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DOE OSTI2015

Optimization of salt-cake removal flowsheet at SRS through incorporation of testing and in-tank waste experience - 15263

Salt-cake removal at SRS may be performed for several reasons: to provide space for evaporator operation (i.e., to precipitate more salt in the drop tank), to provide feed for salt processing (i.e. immobilize the waste), or to remove the salt for tank closure. Many different salt dissolution techniques have been employed in the 40 years that SRS has been performing salt removal, from a basic 'Add, Sit, Remove' method (water is added on top of the salt-cake and time is allowed for diffusion), to performing interstitial liquid removal, or using mixing devices to promote contact with the liquid. Lessons learned from previous salt-cake removal campaigns, in addition to testing and modeling, have led to opportunities for improvements to the overall salt-cake removal process. This includes better understanding of salt properties and behavior during dissolution; the primary concerns for salt dissolution are the release of radiolytic hydrogen and criticality prevention (post-dissolution). Recent developments in salt dissolution include the reuse of dilute supernate and a semi-continuous dissolution (SCD) process, where low volume mixing eductors are used to deliver water near the surface of the salt-cake at the same rate as the salt solution is removed and transferred to a receipt tank. (authors)

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DOE OSTI2019

Tank Closure Cesium Removal Project CST Simulant Cesium Batch Contact Kinetics Test Results and Column Performance Predictions

To evaluate the cesium removal capabilities of the caustic-washed R9120-B version of crystalline silicotitanate (CST) ion exchange media from Savannah River Site (SRS) waste during Tank Closure Cesium Removal (TCCR) column processing, batch contact kinetics tests were performed at 23 °C in SRS Average simulant. Cesium loading kinetics data was also collected (in current and previous testing; King, et al., 2018a) using the same simulant for an older CST sample (IE- 911) which was caustic-washed by the vendor many years ago and was recently retrieved from drums stored at SRS. (Note that R9120-B and IE-911 are believed to be identical materials, after the R9120-B was caustic washed by SRNL, and differ only by particle size and manufacturing date, and thus can be considered different batches). Although cesium loading kinetics testing can be conducted more rigorously using dynamic column tests, the batch contact method requires less simulant and considerably less experimental effort. Average percent removal data for batch contact samples for the two CST batches are provided in Table ES-1. Faster cesium uptake was observed with IE-911, although the loading rates for the two media batches were similar. After 5.5 hours of CST contact, 81% cesium removal was observed with R9120-B versus 86% removal with IE-911. Although these differences are small, the trend is consistent across the timed results, with the larger bead material always loading slower. These seemingly small differences in percent removal are believed to have a large impact on the column performance. Cesium loading curves for the two CST batches are provided in Figure ES-1 below. Faster cesium loading with IE-911 CST is presumed to be associated with the smaller average particle diameter (408 μm mean volume diameter) versus R9120-B (572 μm diameter). Cesium loading kinetic data was needed to conduct column performance comparisons (i.e., model predictions) for the two CST batches.

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DOE OSTI2015

Optimization of Saltcake Removal Flowsheet at SRS through Incorporation of Testing and In-Tank Waste Experience

Saltcake removal at SRS may be performed for several reasons: to provide space for evaporator operation (i.e., to precipitate more salt in the drop tank), to provide feed for salt processing (i.e. immobilize the waste), or to remove the salt for tank closure. Many different salt dissolution techniques have been employed in the 40 years that SRS has been performing salt removal, from a basic “Add, Sit, Remove” method (water is added on top of the saltcake and time is allowed for diffusion), to performing interstitial liquid removal, or using mixing devices to promote contact with the liquid. Lessons learned from previous saltcake removal campaigns, in addition to testing and modeling, have led to opportunities for improvements to the overall saltcake removal process. This includes better understanding of salt properties and behavior during dissolution; the primary concerns for salt dissolution are the release of radiolytic hydrogen and criticality prevention (post-dissolution). Recent developments in salt dissolution include the reuse of dilute supernate and a semi-continuous dissolution (SCD) process, where low volume mixing eductors are used to deliver water near the surface of the saltcake at the same rate as the salt solution is removed and transferred to a receipt tank.

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DOE OSTI2016

Results for the Fourth Quarter Calendar Year 2015 Tank 50H Salt Solution Sample

In this memorandum, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2015 (CY15) sample of Tank 50H salt solution are presented in tabulated form. The Fourth Quarter CY15 Tank 50H samples were obtained on October 29, 2015 and received at Savannah River National Laboratory (SRNL) on October 30, 2015. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter Calendar Year 2015 (CY15) sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan.

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DOE OSTI2015

Actinide Removal Process October 2014 Sample Analysis

Filtration within the Actinide Removal Process (ARP) currently limits the throughput in interim salt processing at the Savannah River Site. In this process, batches of salt solution with Monosodium Titanate (MST) sorbent are concentrated by crossflow filtration. The filtrate is subsequently processed to remove cesium in the Modular Caustic Side Solvent Extraction Unit (MCU) followed by disposal in saltstone grout. The concentrated MST slurry is washed and sent to the Defense Waste Processing Facility (DWPF) for vitrification. During past ARP processing, there has been a degradation of filter performance manifested as the inability to maintain high filtrate flux throughout a multi-batch cycles during Salt Batches 6, 5 and 4. The objectives of the current effort for Salt Batch 7 samples, which are similar to the most recent ARP sample analysis from 2013, were to characterize the feed streams, to determine if solids (in addition to MST) are precipitating and causing the degraded performance of the filters, and to assess the particle size and rheological data to address potential filtration impacts. The samples from the LWPT and LWHT were obtained from several stages of processing of Salt Batch 7B, Cycle 1, Batch 67.

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DOE OSTI2016

The Path to Nitrate Salt Disposition

The topic is presented in a series of slides arranged according to the following outline: LANL nitrate salt incident as thermal runaway (thermally sensitive surrogates, full-scale tests), temperature control for processing, treatment options and down selection, assessment of engineering options, anticipated control set for treatment, and summary of the overall steps for RNS.

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DOE OSTI2020

Proposed guidance for preparing and reviewing a molten salt non-power production or utilization facility application

Development of non-power molten salt reactors (MSRs) are under consideration to further establish an MSR experience base, support the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.43(e), and provide any additional analyses needed for development of a full-scale MSR. Guidance provided in this report is based on MSRs operating with liquid fuel (i.e., fuel dissolved within a molten salt). These reactors, unless owned by the DOE or DOD, will require licensing by the US Nuclear Regulatory Commission (NRC) staff. Standard review plan (SRP) guidance for large light water reactors (LWRs) is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants; Light Water Reactor (LWR) Edition. However, NRC staff observed that NUREG-0800 is very cumbersome to apply to non-power reactors “because of the great differences in complexity and hazards between non-power reactors and nuclear power plants.” Therefore, a program to develop performance-based guidance applicable to non-power reactors was initiated. In 1996, NUREG-1537, Parts 1 and 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, was published. Part 1, the format and content guide, suggests a uniform format for presenting information in non-power reactor applications that is acceptable to the NRC staff, but conformance with the format and content is not required. Part 2, the SRP, ensures the quality and uniformity of the staff review of an application. Unfortunately, the application guidelines and SRP do not provide adequate guidance for all advanced non-LWR technologies and applications. This discrepancy eventually led to the 2012 development of interim staff guidance (ISG) for NUREG-1537, which includes criteria for describing and reviewing aqueous homogeneous reactors (AHRs). Specifically, NUREG-1537 ISG, 2012 expanded the original document to address three areas: 1. updated criteria for heterogeneous non-power reactors, 2. criteria for licensing AHRs, and 3. criteria for licensing a Part 50-licensed isotope production facility. In 2015, the US Department of Energy (DOE) opted to build on the AHR NUREG-1537 ISG experience by performing a gap analysis of the guidance that would be used to license a non-power MSR. MSRs represent one of the advanced non-LWR technologies selected by DOE for development through a multiyear cost share award with Southern Company Services. Under this Advanced Reactor Concepts 2015 (DOE Advanced Reactor Concepts [ARC] 15) award program, the DOE tasked Oak Ridge National Laboratory (ORNL) to evaluate the guidance changes that the NRC may need to consider when licensing an MSR non-power reactor. ORNL staff, with support from Boston Government Services, LLC, focused on five system-related chapters in NUREG-1537 that were considered most relevant to inform the effort that would be required for a non-power MSR applicant. ORNL documented this review in a technical report, ORNL/TM-2018/834, Proposed Guidance for Preparing and Reviewing Molten Salt Non-Power Reactor License Applications (NUREG-1537). The report was subsequently shared with industry and the NRC. The 2018 review was limited in scope, focusing on key system chapters based on the expected significance of each chapter relative to expected differences in addressing advanced non-LWR technologies, specifically non-power MSRs, compared with heterogeneously fueled non-power reactors. In the ORNL report, proposed generic adaptations were suggested for the following NUREG-1537 chapters: Chapter 4, “Reactor Description”; Chapter 5, “Reactor Coolant Systems”; Chapter 6, “Engineered Safety Features”; Chapter 9, “Auxiliary Systems”; Chapter 11, “Radiation Protection Program and Waste Management” The inclusion of Chapter 11 in the previous review effort was intended to provide guidance for categorizing the waste-handling process for an MSR operating with homogenous fuel. The introductions from Parts 1 and 2 of the 2012 NUREG-1537 ISG provide guidance for the application and review of production facilities. After a period of operation, non-power MSRs with homogenous fuel will include gaseous and soluble fission products. The gaseous fission products will be collected and held for decay in an off-gas system. There might also be an initiative to polish or filter the soluble fission products in the fuel salt by some mechanical or chemical means. The treatment and handling of fission products in the non-power MSR fuel salt and the description of this process in the safety analysis report (SAR) must be very precise to avoid the waste treatment facility being construed as a co-located special nuclear material (SNM) fuel cycle facility (see Section 2.3 of this report). Subsequent to the release of ORNL/TM-2018/834, NRC staff expressed a desire to continue the regulatory gap analysis that was begun in that report. This would provide additional clarity and information addressed in certain sections of the original report, while also providing new guidance on certain topics not addressed in the original report. This revision would benefit the NRC staff reviewing applications involving non-power MSR designs and would help developers understand how the NRC staff might approach the review of such applications. The focus of this report is to provide infrastructure support to the NRC staff for the regulatory review of non-power MSRs.

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DOE OSTI2016

Daher-TLI UCSC Wash Process - Lessons Learned from Handling a Legacy Cylinder - 16506

The Daher-TLI UF{sub 6} Cylinder Service Center (UCSC) began operation in 2013 and functions as a dedicated wash, re-certification, storage and disposal facility for 122 cm uranium hexafluoride (UF{sub 6}) cylinders. The primary purpose of the facility is to provide cylinder cleaning and re--certification services to the UF{sub 6} industry for cylinders that previously contained natural or depleted UF6. The facility also has the ability to recover most of the uranium in the form of uranyl peroxide (UO{sub 4}) or triuranium octoxide (U{sub 3}O{sub 8}). Small residues of UF{sub 6} (commonly referred to as a heel) from natural cylinders are removed and treated in a patented three-step purification process that recovers U{sub 3}O{sub 8} that meets the ASTM specification. The wastewater ordinarily contains a mixture of sodium salts (77% sodium sulfate, 23% sodium nitrate) along with minute traces of uranium (1 - 2 mg/L) and fluoride (10 - 20 mg/L). Over a period of 6 months, lab analyses were performed on the filtrate from a legacy cylinder in an attempt to separate uranium and fluoride from this stream. These pilot results established near quantitative recovery of uranium (< 1 mg/L) and near complete separation of fluorides (5 - 15 mg/L). This paper follows the efforts of the Daher-TLI UCSC to process the abnormal heeled UF{sub 6} cylinder. This paper also documents the lessons learned from this experience and how those lessons were used to improve the UCSC wash process. These modifications served to recover 99.87% of the input uranium as UO{sub 4} while simultaneously creating a waste stream with only trace amounts of impurities. (authors)

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DOE OSTI2020

Qualification of Processing Batches for Tank Closure Cesium Removal Through In-Tank Batch Contact Testing - 20470

The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)

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DOE OSTI2019

CST Batch Contact Equilibrium Testing of TCCR Tank 10H Batch 1 and 1A Waste Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that both the Tank 10H Batch 1 (pre-production) and 1A (production) waste supernates are unique solutions for which CST cesium removal performance is lower than predicted by ZAM modeling. Cesium distribution coefficients (K d ), percent removal, and CST loading data for the TCCR production supernate batch are provided in Table ES-1. ZAM-based cesium loading predictions are compared to the observed loading for the production batch in Figure ES-1, where significantly lower performance was observed than predicted by the model. Even lower cesium loading results (versus ZAM) were observed previously for the in-tank “teabag” batch contact samples.

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DOE OSTI2017

Savannah River Remediation Liquid Waste Operations Lifecycle Savings through a Lean Business System at the Savannah River Site - 17245

Savannah River Remediation (SRR) was contracted by the U.S. Department of Energy (DOE) to manage its liquid waste project at the Savannah River Site. The liquid waste project 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 lower radioactive 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. 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. At SRR, a Lean business management system was adopted to address site inefficiencies, aging infrastructure concerns and budget constraints. With yearly escalation and flat budgets, continuous improvement is taking on a larger role - improving the work accomplished for each dollar. This paper focuses on the program results achieved to date, including the lifecycle savings identified through the Rapid Improvement Event (RIE) process. The RIE process is a Lean tool which creates a cross-functional team to analyze, develop and implement improvements for a given process. DOE-EM has recently emphasized the value of contractor's continuous improvement programs through their implementation of Plan of Action and Milestone (POAM): Goal 2 (Increase Efficiency/Improve Performance) Reduce EM Operations Cost in FY16. POAM Goal 2 was a FY 16 effort to achieve 5% cost savings across most programs. This past year's cost savings initiatives will likely become a trend as DOE and its contractors look to do more with less. In support of our own business objectives, as well as, DOE's motivation to improve efficiencies, SRR has held over one hundred formal Lean events and dozens of 5S events that have driven improvement across the Project, including tank closure and production facilities. These improvements have significantly reduced the time required to process material and achieve the overall goal of hazard reduction and tank closure. (authors)

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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 OSTI2018

Thermal evaluations for mixed organic and nitrate salt drums in interim storage at the WCS Federal Waste Facility

Waste Control Specialists LLC (WCS) operates the Federal Waste Facility (FWF) in Texas that is licensed to process and store certain types of mixed low-level radioactive waste packages. The specific waste currently housed at WCS includes some of the transuranic (TRU) waste that originated at the DOE Los Alamos National Laboratory (LANL) and are destined for disposal at the DOE Waste Isolation Pilot Plant (WIPP) facility. The TRU waste within each Standard Waste Box (SWB) was packaged in 55-gallon drums at LANL. Some of the drums contain mixed organic and nitrate salt contents suspected of experiencing a chemical reaction. The drums (up to four drums in a SWB) were then overpacked in a SWB prior to shipment to WCS. In the most common configuration, WCS placed the SWB containing the drums on metal pallets and stored two SWBs into Modular Concrete Canisters (MCCs) in a double stacked array. The MCCs were placed in a staggered array at a segregated area of the FWF and covered with sand to mitigate the heating that occurred when stored above ground on waste pads. The temporary interment is also intended to provide additional containment of the suspect reactive waste to mitigate release of radioactivity in an incident such as that which occurred at WIPP for one such drum. Void space of each MCC was filled with 1-inch pea gravel to facilitate retrievability. Thermocouple probes were placed within each MCC to monitor heat generation from the SWBs and the MCC lids were closed. The primary objective of this analysis was to estimate the heat generation of the SWB drum contents when the SWB package temperature inside the MCC and the FWF ambient temperature were monitored.

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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 OSTI2023

Review of Mercury Sequestration in Cementitious Waste Forms

Decontaminated salt solution from Tank 50 at the Savannah River Site (SRS) is known to contain dissolved mercury species which are present as hydroxides and oxides. These species are reported in analytical measurements as the organic cation [CH 3 Hg] + and the inorganic cation Hg 2+ . The salt solution is mixed with cementitious reagents, blast-furnace slag (BFS) and thermally beneficiated Class F fly ash in the Saltstone Processing Facility and pumped into engineered Saltstone Disposal Units, (SDUs), to produce a solidified waste form. The solidified waste form, saltstone, encapsulates and chemically stabilizes the mercury and other contaminants, sequestering them from release into the environment. With the discovery of methylmercury in the salt solution, testing was performed to determine the extent to which saltstone prepared with methylmercury (MeHg) retains mercury in the Toxicity Characteristic Leaching Procedure (TCLP). Results from this testing indicated that the amount of mercury extracted remained below the regulatory limit, and that the saltstone formulation with a greater proportion of BFS performed slightly better at retaining Hg in the solid.

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DOE OSTI2016

PLAN-TA9-2443(U), Rev. B Remediated Nitrate Salt (RNS) Surrogate Formulation and Testing Standard Procedure

This document identifies scope and some general procedural steps for performing Remediated Nitrate Salt (RNS) Surrogate Formulation and Testing. This Test Plan describes the requirements, responsibilities, and process for preparing and testing a range of chemical surrogates intended to mimic the energetic response of waste created during processing of legacy nitrate salts. The surrogates developed are expected to bound1 the thermal and mechanical sensitivity of such waste, allowing for the development of process parameters required to minimize the risk to worker and public when processing this waste. Such parameters will be based on the worst-case kinetic parameters as derived from APTAC measurements as well as the development of controls to mitigate sensitivities that may exist due to friction, impact, and spark. This Test Plan will define the scope and technical approach for activities that implement Quality Assurance requirements relevant to formulation and testing.

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DOE OSTI2024

KOMPASS-II: Compaction of Crushed salt for Safe Containment – Phase 2

Long-term stable sealing elements are a basic component in the safety concept for a possible repository for heat-emitting radioactive waste in rock salt. The sealing elements will be part of the closure concept for drifts and shafts. They will be made from a welldefinied crushed salt in employ a specific manufacturing process. The use of crushed salt as geotechnical barrier as required by the German Site Selection Act from 2017 /STA 17/ represents a paradigm change in the safety function of crushed salt, since this material was formerly only considered as stabilizing backfill for the host rock. The demonstration of the long-term stability and impermeability of crushed salt is crucial for its use as a geotechnical barrier. The KOMPASS-II project, is a follow-up of the KOMPASS-I project and continues the work with focus on improving the understanding of the thermal-hydraulic-mechanical (THM) coupled processes in crushed salt compaction with the objective to enhance the scientific competence for using crushed salt for the long-term isolation of high-level nuclear waste within rock salt repositories. The project strives for an adequate characterization of the compaction process and the essential influencing parameters, as well as a robust and reliable long-term prognosis using validated constitutive models. For this purpose, experimental studies on long-term compaction tests are combined with microstructural investigations and numerical modeling. The long-term compaction tests in this project focused on the effect of mean stress, deviatoric stress and temperature on the compaction behavior of crushed salt. A laboratory benchmark was performed identifying a variability in compaction behavior. Microstructural investigations were executed with the objective to characterize the influence of pre-compaction procedure, humidity content and grain size/grain size distribution on the overall compaction process of crushed salt with respect to the deformation mechanisms. The created database was used for benchmark calculations aiming for improvement and optimization of a large number of constitutive models available for crushed salt. The models were calibrated, and the improvement process was made visible applying the virtual demonstrator.

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DOE OSTI2017

Experiments and Modeling to Support Field Test Design

Disposition of heat-generating nuclear waste (HGNW) remains a continuing technical and sociopolitical challenge. We define HGNW as the combination of both heat generating defense high level waste (DHLW) and civilian spent nuclear fuel (SNF). Numerous concepts for HGNW management have been proposed and examined internationally, including an extensive focus on geologic disposal (c.f. Brunnengräber et al., 2013). One type of proposed geologic material is salt, so chosen because of its viscoplastic deformation that causes self-repair of damage or deformation induced in the salt by waste emplacement activities (Hansen and Leigh, 2011). Salt as a repository material has been tested at several sites around the world, notably the Morsleben facility in Germany (c.f. Fahland and Heusermann, 2013; Wollrath et al., 2014; Fahland et al., 2015) and at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, NM. Evaluating the technical feasibility of a HGNW repository in salt is an ongoing process involving experiments and numerical modeling of many processes at many facilities.

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