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DOE OSTI2024

Synthesis of results for Brine Availability Test in Salt (BATS) DECOVALEX-2023 Task E

In the 2023 phase of the international collaborative DECOVALEX modeling project, Task E focused on understanding thermal, hydrological, and mechanical (THM) processes related to predicting brine migration in the excavation damaged zone around a heated excavation in salt. Salt is attractive as a disposal medium for radioactive waste because it is self-healing and is essentially impermeable and non-porous in the far field. Investigation of the short-term, near-field behavior is important for radioactive waste disposal because this early period strongly controls the amount of inflowing brine. Brine leads to corrosion of waste forms and waste packages, and possible dissolution of radionuclides with brine transport being a potential transport vector to the accessible environment.The Task was divided into steps. Step 0 included matching unheated brine inflow data from boreholes at the Waste Isolation Pilot Plant (WIPP) and matching temperature observations during a Brine Availability Test in Salt (BATS) heater test. Step 1 included validation of models against a thermo-poroelastic analytical solution, and two-phase flow around an excavation. Finally, Step 2 required all the individual components covered in steps 0 and 1 to come together to match observed brine inflow behavior during the same BATS heater test.There were a range of approaches from the teams, from mechanistic to prescriptive. Given the uncertainties in the problem, some teams used one- or two-dimensional models of the processes, while other teams included more geometrical complexity in three-dimensional models. Task E was a learning experience for the teams involved, and feedback from the modeling teams has led to changes in follow-on BATS experiments at WIPP. The primary Task E lessons learned were the impact of hydrologic initialization methods (wetting up vs. drying down), the difference between confined and unconfined thermal expansion, and the large changes in permeability associated with heating and cooling.

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DOE OSTI2017

Special Radwaste Management Prospects

The initial registration brought to light the characteristics and disposition of wastes that should be disposed of on-site (special radioactive wastes). The criteria for validating the classification of radwastes as special are presented. The established criteria for the origin of wastes and the location of an object and its sanitaryprotective zone do not permit classifying the radwastes of certain large objects as special even though their removal is unjustified both in terms of dose loads and the risk of potential radiation as well as in terms of financial cost. Criteria are proposed for storage sites for special radwastes for inclusion in the developed regulatory documents, regulating their safe operation and transfer to accommodation and storage sites.

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DOE OSTI2019

Small Modular Reactors in Canada: Planning for Waste from the Outset - 19254

Canada is exploring the potential for on- and off-grid applications for small modular reactors (SMRs). SMRs are at an early stage of development and a road-mapping process was launched in Canada in February 2018 to better understand the views on priorities and challenges related to the possible development and deployment of SMRs. In addition to engaging partners, stakeholders and Indigenous representatives through workshops, a number of working groups were formed to advise the Road-map Secretariat on topics such as understanding the value proposition of different SMR technologies, matters associated with regulatory readiness, and issues related to waste management and transportation. The Waste Working Group (Waste WG) brought together experts from current nuclear waste generators, the organization responsible for managing Canada's nuclear fuel waste, government, and the regulator to identify waste disposal and storage considerations for Canadian SMR applications, including an analysis of waste streams, waste transportation, short and long-term disposal and storage requirements, and decommissioning considerations. Radioactive waste has been safely managed within a regulatory framework in Canada for decades. Technologies for waste storage and transport exist and also have been in use for decades, and therefore provide a technical basis to explore any specific differences that would apply to waste streams resulting from the construction, operation and decommissioning of SMRs. The Waste WG concluded that a disposal path for SMR used fuel exists through the Nuclear Waste Management Organisation's (NWMO) Adaptive Phased Management (APM) approach, but that R and D work is needed to demonstrate that SMR waste can meet the acceptance criteria. The Waste WG noted that although projects are progressing to realize disposal facilities for the two largest producers of low and intermediate level waste, small producers including potential SMR operators currently have no defined path to disposal for low and intermediate level waste. The Waste WG also highlighted that there could be potential logistical challenges with the transportation of irradiated reactor cores, in particular for remote locations, which will need to be explored. The Waste WG recognized that some uncertainties remain, and early and open communication and coordination between the SMR developers and the NWMO, and potential waste receivers will be key to ensuring a robust framework for SMR waste management from the outset in Canada. This paper explores the results of the analysis and the key findings, high-level recommendations and actions of the Waste WG. It presents the results of the Waste WG's analysis of waste categories against Canadian SMR technologies that best align to end-user requirements; identification of key gaps in research and development for waste categories; and identification of potential work that could address the key gaps. The Waste WG undertook its work between February and August 2018. (authors)

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DOE OSTI2020

Versatile Remediation Module Final Operational Testing Results

Oak Ridge National Laboratory (ORNL) successfully demonstrated the Versatile Remediation Module (VRM), a prototype module designed and built by ORNL for performing the on-site remote repair of welded stainless-steel storage containers for spent nuclear fuel and high-level radioactive waste. This report describes the VRM prototype and its design features, components, qualification tests, operational procedures, and demonstration results to support continued long-term storage or off-site transportation of spent nuclear fuel and high-level radioactive waste currently stored in storage containers. A remote (i.e., 100 ft away from the simulated radiative environment) demonstration of the VRM was successfully performed on a full-scale mock-up welded stainless-steel canister. The VRM is designed with features to accommodate remediation techniques beyond those currently selected and described in this report. Therefore, many of the VRM’s features could benefit other remote nuclear or nonnuclear applications. The VRM is envisioned to serve as a development center to facilitate and enhance the further development of new remediation technologies.

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DOE OSTI2017

SRR's 'Next Step' in the Lean Journey -Taking Lean to the Workforce - 17247

The Savannah River Site, a U.S. Department of Energy (DOE) facility, contracted Savannah River Remediation (SRR) to manage its liquid waste project, which is a unique, high hazard nuclear waste treatment and disposal facility consisting of 51, one million-gallon underground waste tanks, a solvent extraction processing unit, a vitrification facility for the highest radioactive waste, and a grout solidification and disposal facility for a decontaminated salt solution. Our Project mission is to treat the nuclear waste and close the associated storage and treatment facilities, of which, 8 tanks have been closed to date. At SRR, the Lean business management system has been adopted to address site inefficiencies, aging infrastructure concerns and budget constraints. The Lean philosophy took root decades ago in the manufacturing sector and started with the work of W. Edwards Deming and his contemporaries. In more recent years the process has matured in the Toyota Production System and migrated into fields as diverse as services industries, health-care, and now the DOE complex. The first step of the SRR Lean Journey was to formally identify areas of improvement and initiate Value Stream Analysis (VSAs) and Rapid Improvement Events (RIEs). SRR has achieved improvements in procedure implementation, work planning and control, project schedule compression, and engineering requirements. SRR is now phasing in the next phase of Lean through Managing for Daily Improvement (MDI), Visual Management and 5S (sort, set in order, shine, standardize, and sustain). MDI is a fundamental Lean principle in that we use our most valuable asset, our employees, to drive continuous improvement versus a smaller population through formal events. Part of MDI includes visual management where the workforce can go to a 'dashboard' to see and understand the company mission and status. 5S is a process where a work group improves performance in an area by addressing physical organization, efficiency, uniformity, standardization, and mechanisms to sustain improvements. 5S stands for: Sort, Set in Order, Shine, Standardize, and Sustain. This paper will focus on the 'Next Step' in SRR's Lean journey by address how SRR is implementing MDI and 5S to improve CONOPs, Project efficiencies, and worker engagement in day to day operations. (authors)

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DOE OSTI2015

Advanced waste form and Melter development for treatment of troublesome high-level wastes

A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe 2 O 3 (also with high Al 2 O 3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both the as poured state and after being slowly cooled according to the canister centerline cooling (CCC) profile. Glass formulation development was also completed on other Hanford tank wastes that were identified to further challenge waste loading due to the presence of appreciable quantities (>750 g) of plutonium in the waste tanks. In addition to containing appreciable Pu quantities, the C-102 waste tank and the 244-TX waste tank contain high concentrations of aluminum and iron, respectively that will further challenge vitrification processing. Glass formulation testing also demonstrated that high waste loadings could be achieved with these tank compositions using the attributes afforded by the CCIM technology.

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DOE OSTI2015

Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design

This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portions of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing.

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DOE OSTI2024

Bonding Environments and Radiation Stabilities of Phosphate Glasses

This report is a summary of some key take-aways presented during the Phosphate Round Table Workshop held in 2020, which was led by PNNL and ANL and funded by DOE-NE. Phosphate glasses have a wide range of commercial and industrial uses due to their unique optical, chemical, and physical properties. They are a candidate material for use as an advanced waste form matrix for immobilizing radioactive waste due to their unique ability to immobilize high fractions of alkali- and sulfur-rich streams in chemically durable glasses. This literature review provides an overview of the structure of phosphate glasses, waste-form related properties of interest (e.g., chemical durability and radiation stability), a summary of how specific composition ratios affect chemical durability (e.g., [Fe]/[P] ratio, [O]/[P] ratio, Fe 2+ /Fe 3+ ratio), as well different ways that phosphate glasses can be affected by radiation. All of these are of interest to the waste form community as phosphate glasses, primarily Fe-P-O glasses, are investigated for usage in immobilizing various types of nuclear waste including U.S. legacy wastes as well as a method for treating salt-based high-level wastes from molten salt reactors and pyroprocessing of used nuclear fuels.

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DOE OSTI2023

Alternative reductants for foam control during vitrification of high-iron High Level Waste (HLW) feeds

Foaming during vitrification of radioactive waste in Joule-Heated Ceramic Melters (JHCM) is exacerbated by trapping of evolving gases, such as CO 2 , NO x and O 2 , beneath a viscous reaction layer. Foaming restricts heat transfer during melting. Sucrose is employed as the baseline additive at the Hanford site in Washington State, USA to reduce foaming. Alternative carbon-based reductant additives were explored in simulated, inactive Hanford high-iron HLW-NG-Fe2 feeds, for both their effect on foaming and to give insight to the behaviour of multivalent species in glass melts under different redox conditions. Graphite, coke (93% C), formic acid and HEDTA additives were compared with sucrose, and a feed with no additive. Graphite and coke additions proved most effective in reducing the maximum foam volume by 51 ± 3% and 54 ± 2%, respectively, compared with 24 ± 5% for sucrose. Lower foaming could result in more efficient vitrification in JHCMs. Reductants also affected redox ratios in the multivalent species present in the feed. The order of reduction, Mn 3+ /Mn 2+ > Cr 6+ /Cr 3+ > Ce 3+ /Ce 4+ > Fe 3+ /Fe 2+ was as predicted on the basis of their redox potentials. There is less reduction overall, particularly in the Fe 3+ → Fe 2+ , than predicted by the calculations, attributed to the oxygenated atmosphere of the experiments.

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DOE OSTI2019

Concept for Operation of the Belgian Geological Disposal Facility for Category B and C Wastes - 19247

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth of between 200 m and 600 m. Within the framework of the first Safety and Feasibility Case (SFC-1) undertaken by ONDRAF/NIRAS, the organization is updating its former GDF layout, which dates back from 2003. The SFC-1 GDF concept is characterized by a central zone in which shafts, joined by a central connecting gallery, provide two surface connections. On opposing sides of the central gallery, four (two on each side) access galleries branch off to connect the disposal areas for B- and C-waste, respectively. Disposal dead end galleries, 400 m long, branch off perpendicular to the access galleries. Both waste areas will be operated during two separate time frames. The current underground layout has been significantly modified from the previous 2003 layout. The modifications are largely based on several R and D studies with regard to the future operations of the GDF carried out by BGE Technology GmbH in support to ONDRAF/NIRAS. This paper focuses on major operational processes such as waste transport and emplacement. In addition, the current conceptual design of the backfilling and closure concept is presented. Optimization studies on the further development of the operational process are also presented. (authors)

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DOE OSTI2015

Hanford building 324 REC hot cell decommissioning approach - 15484

The 324 Building is a facility located at the Hanford Site and is undergoing decontamination and demolition (D and D) activities by Washington Closure Hanford (WCH). In October 1986, an in-cell leak of approximately 1.3 million curies occurred with a substantial fraction being lost into the subsurface soil beneath the floor of a large hot cell complex. Sampling under the hot cell floor revealed hot spots in the soil and cobble reading in excess of 10,000 R. In January of 2014, WCH contracted an AREVA lead team with Kurion as primary design agent to remove the B-Cell floor and highly contaminated underlying soils. The objective of this project is to remove the contaminated debris and soil and repackage them into the existing adjacent hot cells for final disposition as monoliths. The approach to this project uses an innovate combination of existing cell equipment, commercial equipment, and special tools to perform a series of completely remote operations. Also, a novel waste handling approach allows all the highly radioactive waste to remain within the existing cells. The remote handling approach to this project is to use customized off-the-shelf excavators which are mounted and operated remotely inside the cell. These can be moved and replaced remotely with a -Plug-and-Play- connection method. All the cables and utilities for the excavator and tools are managed through this cell connection with connections in a worker-accessible gallery space. The remote excavators are wall-mounted, which frees up the floor for the excavator work. The design also includes a transfer barrier which allows the B-Cell door to remain open during waste transfers while preventing spread of contamination between cells. The project balances the use of existing cell equipment such as cranes and shield doors, with the use of new remote equipment Existing equipment will be removed from the floor of the hot cell to adjacent cells. The concrete floor and stainless steel liner will then be remotely cut and removed, revealing the contaminated soil beneath the cell. The contaminated soil will be mixed with grout in the cell using the excavators and then placed in rigid bins. These bins will then be placed in the adjacent cells using existing overhead cranes. The waste containers allow optimal utilization of the limited adjacent hot cell space and improve the characteristics of the final waste monolith package. The system will use each bin once which significantly decreases the risk of contamination spread. The project includes a full-scale mockup facility in which all installation and operational activities will be tested and proven prior to hot cell construction activities. The mockup facility will also be used for operator training before actual hot work starts. This facility will remain available for use during operations with a full set of spares for redundancy and to support operations modeling during construction and remediation activities. (authors)

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DOE OSTI2024

DECOVALEX-2023: Task E Final Report

This is the Task E final report for DECOVALEX-2023. Task E is focused on understanding thermal, two-phase hydrological, and mechanical (TH2M) processes, especially related to predicting brine migration in the excavation damaged zone around a heated excavation in salt. Salt is attractive as a disposal medium for radioactive waste because it is self-healing and is essentially impermeable and essentially non-porous in the far field (away from excavations). Investigation of the short-term (days to years) near-field (centimeters to tens of meters) behavior of salt is important for radioactive waste disposal because this early period strongly controls the amount of brine in a salt repository. Brine leads to corrosion of waste forms and waste packages, and possible dissolution of radionuclides with brine transport being a potential transport vector to the accessible environment. The main test case used in Task E is the ongoing Brine Availability Test in Salt (BATS) heater test located underground at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, USA. The Task was divided into a series of Steps. Step 0 was an introduction to processes in salt, that included matching historical unheated brine inflow data from boreholes at WIPP and matching temperature observations during BATS heater test 1a. Step 1 included validation of models against a thermo-poroelastic analytical solution relevant to heated boreholes in salt, and two-phase flow around an excavation in salt. Step 2 required all the individual components covered in steps 0 and 1 to come together to match observed brine inflow behavior during the BATS 1a heater test. There were a range of approaches from the teams, from mechanistic to prescriptive. Given the uncertainties in the problem, some teams used one- or two-dimensional models of the processes, while other teams included more geometrical complexity in three-dimensional models. The key learning points from Task E have been: • Heat conduction through salt typically requires non-linear thermal conductivity (as a function of temperature), but most models do a good job matching observations, given appropriate adjustments to the applied power and some thermocouple locations. • Thermal pressurization requires coupled thermal-hydrological-mechanical (THM) responses that consider the thermal expansion of the fluid and solid phases. • Initialization of two-phase flow models around a borehole or excavation in salt are more realistically represented as “wetting up”, rather than “drying down” (i.e., the initial state after excavation is mostly dry, rather than mostly wet). • The BATS 1a heater test includes a significant release of brine after the end of heating, which requires a large increase in permeability to recreate. Task E has been a great learning experience for all the teams involved, and feedback from the modeling teams has led to changes in the design of follow-on BATS experiments, which are now ongoing underground at WIPP. There was a balance throughout the task between freedom to model phenomena how each team saw fit, and prescriptiveness in problem design to bring the modeling teams closer together to allow attribution of smaller differences between models to different modeling choices. The modeling approaches seem to go through two phases: an early phase of discovery or testing, and a later phase of refinement and improvement. In future modeling efforts, different field data could be used (e.g., BATS 2) and more time should be included in the processes for teams to make multiple model refinement or even significant changes to their conceptual model or setup, based on lessons learned from the modeling exercise.

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DOE OSTI2024

Characterization of Precipitate Reactor Feed Tank (PRFT) Batches 15, 22, and 26 from the Defense Waste Processing Facility (DWPF)

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) processes a Monosodium Titanate/Sludge Solids (MST/SS) waste stream received from the Salt Waste Processing Facility (SWPF) via the Precipitate Reactor Feed Tank (PRFT). During processing, DWPF is required to provide evidence of compliance with the Waste Acceptance Product Specifications (WAPS) to ensure acceptance of their vitrified high-level waste (HLW) into the Civilian Radioactive Waste Management System. Production Records must document the constituents of the MST/SS material in the PRFT from each salt batch (StB) processed at SWPF. Savannah River Mission Completion (SRMC) has requested Savannah River National Laboratory (SRNL) to analyze PRFT samples representing each SWPF salt batch for thirty-two radionuclides. Additionally, elemental analysis of PRFT slurry and MST/SS solids was performed to aid SRMC in further refinement of the inputs and assumptions used in future frit development and Material Tracking Program calculations. The analyses of PRFT Batches 15, 22, and 26, which corresponds to material from the processing of StB4, StB5, and StB7, respectively, are reported herein. The unwashed dried solids of the PRFT batches were found to be 86-87% MST. Additionally, the total sulfur values are well below the assumed 982 mg of sulfate/kg of PRFT slurry used in Material Tracking Program calculations.

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DOE OSTI2016

Deep Borehole Field Test Conceptual Design Report

This report documents conceptual design development for the Deep Borehole Field Test (DBFT), including test packages (simulated waste packages, not containing waste) and a system for demonstrating emplacement and retrieval of those packages in the planned Field Test Borehole (FTB). For the DBFT to have demonstration value, it must be based on conceptualization of a deep borehole disposal (DBD) system. This document therefore identifies key options for a DBD system, describes an updated reference DBD concept, and derives a recommended concept for the DBFT demonstration. The objective of the DBFT is to confirm the safety and feasibility of the DBD concept for long-term isolation of radioactive waste. The conceptual design described in this report will demonstrate equipment and operations for safe waste handling and downhole emplacement of test packages, while contributing to an evaluation of the overall safety and practicality of the DBD concept. The DBFT also includes drilling and downhole characterization investigations that are described elsewhere (see Section 1). Importantly, no radioactive waste will be used in the DBFT, nor will the DBFT site be used for disposal of any type of waste. The foremost performance objective for conduct of the DBFT is to demonstrate safe operations in all aspects of the test.

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DOE OSTI2016

WTP Modifications to Support DFLAW: LAW, BOF, Lab - 16219

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is a one-of-a-kind nuclear waste treatment facility being designed and constructed for the U.S. Department of Energy by Bechtel National, Inc. and principal subcontractor URS to process and vitrify radioactive waste stored in underground tanks at the Hanford Site. The Tank Operations Contractor, Washington River Protection Solutions, has responsibility to retrieve, stage, pretreat, and deliver the tank waste feed to WTP. To treat tank waste as soon as practicable, a scenario is being considered to deliver pretreated feed straight to WTP's Low-Activity Waste (LAW) Vitrification Facility. This scenario is called the Direct Feed LAW (DFLAW) program. Balance of Facility utility and system facilities are designed for full WTP operation. Thirteen of these are needed to support reduced throughput during DFLAW operation, of which eight require moderate to extensive changes. Extensive changes include permanent plant modifications that allow DFLAW to operate if the Pretreatment (PT) Facility is unavailable. Under full WTP operation, effluents from the LAW Facility will be returned to the PT Facility for further processing and/or transfer to the Liquid Effluent Retention Facility/Effluent Treatment Facility. The PT Facility also handles all flush water from the tank farm transfers. To address these requirements, a Value Improvement Project was performed. The recommended approach is to provide a 'split stream' Effluent Management Facility (EMF) with an evaporator to support DFLAW. The required modifications for DFLAW facilities are discussed in this paper. (authors)

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DOE OSTI2024

Sorption of Metals and Radionuclides in Cementitious Leachate Impacted Sediments from the Savannah River Site

Performance assessments (PAs) are calculations used to establish the risk posed by radioactive waste disposal facilities in the Savannah River Site (SRS). The distribution coefficient (Kd; solid/liquid concentration ratio; Csolid/liquid) is one of the key geochemical parameters used in the PA to provide a measure of how strongly the dissolved contaminants bind to solid phases; the greater the Kd value, the greater the tendency to bind to the solids. The overall purpose of this study was to measure this key parameter for several constituents of concern in the cementitious leachate impacted SRS sediment environments. Particular attention was directed at quantifying the influence of cementitious leachate from various stages of cement aging on the sorption of radioactive waste and heavy metals on SRS sediments.

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DOE OSTI2017

Savannah River Remediation Liquid Waste Operations Making the Workplace Safer - Lean and Human Performance Improvement Initiatives - 17244

The Savannah River Site, a U.S. Department of Energy facility, contracted Savannah River Remediation (SRR) to manage its liquid waste project, which is a unique, high hazard nuclear waste treatment and disposal facility consisting of 51, one million gallon underground waste tanks, a solvent extraction processing unit, a vitrification facility for the highest radioactive waste, and a grout solidification and disposal facility for a decontaminated salt solution. Our Project mission is to treat the nuclear waste and close the associated storage and treatment facilities, of which, 8 of the 51 tanks have been closed to date. Over the past three years, SRR implemented a Lean Business Management System to address site inefficiencies, aging infrastructure concerns and budget constraints. The Lean philosophy started with the work of W. Edwards Deming and his contemporaries. In more recent years the process has matured in the Toyota Production System and migrated into fields as diverse as services industries, health-care, and now the DOE complex. Like safety, continuous improvement is a fundamental core business value that starts at the top. The Lean Business System, led by our executive management team, drives quality, production and waste elimination by focusing on People using a structured Process that allows the workforce to identify and fix problems in their workplace. Many inefficiencies and quality related issues start with the desire to make systems fail-safe without defining the core requirements for which that process or program was created. This leads to added complexity and latent or inherent human performance issues embedded within our processes. While Lean can't fix all problems, a combination of strong nuclear safety culture, robust conduct of operations, active human performance improvement process, and a business practice of continuously looking for improvements leads to excellence across our industry and creates a culture that is worthy of promoting. Without strong management, our systems fail because no one sponsors or encourages those attributes we covet. Leadership starts at the top with our customer defining what provides 'value' to them, and filters down through to the organization and ultimately the workforce in day to day conduct of operations. It is imperative that our customer drives these desirable ideals through recognition and contractual requirements to encourage and support those behavioral traits that are critical to our ongoing success.This paper focuses on how the Lean Business Management System supports those attributes that we value: a strong nuclear safety culture; disciplined conduct of operations; and a culture of robust human performance and continuous improvement. These attributes in our daily work environment propel a safer, more consistent, more efficient workplace. (authors)

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DOE OSTI2024

Migration of 60 Co, 133 Ba, 137 Cs, and 152 Eu from cementitious wasteforms in field lysimeter experiments

Safe and effective storage of radioactive waste is essential to protect 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 work analyzed the vadose zone migration of radionuclides from cementitious wasteforms at the Savannah River Site after ten years. The observed radionuclides 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 in the field 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 137 Cs, migration through the sediment control system increased with decreasing ionic potential (ionic charge/radius); migration order: 152 Eu< 137 Cs< 60 Co< 133 Ba. Overall, the cementitious wasteforms were observed to decrease radionuclide migration extent relative to the filter paper. In both cementitious wasteforms, the migration extent increased in the order 152 Eu< 133 Ba< 60 Co< 137 Cs. 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 actinide element isotopes.

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