Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Waste Forms”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 775 records · Page 43

Combined effects of radiation damage and He accumulation on bubble nucleation in Gd2Ti2O7

Pyrochlores have long been considered as host phases for long-term immobilization of radioactive waste nuclides that would undergo ..alpha..-decay for hundreds of thousands of years. This work utilizes ion-beam irradiations to examine the combined effects of radiation damage and He accumulation on bubble formation in Gd2Ti2O7 over relevant waste-form timescales. Helium bubbles are not observed in pre-damaged Gd2Ti2O7 implanted with 2 x 1016 He/cm2, even after post-implantation irradiations with 7 MeV Au3+ at 300, 500, and 700 K. However, He bubbles with average diameters of 1.5 nm and 2.1 nm are observed in pre-damaged (amorphous) Gd2Ti2O7 and pristine Gd2Ti2O7, respectively, after implantation of 2 x 1017 He/cm2. The critical He concentration for bubble nucleation in Gd2Ti2O7 is estimated to be 6 at.% He.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Framework for The Analysis of US Permanent Geologic Disposal Options for Radioactive Waste

The U.S. study of permanent geologic disposal options of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) provided a technical basis for informing policy decisions regarding strategies for the management and disposal of radioactive waste requiring geologic isolation through the evaluation of potential impacts of waste forms on the feasibility and performance of representative generic concepts for geologic disposal. The goal of the study was to help inform relevant policy questions including: Is a "one-size-fits-all" repository a good strategic option for disposal? Do different waste types and forms perform differently enough in different disposal concepts that they warrant different treatment? Do some disposal concepts perform significantly better with or without specific waste types or forms?

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Overview of Features, Events, and Processes Affecting the Postclosure Performance of a Deep Borehole Disposal Facility - 17168

The United States Department of Energy's Office of Nuclear Energy (DOE-NE) is evaluating the feasibility of disposing of smaller DOE-managed waste forms in 5-km deep boreholes drilled in crystalline rock. It is thought that both the depth and the geological conditions at that depth will serve to isolate the waste from the accessible environment after the disposal facility is closed. This paper discusses the features, events, and processes (FEPs) that could affect the postclosure performance of a deep borehole disposal (DBD) facility and identifies those that would likely need to be considered in the context of conducting a performance assessment analysis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The U.S. Nuclear Waste Technical Review Board Intern Program: Developing Next Generation Nuclear Waste Professionals - 16102

The U.S. Nuclear Waste Technical Review Board (Board) created an internship program in the fall of 2013 for undergraduate and graduate students interested in management and disposal of spent nuclear fuel and high-level radioactive waste. Interns must select a project topic in a technical area related to the Board's mission and prepare a well-supported briefing for the agency. The ability to review technical and scientific information in a critical and unbiased manner is essential for completion of the intern's project and this is a particular focus of the interaction between the mentor and the intern. The interns from 2014 and 2015 selected and reviewed topics related to the areas of commingling waste forms for disposal and deep borehole disposal. In addition to the technical project, interns acquire a broad perspective of the federal agencies involved in managing and disposing of radioactive waste while participating in inter-agency networking opportunities. In this review of the Board internship experience, two past interns discuss the personal impact of the program and how they believe the experience prepared them to face the challenges of a career in management and disposal of spent nuclear fuel and high-level radioactive waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tank Side Cesium Removal System Project and Technology Maturation Program - 19519

Washington River Protection Solutions (WRPS) is the Tank Operating Contractor (TOC) for the U.S. Department of Energy-Office of River Protection (DOE-ORP) on the Hanford Site. The Hanford Site stores an estimated 56 million gallons of mixed radioactive and chemically hazardous waste in large underground tanks. WRPS is in the process of designing the Tank Side Cesium Removal (TSCR) system to produce Low Activity Waste (LAW) from Hanford tank waste. The LAW will be transferred to the Waste Treatment and Immobilization Plant (WTP) LAW Vitrification Facility, where the hazardous constituents in the LAW will be immobilized in a durable glass waste form for disposal. The TSCR Project demonstrates a tank-side treatment system for providing feed to the WTP LAW Vitrification Facility. This system removes cesium from supernatant tank waste using non-elutable cesium ion exchange media to meet the applicable waste acceptance criteria for the WTP. In support of the project, a tailored Technology Maturation/Testing Plan (TMP) has been developed to identify principal Technology Elements (TEs) for the proposed process, designate Critical Technology Elements (CTEs), identify gaps in the existing knowledge base and identify the necessary activities to mature the technologies for successful deployment of TSCR. The project scope and technology maturation approach are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in coordination with South Carolina Department of Health and Environmental Control (SCDHEC). DOE has completed or begun additional work related to salt waste disposal to address these factors. NRC staff continues to evaluate information related to the performance of the SDF and has been working with DOE and SCDHEC to resolve NRC staff's technical concerns. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A Decade of Innovative Approaches to Treating and Dispositioning Radioactive Waste at the Savannah River Site - 20620

This year marks the 70. anniversary of the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina. SRS is a key industrial complex responsible for environmental stewardship, environmental cleanup, waste management, and disposition of nuclear materials. SRS also continues to have a role in critical defense-related activities and the reprocessing of used reactor fuel. The SRS encompasses 803 square kilometers (310 square miles) in parts of Aiken, Barnwell, and Allendale counties. The Liquid Waste Mission starts with the safe receipt and storage of radioactive liquid waste in the waste tanks, which principally includes Cold War legacy waste, but also waste from support of National Aeronautics Space Administration missions, medical isotope production, and research activities. This 133,000 cubic meters (m{sup 3}) (35 million gallons [Mgal]) of high-level radioactive waste (HLW) is currently held in 43 large underground waste tanks. The capacity of each of these tanks range from 2,800 m{sup 3} to 5,000 m{sup 3} (0.75 to 1.33 Mgal) and were placed in operation between 1954 and 1986. In July 2009, the DOE awarded a stand-alone contract to execute the Liquid Waste Mission at SRS. The Liquid Waste Contractor selected was Savannah River Remediation LLC1 (SRR), which is responsible for receipt, storage, retrieval and treatment of all HLW, disposal of the decontaminated low-activity waste fraction, as well as operationally closing cleaned HLW tanks. Since July 2009, SRR has successfully grouted and operationally closed six large underground tanks, poured 1,476 canisters of vitrified HLW (a stable glass waste form), treated approximately 34,200 m{sup 3} (9.0 Mgal) of HLW, and dispositioned over 41,200 m{sup 3} (10.9 Mgal) of decontaminated salt solution as low-level waste (LLW) into the Saltstone Disposal Units (SDUs) being constructed by SRR on site. The innovative work performed by SRR in the LLW portion of its mission has earned it the Richard S. Hodes Award. The work conducted by SRR is the same work championed by the late Richard S. Hodes, a respected physician, statesmen, and Chairman of the Southeast Compact Commission for Low-Level Radioactive Waste Management. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tank Side Cesium Removal System Project and Technology Maturation Program - 20101

Washington River Protection Solutions (WRPS) is the Tank Operating Contractor (TOC) for the U.S. Department of Energy-Office of River Protection (DOE-ORP) on the Hanford Site. The Hanford Site stores an estimated 56 million gallons of mixed radioactive and chemically hazardous waste in large underground tanks. WRPS is in the process of designing the Tank Side Cesium Removal (TSCR) system to produce a Low Activity Waste (LAW) feed from existing mixed Hanford tank waste. The LAW will be transferred to the Waste Treatment and Immobilization Plant (WTP) LAW Vitrification Facility, where it will be immobilized in a durable glass waste form for disposal. The TSCR Project demonstrates a tank-side treatment system for providing feed to the WTP LAW Vitrification Facility. This system removes undissolved solids and cesium from tank waste supernatant using non-elutable ion exchange media to meet the applicable waste acceptance criteria for the WTP. In support of the project, technology testing has been performed to answer design questions and reduce risk. The project scope and technology testing approach are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Stability of Super Absorbent Polymers in Sodium Nitrate - 16222

The stability of two different chemical types of cross-linked super absorbent polymers has been explored using a surrogate waste of 3M Sodium Nitrate (NaNO{sub 3}) that had pH adjusted using concentrated Nitric Acid (HNO{sub 3}) to pH of 1, 2, 4 and 6. Tests showed both types of polymer gelled the surrogate waste liquids. On-going stability tests revealed that there was no breakdown of the waste form after 11 months and no release of free liquids. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Goethite Bench-scale and Large-scale Preparation Tests

The Hanford Waste Treatment and Immobilization Plant (WTP) is the keystone for cleanup of high-level radioactive waste from our nation's nuclear defense program. The WTP will process high-level waste from the Hanford tanks and produce immobilized high-level waste glass for disposal at a national repository, low activity waste (LAW) glass, and liquid effluent from the vitrification off-gas scrubbers. The liquid effluent will be stabilized into a secondary waste form (e.g. grout-like material) and disposed on the Hanford site in the Integrated Disposal Facility (IDF) along with the low-activity waste glass. The major long-term environmental impact at Hanford results from technetium that volatilizes from the WTP melters and finally resides in the secondary waste. Laboratory studies have indicated that pertechnetate ({sup 99}TcO{sub 4}{sup -}) can be reduced and captured into a solid solution of {alpha}-FeOOH, goethite (Um 2010). Goethite is a stable mineral and can significantly retard the release of technetium to the environment from the IDF. The laboratory studies were conducted using reaction times of many days, which is typical of environmental subsurface reactions that were the genesis of this new process. This study was the first step in considering adaptation of the slow laboratory steps to a larger-scale and faster process that could be conducted either within the WTP or within the effluent treatment facility (ETF). Two levels of scale-up tests were conducted (25x and 400x). The largest scale-up produced slurries of Fe-rich precipitates that contained rhenium as a nonradioactive surrogate for {sup 99}Tc. The slurries were used in melter tests at Vitreous State Laboratory (VSL) to determine whether captured rhenium was less volatile in the vitrification process than rhenium in an unmodified feed. A critical step in the technetium immobilization process is to chemically reduce Tc(VII) in the pertechnetate (TcO{sub 4}{sup -}) to Tc(Iv)by reaction with the ferrous ion, Fe{sup 2+}-Fe{sup 2+} is oxidized to Fe{sup 3+} - in the presence of goethite seed particles. Rhenium does not mimic that process; it is not a strong enough reducing agent to duplicate the TcO{sub 4}{sup -}/Fe{sup 2+} redox reactions. Laboratory tests conducted in parallel with these scaled tests identified modifications to the liquid chemistry necessary to reduce ReO{sub 4}{sup -} and capture rhenium in the solids at levels similar to those achieved by Um (2010) for inclusion of Tc into goethite. By implementing these changes, Re was incorporated into Fe-rich solids for testing at VSL. The changes also changed the phase of iron that was in the slurry product: rather than forming goethite ({alpha}-FeOOH), the process produced magnetite (Fe{sub 3}O{sub 4}). Magnetite was considered by Pacific Northwest National Laboratory (PNNL) and VSL to probably be a better product to improve Re retention in the melter because it decomposes at a higher temperature than goethite (1538 C vs. 136 C). The feasibility tests at VSL were conducted using Re-rich magnetite. The tests did not indicate an improved retention of Re in the glass during vitrification, but they did indicate an improved melting rate (+60%), which could have significant impact on HLW processing. It is still to be shown whether the Re is a solid solution in the magnetite as {sup 99}Tc was determined to be in goethite.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Demonstration of SuperLig{sup R} 605 and SuperLig{sup R} 644 in a Regeneration Flow Sheet to Remove Radioactive Strontium and Cesium from Seawater - 16236

Remediation techniques are needed for efficient removal of radioactive strontium and cesium from seawater. Molecular Recognition Technology, using selective SuperLig{sup R} resins, demonstrates a clean chemistry approach to solving this problem. Molecular Recognition Technology offers high levels of separation of chemically pure metal contaminants, whose subsequent disposal is compatible with simple waste forms (grout and glass). SuperLig{sup R} resins can be regenerated with common reagents and used many times, limited by a dose of 1 E+9 Rad, which assures low volumes of secondary solid waste. The metal selectivity of the SuperLig{sup R} resin ensures that the solid waste is of low residual activity. These properties of sustainability are illustrated in a project that demonstrated the ability of two SuperLig{sup R} resins to selectively remove Cs and Sr from a seawater mimic representative of the contaminated seawater in Fukushima, Dai'ichi, Japan harbor. Removal of radioactive Sr-90 and Cs-134/Cs-137 from seawater using conventional ion exchange or a complexing agent is complicated by the presence of large excesses of competing ions such as Na, K, Ca, and Mg, which can be co-extracted. These competing ions, if co-extracted, reduce the capacity of a single pass resin with the consequent increase in volume of spent resin or decrease in capacity of a regenerable resin. The successful demonstration reported in this study has shown the ability of two proprietary resins, SuperLig{sup R} 605 and SuperLig{sup R} 644, to remove Sr and Cs, respectively, with high selectivity from simulated solutions containing these metal ions together with major constituents of seawater at their normal concentrations. This effort was part of a demonstration project to investigate methods for removal of Sr-90 and Cs-134/Cs-137 from 160,000 cubic meters of seawater impounded at Fukushima, Dai'ichi, Japan harbor. SuperLig{sup R} resins can be regenerated by elution with small volumes of simple acid solutions. SuperLig{sup R} resins have a high tolerance to radiation, with their performance not being degraded by doses of 1.0 E+9 Rad. Estimates, based on the results obtained, are that Cs and Sr levels in the volume of seawater in the Fukushima, Dai'ichi, Japan harbor could be reduced to desired discharge limits in 200 days with some 30 cubic meters of the two SuperLig{sup R} resins and that decontamination factors of 10 or more could be achieved. Results of the demonstration, together with advantages of the Molecular Recognition Technology process over conventional procedures for these separations are presented and discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Decontamination and Melting of Reactor Upper Internals - 18536

In a BWR, the upper reactor internals, i.e. the steam separators and the steam dryer converts the hot water to a dry steam for the turbines. These components have due to the distance to the reactor core low to very low levels of neutron activation, if any. The contamination of these components is usually significant and consists of both loose contamination and contamination bound in oxide layers. The upper reactor internals on a BWR are quite voluminous in relation to their weight, has a complex geometry and are therefore time consuming to cut in pieces that allows high density packing in waste containers for ILW. The incentives for decontamination of the components down to levels that allow melting for volume reduction are therefore high. A performed full-scale treatment project on a steam dryer confirms the significant volume reduction. The steam dryer was decayed prior to segmentation and melting for volume reduction. Cyclife/EDF has, since significant volume reduction is within reach if the inventory of gamma emitting nuclides and by then the dose rate can be reduced, taken benefit of chemical decontamination studies and an analyze of potential treatment results. The analyze confirm that high decontamination factors can be achieved and that the acceptance criteria for melting can be met, with margin. Full scale decontamination results achieved by AREVA NP and others do align fairly well with the test results. Even without the option of clearance/recycling of the metals, due to high residual activity, a high or very high volume reduction factors can be achieved by melting as well as a potential reclassification of the metal in a disposal perspective. One key aspect for success is to keep the volume of the secondary waste from the decontamination down and to make sure that the waste form fulfils the requirements for disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Structural Changes in Molten Salt Fuel and/or Waste Stream Compounds Cs 2 UCl 6 and Cs 2 UO 2 Cl 4 from Room Temperature to Melting and Related Materials [Slides]

Perovskite phases have been explored in recent literature as waste forms for long term storage of radioactive waste from molten salt reactors specifically. Synthesis of new halide perovskite phases can inform these efforts as well as molten salt reactor chemistry in general. Inorganic perovskites have also shown promise as new scintillating materials for detection of X-rays and gamma rays. Modifying the structure and chemistry of these compounds, perhaps with organic components, could be used to synthesize new scintillating compounds or modify the properties of existing ones. Both U 6+ and U 4+ readily form compounds with a striking similarity to vacancy ordered perovskites (Cs 2 UO 2 Cl 4 and Cs 2 UCl 6 ). Each of these compounds is relevant to molten salt reactor chemistry and/or waste streams. These phases could from in uranium chloride salt systems or waste streams: Cs being a fission product.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Multicomponent pyrochlore solid solutions with uranium incorporation – A new perspective of materials design for nuclear applications

Multicomponent pyrochlore solid solutions with and without uranium incorporation were fabricated and their thermal-mechanical properties were characterized. Multicomponent pyrochlore solid solutions without uranium exhibit comparable thermal conductivity and higher mechanical strength compared to baseline single component rare-earth titanate pyrochlore (A 2 Ti 2 O 7 ). Uranium incorporation reduces hardness as compared with single component compositions. High entropy pyrochlore with uranium displays the highest thermal conductivity within multicomponent pyrochlore solid solutions with significantly better mechanical properties than UO 2 . The measured thermal conductivity correlates well with A-site cation mixing entropy and a modified size disorder parameter, and thus the size disorder and mixing entropy could be good indicators for predicting thermal conductivity of multicomponent pyrochlore solid solutions. This work opens up the possibility of designing multicomponent oxide solid solutions by controlling their chemical disorder/mixing entropy to achieve acceptable thermal-mechanical properties, desired radiation and corrosion performance for potential nuclear waste form and inert matrix fuel applications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Implementation of Increased Salt and Decreased Sludge Volumes in the Defense Waste Processing Facility - 19597

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). The disposition of high-level waste (HLW) into a glass waste form poured into canisters at the DWPF began in 1996. DWPF processed a sludge only flowsheet until 2008 when the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began treating the salt waste from the underground liquid storage tanks by striking with monosodium titanate (MST) to remove actinides, when necessary to meet downstream acceptance criteria, and removing the MST and entrained sludge solids (SS) by filtration before using an organic solvent to strip the cesium from the salt, creating two new concentrated HLW streams. This change caused the DWPF to begin processing under both a sludge-only and a coupled operations flowsheet where salt streams would only be added when available. This strategy allows for continuous glass canister production at DWPF. Recently, DWPF completed a 10-month melter replacement in December 2017 and the Concentrate, Storage, and Transfer Facility (CSTF) has completed repairs to the 3H Evaporator system. These facility outages have caused delays in preparing the next batch of sludge feed for DWPF. Concurrent with these facility outages, the Salt Waste Processing Facility (SWPF) has been undergoing system testing and commissioning, working toward startup. SWPF is designed to process salt waste in the same manner as ARP/MCU, but with much higher throughput volumes produced per unit time. In order to process the SWPF salt streams efficiently, DWPF will implement a coupled operation only flowsheet. Based on the facility outages mentioned above, efforts have been made to identify strategies to minimize sludge volumes and maximize salt volumes in the DWPF until the next sludge batch is ready for processing in DWPF. Savannah River Remediation (SRR) recently performed an evaluation that examines the effects of reducing the volume of sludge and maximizing the addition of the cesium-rich concentrated salt solution in each Sludge Receipt and Adjustment Tank (SRAT) batch in order to extend the existing sludge batch as long as possible until the next sludge batch is ready. These effects include a reduction of weight percent solids in the DWPF processing vessels, an increase in cycle time, a new canister heat generation value that remains within regulatory limits, and higher gamma source that remains within bulk facility gamma shielding. Balancing the above parameters resulted in a small reduction of sludge volume per SRAT batch under SWPF operation and maximizing the salt volumes for both ARP/MCU and SWPF operation. This slight reduction in sludge volume and increase in salt volumes allows SRR to mitigate some of the schedule impact incurred by the facility outages. This strategy was evaluated for both ARP/MCU and SWPF operation. This strategy has been successfully implemented under ARP/MCU operation using small step changes in the current process. This paper provides an overview of the evaluation performed for DWPF with ARP/MCU and DWPF with SWPF operations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

External Engagement During Evaluations of Shutdown Nuclear Power Plant Sites - 19221

This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. The U.S. Department of Energy Office of Integrated Waste Management (DOE-IWM) continues to conduct evaluations of removing spent nuclear fuel (SNF) from shutdown nuclear power plant sites. The 14 shutdown sites included in the evaluation to date have nuclear power reactors that have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The 14 shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun. The Oyster Creek Nuclear Generating Station ceased operations in September 2018 and will be included in future evaluations. A key part of the shutdown site evaluations are visits to the sites. These site visits typically take place over 3 days. The first day is spent at the nuclear power plant site. The second day is spent evaluating the near-site transportation infrastructure, including rail infrastructure, potential heavy-haul routes, and potential rail and barge transload locations. The third day is often spent meeting with community engagement or advisory panels. External engagement activities during site visits are conducted in two areas. The first area involves inviting tribal and state representatives, State Regional Group (SRG) representatives, and Federal Railroad Administration (FRA) representatives to participate in the entire site visit. The SRG representatives are from the Southern States Energy Board, the Western Interstate Energy Board, the Council of State Governments - Midwest, or the Council of State Governments - Eastern Regional Conference, according to the location of the nuclear power plant. State representatives typically represent their state department of energy, state department of environmental or natural resources, State Rail Safety Participation Program, state police, state department of transportation, radiation protection organization, or emergency management organization. Tribal representatives explore issues associated with cultural affiliation and tribal involvement with past and present site activities. As the regulator of the U.S. railroads, the FRA representatives bring unique experience to the site visits. They also coordinate meetings with the railroads that serve the shutdown sites. The second area of external engagement activities during site visits involves meeting with local community engagement or advisory panels. Meeting with local community engagement or advisory panels provides an opportunity to inform these panels on DOE activities and the roles and responsibilities of federal agencies during site decommissioning. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NDA Challenges during Cleanup of the Hanford 618-10 Burial Ground Trenches - 15542

The 618-10 Burial Ground on the Hanford Reservation was an early trench waste site used for the disposal of waste from the 300 Area at Hanford. Much of this waste consisted of residues from evaluations of failed fuel as well as various process and separations studies associated with the Department of Energy (DOE) mission which were performed in the 300 Area between March 1954 and September 1963. Much of this waste contained significant concentrations of fission products as well as Transuranic (TRU) radionuclides. Waste forms consist of drums and boxes which were disposed of in the trenches. Many of the drums are shielded containers, cemented sludge, or cemented debris. There are no identifying marks on the containers to correlate them to a specific waste stream. In addition there is some waste in the trenches which came from other locations outside of the 300 area. Cleanup of the 618-10 burial ground is covered under the Washington Closure Hanford (WCH) contract. Due to the proximity of the 618-10 site to the Columbia River, waste is being removed from the 618-10 site, stabilized, and disposed of into either the Environmental Restoration Disposal Facility (ERDF) if the containers are determined to be LLW or transferred to the Hanford Central Waste Complex (CWC) if the containers are suspected to be TRU waste. In order to determine the correct disposal path Non Destructive Assay (NDA) must be performed on all of the containers excavated from the trenches. The NDA of these containers is performed using germanium gamma spectroscopy with an In Situ Object Counting System (ISOCS) and with a neutron slab counter. Due to the challenges associated with these difficult waste streams, matrices, and lack of information to associate these containers with a particular process, the NDA program has had to develop a more robust process to ensure the correct disposition of the waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Reconfirmation of frit 803 based on the January 2016 sludge batch 9 reprojection

On January 11, 2016, Savannah River Remediation (SRR) provided the Savannah River National Laboratory (SRNL) with a Sludge Batch 9 (SB9) reprojection that was developed from the analyzed composition of a Tank 51 sample. This sample was collected after field washing had been completed in Tank 51 to support the alternate reductant task. Based on this reprojection, Frit 803 is still a viable option for the processing of SB9 under sludge-only operations and coupled (Actinide Removal Process (ARP) product with and without monosodium titanate (MST)) operations. The maximum projected volumes of ARP product that can be transferred from the Precipitate Reactor Feed Tank (PRFT) per Sludge Receipt and Adjustment Tank (SRAT) batch and the resulting Na 2 O concentrations in the SRAT for coupled operations were determined. The Na 2 O concentrations in the SRAT resulting from the maximum projected ARP product transfer volumes are consistent with those from the previous assessments that were based on the August 2015 projections. Regardless of the presence or absence of MST in the ARP product, the contribution of Na 2 O to the resulting glass will be similar at the same waste loading (WL). These projected volumes of ARP product are not anticipated to be an issue for SB9. The actual transfer volumes from the PRFT to the SRAT are determined based upon the analyzed Na 2 O concentrations in the PRFT samples, which has resulted in larger transfer volumes than those allowed by the projections for Sludge Batch 8 (SB8). An operating window of 32-40% WL around the nominal WL of 36% is achievable for both sludge-only and coupled operations; however, each of the glass systems studied does become limited by waste form affecting constraints (durability) at higher volumes of ARP product and WLs of 41-42%.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗