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DOE OSTI2019

International Low Level Waste Management Practices and Technical Bases for Very Low Level Waste Classification - 19415

Different countries around the globe use varying schemes for the classification of radioactive waste. Many countries use adaptations of the International Atomic Energy Agency (IAEA) waste classification scheme. The United States uses the Class A, B, and C waste classification scheme that is loosely based on the IAEA concepts. Classification may be based on concentration limits or dose rate limits or a combination of concentration and dose rate limits. In some countries the existence of very low-level waste categories allows nuclear power plant waste to be disposed of in hazardous waste or shallow land burial sites separate from low and intermediate level waste disposal facilities. In the United States, the very low-level waste (VLLW) classification does not exist. Nuclear power plants use alternative disposal procedures under 10CFR20.2002 and other licensed processes in agreement states in order to dispose of very low activity waste in hazardous waste disposal sites. [1] Disposal of VLLW in hazardous waste or separate VLLW facilities is generally more cost effective than disposing of waste in radioactive waste sites and is protective of the health and safety of the public. This can become an important factor when large volumes of very low activity waste are generated during decommissioning of nuclear power plants. The Electric Power Research Institute (EPRI) Radiation Safety (RS) Program has developed technical bases for the establishment of a very low-level waste category. EPRI studied how VLLW waste is defined, managed, and disposed of in various countries. This included the study of disposal site design, characteristics, and how these sites protect the health and safety of the environment and the public from low levels of radioactivity. EPRI conducted a cost benefit analysis to understand the economic impact of disposing VLLW in hazardous waste disposal sites versus low-level waste disposal sites. Finally, EPRI calculated potential radionuclide limits that may be used to define very low-level waste. The purpose of this work was to provide an example of how one might approach defining limits for VLLW. The same or similar approaches could be applied to derive radionuclide limits for different dose limit objectives. This paper presents the results of research that explored the waste management and disposal practices at six countries around the world with additional focus and details on the implementation of the very low-level waste category. (authors)

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

EPRI Research Results: International Nuclear Power Plant Waste Classification - 17444

Different countries around the globe use varying schemes for the classification of radioactive waste. Many countries develop country specific waste classification schemes based on the International Atomic Energy Agency (IAEA) waste classification framework. The United States of America use the Class A, B, and C waste classification. Additionally, some countries use concentration limits to classify wastes while others use dose rate limits or a combination of concentration and dose rate limits. General comparisons have been made between the IAEA and the United States classification schemes, but a clear and well researched comparison of the international schemes was not evident in the available research data sets. In 2013 Electric Power Research Institute (EPRI) initiated the Comparison of Global Low and Intermediate Level Waste Management Methods project. The objective of the project was to understand how six countries managed their low and intermediate level wastes from generation to disposal. A major task of this project was to conduct a comparison of the international waste classification schemes. EPRI developed a set of reference waste volumes that are representative of nuclear power plant low-level radioactive waste in form and radioactivity concentrations (e.g. dry active waste, high activity resin, low activity resin, high activity filters and low activity filters). These waste volumes were then classified using the waste classification systems in place in several countries around the world. In 2016, EPRI presented a paper providing a status update on this research. This 2017 paper provides the final results of that research, show where comparisons and correlations can be made, and provide a discussion on how this information might inform international radioactive research and development and benchmarking of radioactive waste management strategies, operational experiences, best practices, and technology implementation. (authors)

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

Panel Session 36: Hot Topics in US Commercial Low-Level Waste Management

This panel focused on emerging issues in commercial LLW management in the US from the perspective of active members of the Low-Level Radioactive Waste Forum, Inc. State, compact, federal and industry officials shared their views on a variety of timely and significant topics related to low-level radioactive waste management, disposal and related issues. Session no. 36 Co-Chairs, Patricia Holahan and Daniel Shrum, introduced the panelists and provided a summary of the purpose of the session. This Panel Session focused on several current and emerging issues impacting the commercial Low-Level Radioactive Waste (LLRW) management in the US from the perspective of active members of the Low-Level Waste Forum, Inc., States, Compact, as well as Federal and nuclear industries. The panelists provided different perspectives on number of key issues such as: LLW disposal facility status and updates (e.g.; in terms of disposal capacity, plans, and waste acceptance conditions). The panelists also addressed waste disposal rates and potential impacts of ongoing regulatory developments by the NRC and the concerned States. Other timely topics pertaining to the revision of 10CFR Part 61, the Greater-than-Class C (GTCC) and Transuranic (TRU) waste, and Very Low-Level Waste (VLLW) disposal issues were also addressed at this Session. Panelists with presentations: Hot Topics in US Commercial LLW Management (Joe Weismann); Texas LLRW Compact (David Carlson); Texas Low-Level Radioactive Waste Disposal Compact Commission (John Salsman); Barnwell LLRW Disposal Facility (Benjamin Smith, Bill Carver)

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

Lessons Learned from More than Three Decades of Decommissioning and RAW Management in Slovakia - 20074

The purpose of this paper is to describe utilization of experience of JAVYS Company acquired while acting as the Slovak national decommissioning and radioactive waste management entity. JAVYS manages both the decommissioning of A1 NPP (shut down after series of accidents) and V1 NPP (shut down after standard operation) and related radioactive waste management. Besides JAVYS is responsible for management of institutional radioactive waste in Slovakia. The paper is focusing on selected successfully completed projects and related lessons learned from development of special procedures and implementation of activities within decommissioning and management of generated waste streams from cradle to grave. Recently, experience and lessons learned from all aforementioned activities are used internationally to provide assistance to other countries in order to overcome difficulties and to successfully implement various projects dealing with decommissioning and radioactive waste management. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Development of an ERT‐Based Framework for Bentonite Buffers Monitoring From Laboratory Tests: 1. Characterizing Thermal–Hydrological–Mechanical Processes

Abstract Bentonite clay is widely used in engineered barrier systems for the permanent disposal of high‐level radioactive waste due to its low permeability, high swelling capacity, and thermal stability. However, the complex thermal‐hydrological‐mechanical (THM) processes induced by heating from decaying radioactive waste and hydration from surrounding rock can lead to heterogeneous changes that are difficult to measure and predict. This study develops an Electrical Resistivity Tomography (ERT)‐based framework for monitoring THM processes, progressing from sample‐scale to bench‐scale tests, to inform field‐scale applications. Sample‐scale tests analyzed small bentonite samples under controlled variations in water content, temperature, and porosity to establish fundamental resistivity relationships. Bench‐scale tests involved larger bentonite columns subjected to heating (up to 200°C) and hydration under controlled pressure, simulating repository conditions. ERT measurements, complemented by X‐ray CT imaging, temperature monitoring, and tracing sensors, revealed coupled THM processes, such as hydration‐induced compression, swelling, and thermal gradients, leading to complex resistivity patterns. The results demonstrate the potential of ERT for capturing THM‐induced resistivity changes, though challenges remain in upscaling and quantitative analysis. This study evaluates laboratory test capabilities and proposes future improvements for understanding THM‐induced resistivity responses. A conceptual framework for ERT implementation in field‐scale monitoring is presented, synthesizing findings from both scales and exploring how ERT data can inform long‐term modeling and reduce prediction uncertainties. Overall, this ERT‐based framework offers a robust method for monitoring bentonite buffers, aiding in early issue detection and supporting the safe long‐term disposal of radioactive waste in geological repositories, while highlighting the need for future development. Plain Language Summary Bentonite clay is crucial in engineered barrier systems (EBS) for containing high‐level radioactive waste due to its ability to absorb water, swell, seal and remain stable under high temperatures. When bentonite absorbs water and heats up from radioactive decay, it experiences complex changes in its physical and mechanical properties. Understanding these changes is important for ensuring the long‐term safety and effectiveness of EBS. This study used Electrical Resistivity Tomography (ERT), a non‐invasive method that measures electrical conductivity to monitor these changes during laboratory experiments. The ERT data revealed significant variations in resistivity corresponding to changes in water content, temperature, and density, providing detailed spatial and temporal insights into the behavior of bentonite. These findings enhance our ability to predict the long‐term performance of bentonite barriers, ensuring the safe containment of radioactive waste. By improving our understanding of bentonite's behavior, this research supports the development of more reliable and effective barrier systems for radioactive waste disposal, protecting the environment and public health. Key Points ERT monitoring was employed to capture resistivity changes in bentonite during controlled heating and hydration experiments, providing insights into THM processes ERT data reveal significant resistivity changes correlated with water content, temperature, and mechanical effects, enhancing the understanding of THM dynamics in bentonite This study explores the potential of the framework for application in field‐scale EBS monitoring, emphasizing the need for integrating additional geophysical methods for comprehensive subsurface imaging

Chen, Hang↗
DOE OSTI2020

Integrating Commercial Ready Autonomous Robotic Platforms to Enhance Worker Safety and Efficiency - 20053

This paper will explore the challenges and triumphs of implementing advanced robotic systems and the supporting technology for the waste management industry. Deployment-ready solutions for autonomous robotic platforms are new to the radioactive waste management industry and have been slow to implement across the complex. Specifically, it will explore Hanford Tank Operations Contractor's implementation of ground-based autonomous robotic platforms to enhance workforce safety and efficiency. The first objective of the project is to demonstrate safe operation of autonomous instrumented robots, followed by a scaling up of capabilities for the platform to meet any additional requests of internal stakeholders. Autonomous robotics platforms offer a fantastic range of capabilities that will aid in decreasing long-term liability of radioactive waste management industry by enhancing clean-up efforts and worker safety. Over the past five years, deployment-ready solutions for autonomous robotic platforms have come a long way. Autonomous robotic platforms are increasingly available to purchase from many vendors, often for prices under $100,000. Additionally, the vendors who produce these products are efficient in accommodating customized solutions to unique challenges, such as those often found in the radioactive waste management industry. However, the newness of this technology to the radioactive waste management industry and has resulted in a slow implementation across the complex. This paper will explore the challenges and triumphs of implementing advanced robotic systems and supporting technology for the waste management industry. Specifically, it will explore Hanford Tank Operations Contractor's implementation of ground based autonomous robotic platforms to enhance workforce safety and efficiency. While the company has utilized robots in many parts of the tank farms mission, this was first attempted in the deployment of an autonomous platform that required no operator with only minor maintenance and created challenges not yet encountered. These challenges included stakeholder buy-in, operational safety, procurement uncertainties, adaption of 'off the shelf products, overcoming the 'Fukushima complexity' perception, security concerns, and the initial criteria down select. Hanford Tank Operations Contractor's management identified the need for a workforce protecting low cost robotics solution in the fall of 2017. The project scope generated was part of the response to the Tank Vapor Assessment Team recommendation to enhance the safety of the workforce by utilizing new technology. Since the projects inception, it has been incorporated into the Tank Stewardship Plan. This project was unique to Hanford Tank Operations Contractor, as the company does not frequently incorporate advanced technology for field deployment quickly. The project was also progressive to the Hanford Tank Operations Contractor as it functions as a partnership project with a local university. The first objective of the project is to demonstrate safe operations of autonomous instrumented robotics, monitoring and collecting data, and the wireless transmission of data to a central computing system. Following the completion of this first objective, the project will scale up capabilities of the platform to meet the requests of additional internal stakeholders throughout the tank farms - including completion of specific in-field tasks. (authors)

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

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 18327

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to cleanup radioactive waste at the Hanford Site, located in southeastern Washington State. The overall WTP Project objective is to design, build, and commission the facilities and systems that will treat and immobilize approximately 56 million gallons of radioactive waste stored in Hanford's 177 underground storage tanks. The combination of technologies in the WTP makes it a first-of-a-kind facility at a scale larger than ever, built for radioactive waste processing. DOE is proceeding with its approach to complete the WTP Project in phases, which includes the following principal actions: - Completing construction of the Low-Activity Waste (LAW) Facility, Balance of Facilities (BOF), and Analytical Laboratory (LAB) (collectively referred to as LBL); - Implementing a direct-feed low-activity waste (DFLAW) process to treat low-activity waste independent of the Pretreatment (PT) Facility; - Resolving the technical issues associated with primarily the PT Facility; - Completing the design and construction of the High-Level Waste (HLW) and PT facilities to enable integrated WTP operations. In March 2016, a federal judge amended the 2010 Consent Decree 1 completion milestones for the Hanford WTP, which generally aligns with DOE's plans for completing the WTP facilities in sequence: - Complete LAW Facility hot commissioning: December 31, 2023; - Complete HLW Facility hot commissioning: December 31, 2033; - Complete PT Facility hot commissioning: December 31, 2033; - Achieve initial plant operations for the WTP: December 31, 2036. Having established milestones for completing the WTP facilities has enabled DOE to formally implement its plans for completing the WTP facilities in sequence, starting with the LAW Facility. In December 2016, DOE executed a contract modification and approved a change to the project's performance baseline to complete the LBL facilities and incorporate modifications needed to allow direct feed of low-activity waste to the LAW Facility. Significant progress has been made toward completing the design, construction, and planning for commissioning of the facilities needed to enable the DFLAW operations to commence as early as January 2022. In 2012, DOE partially suspended production work on the WTP's PT Facility and to a lesser degree the HLW Facility because of unresolved technical issues associated with those facilities. In September 2017, DOE acknowledged the WTP contractor had met all the criteria for full resumption of engineering, procurement, and construction (EPC) activities for the HLW Facility. By December 2017, DOE had completed the actions needed to resolve the six of the eight primary technical issues associated with the PT Facility, with the remaining two issues expected to be resolved in the first quarter of 2018. Although significant progress has been made to prepare for resuming production activities for the HLW and PT facilities, DOE has made the decision to constrain EPC work on those facilities in the near term based on DOE's higher priority of completing the WTP facilities needed to begin DFLAW operations. (authors)

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

Development of the Licensing Procedure and Regulatory Framework for the Spent Fuel Storage facility in Korea - 16423

The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the future national policy for spent fuel management of Korea has been established, taking into consideration the national/international trends on policy and technology development. Regardless of the recommendation report of PECOS, Spent fuel storage casks are necessary in Korea. Therefore, various types of spent fuel storage casks are now being developed by Korea vendors. For example, Concrete storage casks and dual purpose metal casks are now under development by the KORAD (Korea Radioactive waste agency) for future use in the interim storage facility. Unfortunately, there is no independent licensing procedure for spent fuel storage casks in Nuclear Safety Act (NSA) because is the casks are considered an integral part of the interim storage facility safety equipment. Furthermore, the definition of radioactive waste in the NSA was changed recently. Now, spent fuel is separated from the definition of 'radioactive waste'. Spent fuel can only become radioactive waste by the decision through the meeting of the relative ministers of Korea government. For that reason, the licensing procedure for interim spent fuel storage facilities was separated from the current 'disposal facility, etc.' in accordance with the article 63 of NSA. The aim of this study is to develop the licensing procedure for the storage cask, and also to develop a revised draft of interim storage facility procedure within the NSA. Once this independent licensing procedure on the spent fuel storage cask is introduced to the NSA, it is expected that developers or operators can develop and commercialize the storage cask in Korea. In this study, an independent licensing procedure for the spent fuel storage facility was developed including the storage cask licensing procedure for the NSA. The procedure was composed of the permit for construction and operation of interim storage facility, criteria for permit, inspection criteria, periodic safety review, and decommissioning criteria, etc. The detailed contents of these requirements were also developed to introduce the regulatory framework for the interim storage facility. Particularly, periodic safety review and decommissioning criteria were developed in accordance with the recommendation of IRRS (International Regulatory Review Service) of IAEA. The developed licensing procedure for spent fuel storage facility will be introduced in NSA in response to the revised definition of NSA and the recommendation report of PECOS. (authors)

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

Waste is our Product. Developing, Implementing and Optimising Waste Management at Dounreay - 19039

The UK's Fast Breeder Reactor Research programme was centred at Dounreay, in the very north of Scotland, from 1955. Since 1994, the site has been focused on decommissioning, and represents one of the most complex, technically challenging and hazardous nuclear clean-up projects in Europe. Since 2012, Dounreay Site Restoration Ltd (DSRL) has been contracted by the NDA as the Site License Company to deliver the defined scope of work, under a Target Cost contract, to decommission the site to achieve the Interim End State by 2030-2033, whilst managing the transition of its people. All parties, including DSRL's Parent Body Organisation Cavendish Dounreay Partnership Ltd (a consortium of Cavendish Nuclear, AECOM, and Jacobs), recognise that there are significant technical and regulatory challenges to achieve this, as well as safely managing the transition for our workforce on what is effectively a 'closure' contract. The product of decommissioning is waste; over the course of the programme, a full range of higher activity and lower level radioactive wastes will be generated, and must be safely managed. This paper will address: - The Policy issues relating to radioactive waste management at Dounreay, including particular issues relating to implementation of Scottish policy, including how Waste Management is Optimised. - An overview of the decommissioning programme and the delivery of an Interim End State; - The type, range and nature of radioactive wastes requiring safe management at Dounreay; - Higher Activity Wastes (HAW): - Setting higher activity waste strategy; defining products and interim storage requirements; - Technical, regulatory and stakeholder issues for HAW strategy; - Lower Activity Wastes: - Setting Low Level Waste (LLW) strategy; defining a disposal product and development and status for Dounreay's LLW Disposal Vaults; - Implementing LLW strategy; - Technical, regulatory and stakeholder issues for LLW strategy; - Optimising Waste Management -Optimising waste from generation (from decommissioning workfaces) through to disposal -why, and how; - The Waste Informed Decommissioning model as a methodology and approach; - Impact and importance of Waste Management on a target cost contract; - Establishment of a Waste Optimisation Programme to drive ownership, delivery and culture change; - Management responsibilities and informing organisational design; - Sharing and learning from Best Practice; - The Challenge - how to ensure workforce and teams are engaged in delivering optimisation requirements; - Conclusions. (authors)

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

Hanford Site Composite Analysis Technical Approach Description: Vadose Zone

The U.S. Department of Energy (DOE) in DOE O 435.1 Chg. 1, Radioactive Waste Management, and DOE M 435.1 Chg 1, Radioactive Waste Management Manual, requires the preparation and maintenance of a composite analysis (CA). The primary purpose of the CA is to provide a reasonable expectation that the primary public dose limit is not likely to be exceeded by multiple source terms that may significantly interact with plumes originating at a low-level waste disposal facility. The CA is used to facilitate planning and land use decisions that help assure disposal facility authorization will not result in long-term compliance problems; or, to determine management alternatives, corrective actions, or assessment needs, if potential problems are identified.

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

A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor

The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures. Both types of HTGR have the tristructural isotropic (TRISO) fuel kernel at the heart of the fuel design. For the prismatic block-type HTGR, the TRISO particles are overcoated with a resinated graphitic matrix and pressed into fuel compacts, which are then heat treated and placed in the fuel channels of the prismatic-block-shaped fuel assemblies. For the pebble-bed-type HTGR, the TRISO particles are dispersed in a graphitic-matrix sphere, which is the basic unit for the reactor core. Despite having very different fuel designs, both types of HTGR are graphite-moderated, gas-cooled, thermal reactors using many of the same materials. As a result, both prismatic-block-type and pebble-bed-type HTGRs have similar radioactive waste streams, all of which require safe and secure storage and eventual disposition. Modern HTGR designs are based on a long and rich operating history of several different graphite-moderated, gas-cooled, thermal reactors. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the spent nuclear fuel (SNF) and radioactive wastes associated with operating these reactors. This white paper will, (1) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste; and (2) identify the key radioactive waste streams from HTGRs and their characteristics. Idaho National Laboratory (INL) has significant experience in the management of SNF from HTGR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new HTGR-type small modular reactor, or microreactor intended for deployment at the INL site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Deep Borehole Disposal Concept: Development of Universal Canister Concept of Operations

This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

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

Solidification of Low and Intermediate Level Waste - A Comparison between Different Methods - 19584

The treatment of radioactive materials results in products that have to be suitable for a period of interim storage as well as for disposal in a repository. Considering available options for treatment of radioactive waste, future disposal also has to be taken into account. Up to now, only cementation and encapsulation in geopolymer (only to limited extent) are relevant treatment methods for the solidification of German waste for disposal in the repository Grube Konrad. However, a synopsis of other solidification technologies is considered to be helpful in order to be prepared for future developments in the fields of radioactive waste treatment. Treatment methods addressed in this paper are: - Cementation; - Encapsulation into geopolymer matrix; - Vitrification; - SYNROC; - Geomelting. The criteria for evaluation of the treatment methods are as follows: - Requirements concerning waste properties; - Safety of treatment facilities; - Fulfillment of waste acceptance criteria for repository Grube Konrad; - Fulfillment of requirements for near surface disposal; - Suitability for long-term interim storage. (authors)

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

Special Analysis: 2016-003 Upgrade of Area G PA=CA Model to Updated Versions of GoldSim Software and to LANL Analysts

The Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several onsite and offsite exposure scenarios. The assessments are based on existing site and disposal facility data, and assumptions about future rates and methods of waste disposal.

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

Geochemical Data Package for Performance Assessment Calculations Related to the Savannah River Site

The Savannah River Site (SRS) disposes of low-level radioactive waste (LLW) and stabilizes high-level radioactive waste (HLW) tanks in the subsurface environment. Calculations used to establish the radiological limits of these facilities are referred to as Performance Assessments (PA), Special Analyses (SA), and Composite Analyses (CA). The objective of this document is to revise existing geochemical input values used for these calculations. This work builds on earlier compilations of geochemical data (2007, 2010), referred to a geochemical data packages. This work is being conducted as part of the on-going maintenance program of the SRS PA programs that periodically updates calculations and data packages when new information becomes available. Because application of values without full understanding of their original purpose may lead to misuse, this document also provides the geochemical conceptual model, the approach used for selecting the values, the justification for selecting data, and the assumptions made to assure that the conceptual and numerical geochemical models are reasonably conservative (i.e., bias the recommended input values to reflect conditions that will tend to predict the maximum risk to the hypothetical recipient). This document provides 1088 input parameters for geochemical parameters describing transport processes for 64 elements (>740 radioisotopes) potentially occurring within eight subsurface disposal or tank closure areas: Slit Trenches (ST), Engineered Trenches (ET), Low Activity Waste Vault (LAWV), Intermediate Level (ILV) Vaults, Naval Reactor Component Disposal Areas (NRCDA), Components-in-Grout (CIG) Trenches, Saltstone Facility, and Closed Liquid Waste Tanks. The geochemical parameters described here are the distribution coefficient, Kd value, apparent solubility concentration, k s value, and the cementitious leachate impact factor.

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

Bubble Formation and Lattice Parameter Changes Resulting from He Irradiation of Defect-Fluorite Gd2Zr2O7

Pyrochlores have long been considered as potential candidates for advanced ceramic waste-forms for the immobilization of radioactive waste nuclides. This work provides evidence that Gd2Zr2O7, often considered the most radiation tolerant pyrochlore, could be susceptible to radiation damage in the form of bubble nucleation at the highest He doses expected over geological time. Ion irradiations were utilized to experimentally simulate the radiation damage and He accumulation produced by ..alpha..-decay. Samples were pre-damaged using 7 MeV Au3+ to induce the pyrochlore to defect-fluorite phase transformation, which would occur due to ..alpha..-recoil damage within several hundred years of storage in a Gd2Zr2O7 waste-form. These samples were then implanted to various He concentrations in order to study the long-term effects of He accumulation. Helium bubbles 1-3 nm in diameter were observed in TEM at a concentration of 4.6 at.% He. Some bubbles remained isolated, while others formed chains 10-30 nm in length parallel to the surface. GIXRD measurements showed lattice swelling after irradiating pristine Gd2Zr2O7 with 7 MeV Au3+ to a fluence of 2.2 x 1015 Au/cm2. An increase in lattice swelling was also measured after 2.2 x 1015 Au/cm2 + 2 x 1015 He/cm2 and 2.2 x 1015 Au/cm2 + 2 x 1016 He/cm2. A decrease in lattice swelling was measured after irradiation with 2.2 x 1015 Au/cm2 + 2 x 1017 He/cm2, the fluence where bubbles and bubble chains were observed in TEM. Bubble chains are thought to form in order to reduce lattice strain normal to the surface, which is produced by the Au and He irradiation damage.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A Preliminary Radiological Risk Assessment Model for Disposition of Remote-Handled Transuranic Wastes at Los Alamos National Laboratory Area G - 20116

The U.S. Department of Energy (DOE) operates a low-level radioactive waste (LLW) disposal site at Material Disposal Area G, in Los Alamos, New Mexico, USA. Area G has been the primary LLW disposal site for Los Alamos National Laboratory (LANL) since the 1960's. In addition to LLW, Area G is host to a variety of other wastes, the disposition of which must be determined before closure of the site. A probabilistic Radiological Risk Assessment (RRA) for Area G is used in order to support decision making regarding some wastes that are not addressed in the extant Area G Performance Assessment (PA) and Composite Analysis (CA). Between 1979 and 1987, 33 special shafts were augered into the Bandelier Tuff at Area G. This volcanic tuff is present across Pajarito Plateau on the eastern slopes of the Jemez Mountains, and varies widely in its consistency, from weakly indurated non-welded layers to welded layers that uphold the mesa cliffs of the Plateau. These mesas are home to LANL, Area G, and the townsites of Los Alamos and White Rock, with residences about 1400 m from Area G. The 33 Shafts were lined with steel casing, and contain remote-handled (RH) transuranic wastes (TRU) resulting from experiments and analysis performed in special glove boxes at the Chemistry and Metallurgy Research (CMR) facility at LANL. Some of these wastes originated as used nuclear fuel. The purpose of the Area G RRA is to evaluate the potential future risk to humans and the environment from the RH TRU in the 33 Shafts in the context of the risk associated with the surrounding wastes at Area G. The analysis is responsive to expectations outlined in DOE Order 458.1, Radiation Protection of the Public and the Environment, and is informed by the Manual and Guidance accompanying DOE O 435.1, Radioactive Waste Management. Because the waste meets the definition of TRU, the regulatory context necessarily takes into consideration the regulation governing the disposal of TRU from the U.S. Environmental Protection Agency (EPA): 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. Given the broader regulatory context for the RRA, the analysis is subject to different assumptions from those made in the existing DOE O 435.1 PA and CA, such as allowing for future occupation of the site. The analysis begins with a comprehensive evaluation of features, events, processes, and exposure scenarios (FEPS) for Area G and the wastes it contains. These FEPSs are screened to eliminate from further consideration those of extremely low probability and/or consequence, and a conceptual site model (CSM) is subsequently developed. The scope and structure of the Area G RRA Model is informed by this CSM, and the Area G RRA Model is developed using the GoldSim systems analysis modeling platform. This paper presents the initial version of a defensible, transparent, and reasonably realistic model, which is based on the state of knowledge of the wastes, the site, and the FEPSs that govern contaminant transport from wastes into the environment and subsequent exposures to humans and other biota. Probabilistic model input distributions represent uncertainties inherent in the real and modeled systems. The results of the Area G RRA Model inform decisions regarding the disposition of the RH TRU in the 33 Shafts. (authors)

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

Cliff Retreat Characterization at Technical Area 54, Los Alamos National Laboratory, Los Alamos, NM

Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of Laboratory functions; environmental restoration and decontamination and decommissioned waste is generated as contaminated sites and facilities undergo cleanup or remediation. The majority of this waste is low-level radioactive waste and is disposed of at the Technical Area (TA) 54, Area G disposal facility. Area G includes Material Disposal Area (MDA) G, the portion of the facility that currently receives waste, and the Zone 4 expansion area located immediately west of MDAG. The Laboratory no longer plans to use the Zone 4 expansion area.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗