Several years prior to the Lab23-EM001 Call Announcement, Washington River Protection Solutions (WRPS) had done a review in identifying tank repair feasibility. [1, 2] The most viable methods of repair identified for the Floor of a Primary Tank identified were Grout Stabilization and Polymer Grout. Grout Stabilization was described as a self-leveling grout which fills the entire bottom of the tank and creates a new or false bottom. These types of materials have been used at the Savannah River Site (SRS) to fill and close waste tanks as well as contaminated spaces within decommissioned reactors. No mention was made as to the chemical composition of these Stabilizing Grouts in the review, however, grouts used at SRS for the mentioned applications have always been a cementitious grout material. A thickness of approximately 12 inches was suggested in the review. It was also suggested that an additional polymeric coating could be applied to this fresh and clean grout surface.

BLUE, KAREEN↗
DOE OSTI2025

Reactors Section Presentation - INL, HPS Society 2025

The ongoing Nuclear Renaissance includes many next generation reactor designs and testing, this means translates to increased demand for Health Physicists. These opportunities are not limited to just reactors but the entire fuel cycle from mining and extraction from existing tailings, R&D support efforts at new start-up companies, to shielding modeling and design, to Recycling and Reprocessing, and eventually Decommissioning.

Health Physics↗