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At least 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Waste Minimization of HyBRID Decontamination Process - 17060

The Korea Atomic Energy Research Institute (KAERI) developed a Hydrazine base reductive metal ion decontamination (HyBRID) and related pre-oxidation techniques for use in decommissioning a nuclear facility. Inorganic chemicals were used to increase the integrity of the radioactive waste in a repository. Decomposition and precipitation of residual reagents with inorganic chemicals were introduced to minimize the final waste volume. The waste volume was calculated theoretically and demonstrated in a small-scale waste treatment system and bench-scale candle filtration system. The total waste volume of the HyBRID process was less than 1/10 of that of commercial processes, and spent ion-exchange resins were significantly reduced to 5% of that of commercial processes. (authors)

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

Development of Remote Laser Decontamination System - 18183

A cleaning method using a laser ablation technology is widely used in various industries such as automobile, aviation, and semiconductor industries. Recently, applying this technology to decontamination works has been receiving attention as a useful and effective decontamination method. The laser decontamination method allows for decreasing a generation of secondary radioactive waste because materials like abrasive and water are not required. This is one of big advantages in use for decontamination works. Additionally, the system is compact and requires less than 100 W electric power. These features provide a low running cost. Remote operation obviously plays an important role in terms of a dose reduction in highly-radioactive and contaminated environment. The laser decontamination method is also suitable for remote works using manipulators as the laser head is light and compact, and a reaction force during operation is small, compared to the other types of decontamination tools. This paper provides an outline and a capability of the remote laser decontamination system developed by IHI Corporation (IHI) with some results of component tests for this system. (authors)

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

Case study in corporate memory recovery: Hanford tank farms miscellaneous underground waste storage tanks - 15344

In addition to managing the 177 underground waste storage tanks containing 212,000 m{sup 3} (56 million gal) of radioactive waste at the U. S. Department of Energy's Hanford Site 200 Area Tank Farms, Washington River Protection Solutions LLC is responsible for managing numerous small catch tanks and special surveillance facilities. These are collectively known as 'MUSTs' - Miscellaneous Underground Storage Tanks. The MUSTs typically collected drainage and flushes during waste transfer system piping changes; special surveillance facilities supported Tank Farm processes including post-World War II uranium recovery and later fission product recovery from tank wastes. Most were removed from service following deactivation of the single-shell tank system in 1980 and stabilized by pumping the remaining liquids from them. The MUSTs were isolated by blanking connecting transfer lines and adding weatherproofing to prevent rainwater entry. Over the next 30 years MUST operating records were dispersed into large electronic databases or transferred to the National Archives Regional Center in Seattle, Washington. During 2014 an effort to re-acquire the historical bases for the MUSTs' published waste volumes was undertaken. Corporate Memory Recovery from a variety of record sources allowed waste volumes to be initially determined for 21 MUSTs, and waste volumes to be adjusted for 37 others. Precursors and symptoms of Corporate Memory Loss were identified in the context of MUST records recovery. (authors)

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

Long-term performance of engineered barriers for high-level waste repositories - 15358

The evolution of the engineered barrier system (EBS) of geological repositories for high-level radioactive waste (HLW) and spent fuel (SF) has been the subject of many national and international research programs during recent years. The emphasis of the research activities was on the elaboration of a detailed understanding of the complex coupled thermo-hydro-mechanical and -chemical (THM-C) processes, which are expected to evolve in the early post-closure period in the near field. From the perspective of radiological long-term safety, an in-depth understanding of these coupled processes is of great significance, because the evolution of the EBS during the early post-closure phase may have a non-negligible impact on the safety functions at later stages of the repository's lifetime. Process interactions during the re-saturation phase (heat pulse, gas generation, non-uniform water uptake from the host rock) could impair the homogeneity of the safety-relevant properties in the EBS (e.g. swelling pressure, hydraulic conductivity, diffusivity). The 7. Framework EURATOM PEBS project (long-term Performance of Engineered Barrier Systems) was initiated in 2010 and completed in March 2014. The project aimed at evaluating the sealing and barrier performance of clay-based EBS over time. The project approach included experiments, model development and consideration of the potential impact on long-term safety functions. The specific project aims were to: -deepen the understanding of the THM and THM-C evolution of the EBS system with time, - provide an enhanced quantitative basis for relating the evolutionary behavior to the safety functions, - clarify further the significance of residual uncertainties for long-term performance assessment. It can be concluded that uncertainties in process understanding occurring in the re-saturation period have been better constrained through PEBS studies, thus the uncertainties in the long-term performance of bentonite barriers have been reduced in some areas. Improvements are in the areas of evolution of materials properties and model development and testing at various scales. On-going in-situ experiments in underground research laboratories may play an important role in further confirming bentonite performance over periods of 10 - 20 years. (authors)

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

Radiation damage effects in chlorite investigated using microfocus synchrotron techniques

A detailed understanding of the mechanisms and effects of radiation damage in phyllosilicate minerals is a necessary component of the evaluation of the safety case for a deep geological disposal facility (GDF) for radioactive waste. Structural and chemical changes induced by a-particle damage will affect these mineral’s performance as a reactive barrier material (both in the near and far-field) over timescales relevant to GDF integrity. In this study, two examples of chlorite group minerals have been irradiated at a-particle doses comparable to those predicted to be experienced by the clay buffer material surrounding high level radioactive waste canisters. Crystallographic aberrations induced by the focused 4He2+ ion beam are revealed via high-resolution, microfocus X-ray diffraction mapping. Interlayer collapse by up to 0.5 Å is prevalent across both macro- and micro-crystalline samples, with the macro-crystalline specimen displaying a breakdown of the phyllosilicate structure into loosely-connected, multi-oriented crystallites displaying variable lattice parameters. The damaged lattice parameters suggest a localised breakdown and collapse of the OH- rich, ‘brucite-like’ interlayer. Microfocus Fe K-edge X-ray absorption spectroscopy illustrates this defect accumulation, manifest as a severe damping of the X-ray absorption edge. Subtle Fe2+/Fe3+ speciation changes are apparent across the damaged structures. A trend towards Fe reduction is evident at depth in the damaged structures at certain doses (8.76 x 1015 alpha particles/ cm2). Interestingly, this reductive trend does not increase with radiation dose, indeed at the maximum dose (1.26 x 1016 alpha particles/ cm2) administered in this study, there is evidence for a slight increase in Fe binding energy, suggesting the development of a depth-dependant redox gradient concurrent with light ion damage. At the doses examined here, these damaged structures are likely highly reactive, as sorption capacity will, to an extent, be largely enhanced by lattice disruption and an increase in available ‘edge’ sites.

Bower, William R.↗
DOE OSTI2020

Lessons Learned from the First Year of Property Remediation - 20475

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste (LLRW) in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL) on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI compliantly, on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The larger of the two PHAI scopes of work - the Port Hope Project - involves the cleanup of approximately 1.2 million cubic metres of historic LLRW from 18 major sites in Port Hope, the construction of an engineered aboveground mound where the waste will be safely contained and the long-term maintenance and monitoring of the facility. Construction of the waste water treatment plant, the first three of four cells in the mound and supporting infrastructure is complete, with construction of the fourth cell in progress. In 2017 with the movement of on-site waste at this facility began and is now complete. Waste excavation and transportation from off-site locations within the community began in 2018; the cleanup of four temporary storage sites is now complete and remediation at several other major sites is underway. The Port Hope Project also includes a survey of the interior and exterior investigation of approximately 4,800 properties, primarily residential, in the urban area of Port Hope for the presence or absence of historic LLRW. The current projection is that approximately 1,250 privately owned residential and commercial properties known as Small-Scale Sites (SSS) will be identified as having waste exceeding the PHAI clean-up criteria. This paper will outline some of the challenges and lessons learned from the first year of SSS remediation including contractual, data and design issues; interaction with and managing expectations of stakeholders; weather, site and structure conditions as well as municipal permitting, resourcing, transportation and remediation verification. (authors)

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

The Next Generation System Analysis Model: Capabilities for Simulating a Waste Management System - 19131

The U.S. Department of Energy (DOE) Office of Nuclear Energy Office of Spent Fuel and Waste Disposition's Integrated Waste Management System (IWMS) program is applying integrated waste management system architecture analysis, system engineering, and decision analysis principles to inform potential future decisions regarding potential nuclear waste management system architectures. Architecture analyses of the IWMS are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all major aspects of the back end of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The system architecture analyses consider alternative disposition pathways for commercial spent nuclear fuel (SNF) being stored dry in at-reactor Independent Spent Fuel Storage Installations (ISFSIs) or wet in spent-fuel pools. One method for evaluating these alternatives is simulating the movement of SNF throughout the system components. Successfully simulating a system on the scale of a national waste management system requires - accurate baseline information (e.g., historic fuel discharges from reactors and dry storage loadings at reactor sites), - modeling all logistic interactions between components of the proposed system, and - tracking and reporting all relevant details regarding the simulation's execution. The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool designed for the express purpose of modeling the IWMS. It is based on the Performance Analysis Tool, a simulation platform that has been used to model logistics for the Department of Defense and the Federal Emergency Management Agency. Using existing, government-owned software allows NGSAM developers to modify the software at the level of source code and to develop flexible usability and modeling solutions for developing and running simulations of the IWMS. NGSAM imports data from the Oak Ridge National Laboratory (ORNL) Unified Database (e.g., historic assembly information, thermal profiles for assembly heat, at-reactor dry storage loadings) to ensure that the simulation initializes with a realistic representation of the state of commercial SNF in the United States. NGSAM is designed with the flexibility to model and track the behavior of virtually any IWMS architecture assumptions or design considerations. For example, NGSAM can model multiple interim storage facilities or repositories or none at all. Similarly, all assumptions regarding shipment schedules, operating schedules, and reactor operation assumptions can be modeled and adjusted in NGSAM with relative ease. The development of NGSAM was targeted to waste management system analysts as the primary users. With that in mind, analysts from ORNL, Sandia National Laboratories (SNL), and Argonne National Laboratory (ANL) were intimately involved in the development process. That involvement extended beyond merely defining requirements. System analysts have been involved with design decisions and software benchmarking and tests, and they have provided subject-matter expertise where necessary. The result is a high degree of interaction between professional software developers and system analysts, with a multitude of benefits to both groups. This paper explores some of the underlying methodology of the agile software development process used in the creation of NGSAM. It illustrates the capabilities of NGSAM. Some example use cases are discussed as well, with relevant results providing context for how NGSAM is used by system analysts to gain a greater understanding of alternative waste management system architectures. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be 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 provisions prevail. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

TN{sup R} MW- Multi Waste Package 'All In One Solution' For Conditioning - Transport - Storage - Up To Final Disposal - 17196

For the management of radioactive waste, either during nuclear facility operation or decommissioning, one must consider a diverse array of nuclear waste forms of various types, volumes and activities. Examples are: activated metallic waste; legacy waste; sludge; resins; and orphan waste. Customers are frequently faced with the obligation of undertaking multiple, costly waste management operations. These operations activities include handling, reconditioning or material transfers between packages, as in the case when moving material from on-site storage to transportation, and then from transportation to final disposal. To address this issue, a new - highly flexible up to reversible - cask system, the TN{sup R} MW, is being developed as an 'All in One Solution'. TN{sup R} MW has a total weight of 10 metric tons and is compliant with the 2012 International Atomic Energy Agency (IAEA) regulations. It was developed to provide flexibility and designed with the capability to be adaptable to the various needs of the nuclear industry. From IP-2 and Type A containers to Type B(U) and B(U)F casks, this cask offers different options to sufficiently address the requirements of each step of the waste stream from on-site and international transport to long-term interim storage and final disposal. Licensing and manufacturing of a number of packagings in this TN{sup R} MW family are underway. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Reviewing Spent Nuclear Fuel Dry Cask Storage Programs to Identify Challenges for an Integrated Waste Management System - 17535

Under the Nuclear Waste Policy Act (NWPA), the U.S. Department of Energy (DOE) has responsibility for removing commercial spent nuclear fuel (SNF) from nuclear power plant sites and disposing of the SNF in a deep geologic repository. However, the methods used by commercial nuclear power plants for onsite storage of their SNF define the starting conditions for DOE efforts to develop a waste management system that includes packaging, transportation, storage, and disposal of commercial SNF. The NWPA (as amended) also charges the U.S. Nuclear Waste Technical Review Board (Board) with evaluating the technical and scientific validity of DOE activities related to the management and disposal of SNF and high-level radioactive wastes (HLW). DOE's efforts related to developing an integrated program for storing, transporting, and disposing of SNF and HLW were the focus of a public meeting held by the Board on August 24, 2016, in Washington, D.C. To prepare for the public meeting, the Board's staff compiled information on the dry cask storage systems in service at nuclear power plant sites in order to identify issues that would need to be addressed by DOE in planning to transport SNF away from the sites. This paper is based on a briefing presented to the Board members by the Board's staff before the public meeting. Post-meeting findings and recommendations of the Nuclear Waste Technical Review Board are documented in the Board letter to the Department of Energy dated December 8, 2016. A note about terminology: in this paper, the phrase 'dry cask storage system' is generic and applies to a complete system used for dry storage of commercial SNF.Within these storage systems, commercial SNF is sealed inside a container called either a cask or a canister depending upon the specific type of storage system used. This distinction is discussed in more detail in the body of this paper. When referring to the containers holding SNF in particular systems, this paper uses the terms 'casks' or 'canisters,' as appropriate. When discussing the collective group of all containers, this paper uses the phrase 'casks and canisters'. Dry cask storage of SNF is a widespread practice throughout the U.S. commercial nuclear industry. As of August 1, 2016, more than 2300 storage casks and canisters were loaded at U.S. commercial nuclear power plant sites. This number is expected to increase by more than 200 casks and canisters annually, reaching a total of approximately 4500 by 2025. Multiple vendors have developed their own unique designs of dry cask storage and transportation systems. The majority of storage system designs (75 percent) feature SNF loaded in welded stainless steel canisters that are stored in ventilated concrete overpacks that allow for convection cooling. Some older dry cask storage system designs do not include a welded stainless steel canister; instead, SNF is loaded directly into a heavily shielded metal cask to minimize radiation exposure to workers and prevent atmospheric release of radionuclides. These types of systems are sealed by O-rings compressed under a bolted lid. After loading, dry cask storage systems are located on concrete pads at Independent Spent Fuel Storage Installations (ISFSI) at nuclear power plant sites and will remain there until the time when they will be transported to a consolidated interim storage facility or a deep geologic repository. Eventual transportation of canistered SNF will include removing the stainless steel canisters from the concrete overpacks, possibly repackaging the SNF, and loading the canisters into specially designed transportation casks for shipment by truck, rail, or barge. In the case of the non-canistered casks, some are approved for transportation and can be shipped after impact limiters are installed on the casks. SNF will need to be removed from casks not approved for transportation and transferred to transport-approved casks. This paper highlights five issues and challenges presented by the nuclear industry's use of dry cask storage systems that will need to be addressed by DOE in developing an integrated waste management system. One example, among the five, is that several dry-storage casks and canisters have received Nuclear Regulatory Commission certificates of compliance for storage only and do not have certificates of compliance for transportation. This will pose a challenge when the SNF must be moved to a consolidated interim storage facility or a deep geologic repository. Another example is that, when loaded with SNF, the total weight of new, larger transportation cask designs will exceed 142 metric tonnes (156 tons) which, when combined with the weight of the rail-car, will exceed the gross weight limit for unrestricted operation on some rail lines. This will also affect future transportation planning for shipments of SNF. All five issues and challenges are discussed in detail in the body of the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Implementation Guide for the Revised Branch Technical Position on Concentration Averaging and Encapsulation - 16021

As a result of the research begun in 2007 to better understand the bases of the Branch Technical Position (BTP) on 'Concentration Averaging and Encapsulation' and 10 CFR 61 'Licensing Requirements for Land Disposal of Radioactive Waste', EPRI identified three regulatory initiatives that could be pursued to enhance disposal options for the industry. These initiatives directly follow the NRC strategic assessment. The first was to revise the BTP to allow broader blending of compatible waste types. The second was to examine how 10 CFR 61.58 'Alternative requirements for waste classification and characteristics' could be used to allow for an alternative disposal criterion based on site-specific hydro-geological characteristics and end land use scenarios. And, the third is to develop a technical basis for risk informed disposal regulations in anticipation of a 10CFR 61 rulemaking. By the summer of 2015 these initiatives had been broadly addressed and resulted in the NRC issuing Revision 1 to the BTP in February of 2015. Revision 1 of the BTP (BTP 2015) is a significant re-write of the draft issued in May 2012 and a significant improvement over the 1995 BTP (BTP 1995). The NRC has altered many of the positions in the interest of clarification. The development of an Implementation Guide for applying BTP 2015 to industry-specific waste streams that is vetted and supported by a EPRI sponsored working group consisting of industry participants (including utilities, disposal sites, NRC and Agreement State representatives and industry experts) will ensure compliant and consistent application during waste generation, packaging, processing and disposal activities. The Implementation Guide will provide for a comparison to previous BTP guidance and provide clarification and common understanding of the key concepts contained within the revised BTP. These include averaging constraints, hot spots, waste streams, waste types, discrete items, blending, solidification, encapsulation and alternative approaches. EPRI plans to publish the BTP Implementation Guide in the second quarter of 2016 and it will be available to the public. (authors)

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

A Method for the Assessment of Operational Safety of DGR Using Scenario Development - 18187

This paper treats the subject of the accident analysis for the operational phase of a deep geological repository (DGR) for radioactive waste. It describes the further development of the procedures used in the accident analysis. The procedures of the scenario development for the post-closure phase of DGRs are used as example for this further development. Nine international projects regarding scenario development were analyzed to create an advanced and more systematic method for the accident analysis. (authors)

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

ENRESA's Spent Nuclear Fuel and Special Waste Database - 18115

Empresa Nacional de Residuos Radiactivos S.A. (ENRESA, by its initials in Spanish) Spent Nuclear Fuel (SNF) and Special Waste (SW) database development was initiated in a common effort between ENRESA and UNESA (the Spanish Association of the Electricity Industry) as a consequence of new regulatory requirements in the frame of radioactive waste management plans. Although these requirements were only mandatory for operators of nuclear power plants, ENRESA deemed necessary to have a centralized database which collected the data from the plant operators. The most relevant features of this database are: - The information collected covers not only the general characteristics of the SNF and SW but also the operation details needed for their characterization. - An application, named Gestion de Elementos Combustibles y Residuos Especiales (GECYRE, by its initials in Spanish) has been implemented for easing the queries and the generation of predefined reports. - An agile data transfer scheme, based on XML files, is used for an effective communication between ENRESA and the operators. The database is currently in production phase since 2014. During this phase it has demonstrated its effectiveness as an information repository to address the SNF and SW interim storage projects in Spain foreseen in the near future. (authors)

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

Neutron radiation shielding material polyethylene: consequences of gamma irradiation - 15062

High and ultra-high molecular weight polyethylenes are high performance materials, which are predestined for a wide range of applications due to characteristics like high density, low weight, good dimensional stability, high chemical resistance, and high hydrogen content. These polymers are used in two demanding areas: as implant material in medical technology (only ultra-high molecular weight polyethylenes) and as a component for neutron shielding purposes in casks for storage and transport of radioactive waste. In the medical field (joint replacements) as well as during neutron radiation shielding application, high and ultra-high molecular weight polyethylenes are exposed to gamma irradiation: in the first case requested as sterilization process and for surface-crosslinking and in the second case existing as a side effect of inserting the radioactive material in the cask. Given that polyethylene as shielding material has to withstand any type of degradation affecting safety relevant aspects to be applicable for long-term radiation shielding purposes for instance over a period of 40 years, the durability of the material is of special interest. High molecular weight polyethylene (HMW-PE; LUPOLEN 5261Z; Lyondell Basell) and ultra-high molecular weight polyethylene (UHMW-PE; GUR 4120; Ticona) were subjected to gamma radiation and afterwards thermally treated. The gamma doses used are in the range of 50 to 600 kGy and irradiation takes place at RT using a Co-60 source. The planned thermal treatment will take place at a temperature of 125 deg.C for periods of minimum 30 days and maximum of five years. With the applied conventional analytical techniques it is possible to detect structural changes of both types of polyethylene induced by gamma irradiation and certainly of thermal treatment. Through gamma irradiation melting temperature, crystallinity, and density, respectively increased. Furthermore both polyethylenes get oxidized and cross-linked. With regard to the special application as neutron radiation shielding material in casks for storage and transport of radioactive materials, the impact of irradiation lead to changes of material properties. A consolidated view indicates that the detected changes of the irradiated (U)HMW-PE are not safety relevant for long-term neutron radiation shielding purposes over a period of 40 years in Germany. (authors)

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

Classification methodology for tritiated waste requiring interim storage

Fusion machines like the ITER experimental research facility will use tritium as fuel. Therefore, most of the solid radioactive waste will result not only from activation by 14 MeV neutrons, but also from contamination by tritium. As a consequence, optimizing the treatment process for waste containing tritium (tritiated waste) is a major challenge. This paper summarizes the studies conducted in France within the framework of the French national plan for the management of radioactive materials and waste. The paper recommends a reference program for managing this waste based on its sorting, treatment and packaging by the producer. It also recommends setting up a 50-year temporary storage facility to allow for tritium decay and designing future disposal facilities using tritiated radwaste characteristics as input data. This paper first describes this waste program and then details an optimized classification methodology which takes into account tritium decay over a 50-year storage period. The paper also describes a specific application for purely tritiated waste and discusses the set-up expected to be implemented for ITER decommissioning waste (current assumption). Comparison between this optimized approach and other viable detritiation techniques will be drawn. (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Cobalt-60, Barium-133, Cesium-137, and Europium-152 migration from cementitious sources through sediment under field conditions

Safe and effective storage of radioactive waste is essential to protect human and environmental health. Due to the potential for accidental releases and the severity of the associated risks, it is imperative to further understand radionuclide transport should an accident occur. This study was the second set of measurements conducted in 2022 of an ongoing experiment that has analyzed the vadose zone migration of radionuclides from cementitious wasteforms at the Savannah River Site over the last ten years. The radionuclides introduced within the sources are prominent constituents of radioactive waste or analogs for other groups or series of radionuclides. Lysimeters were first analyzed in 2016 using a collimated high-purity germanium gamma-ray spectrometer to non-destructively measure the concentration of each radionuclide in the sediment column as a function of depth. Following these measurements, the lysimeters were redeployed for another 4 years. All radionuclides in all lysimeters were observed to transport further during the redeployment period; however, the extent of migration varied with the material used for introduction. Except for 137Cs, migration through the sediment control system increased with decreasing ionic potential (ionic charge/radius); migration order: 152Eu<137Cs<60Co<133Ba. Overall, the cementitious wasteforms were observed to decrease radionuclide migration extent relative to natural vadose zone conditions. In both cementitious wasteforms, the migration extent increased in the order 152Eu<133Ba<60Co<137Cs. However, less migration was measured when the radionuclides were incorporated into a reducing grout wasteform. The novelty of this paper is the demonstration of a technique capable of creating non-destructive measurements over decade time scales. Ultimately, this work provides insight into the long-term migration of alkali, alkali earth, divalent transition metal, and trivalent (e.g., lanthanide and actinide element) isotopes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Panel Session 129: Perspectives (US and Non-US) on the Use of Risk and Dose Assessment Tools (R9.3)

This panel focused on various dose and risk assessment tools for assessments, Deactivation and Decommissioning, remediation, and closure of sites of chemical and radioactive wastes. Representatives from the US government agencies, non-US countries and their representative regulatory bodies or authorities, as well as performance/risk assessment practitioners compared and contrasted the various guidance from regulatory agencies and authorities on how tools such as the Preliminary Remediation Goals (PRG) calculator, the Dose Compliance Concentrations (DCC) calculator, and Residual Radioactivity (RESRAD) should be used in support of analyses and decisions for environmental cleanup activities. Panelists with presentations: Cumulative Impact Evaluation: Innovative Tools for Evaluation of Groundwater Protection (Alaa Aly); Superfund Evaluation Process for Alternative Risk and Dose Models (with Focus on EPA and DOE Tools) (Stuart Walker); NRC Staff Perspective on Risk-Informed Approach and Reasonable Safety Assurance in D and D and LLW (Rateb (Boby) Abu Eid)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Video Camera Improvements to Support Hanford Tank Farms - 17141

Hanford Tank Farms consist of 177 underground storage tanks that contain 56 million gallons of radioactive waste. The current mission is to safely manage this waste in preparation to feed the Waste Treatment Plant. Fulfilling this mission includes inspecting each tank regularly to ensure tank integrity and support ongoing operations such as transferring waste from older single-shell tanks to the newer double-shell tanks. Video cameras derived from commercially available equipment have provided the ability to remotely see inside the tanks, reducing some of the challenges of fulfilling this mission. However, there are many challenges and limitations with the currently used video systems. Among some of the challenges are limited tank interior entry points for camera deployment, high radiation dose rates, limited equipment longevity in the environmental conditions, need for high resolution image quality, low light conditions, and in-tank obstructions. Washington River Protection Solutions (WRPS) has identified and tested new video systems that will effectively manage these challenges and increase capability, efficiency, and improve safety. The new camera systems are planned for deployment to support the next Single Shell Tank Retrieval Project in the A/AX tank Farm complex. (authors)

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

Approaches to consider covers and liners in a low-level waste disposal facility performance assessment

On-site disposal cells are in use and being considered at several USDOE sites as the final disposition for large amounts of waste associated with cleanup of contaminated areas and facilities. These disposal cells are typically regulated by States and/or the USEPA in addition to having to comply with requirements in DOE Order 435.1, Radioactive Waste Management. The USDOE-EM Office of Site Restoration formed a working group to foster improved communication and sharing of information for personnel associated with these Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) disposal cells and work towards more consistent assumptions, as appropriate, for technical and policy considerations related to performance and risk assessments in support of a Record of Decision and Disposal Authorization Statement. One task completed by the working group addressed approaches for considering the performance of covers and liners/leachate collection systems in the context of a performance assessment (PA). A document has been prepared which provides recommendations for a general approach to address covers and liners/leachate collection systems in a PA and how to integrate assessments with defense-in-depth considerations such as design, operations and waste acceptance criteria to address uncertainties. Specific information and references are provided for details needed to address the evolution of individual components of cover and liner/leachate collection systems. This information is then synthesized into recommendations for best practices for cover and liner system design and examples of approaches to address the performance of covers and liners as part of a performance assessment of the disposal system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗