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Systematic Analysis of Drivers and Barriers of Cross-Sectorial Learning Between Nuclear and Oil and Gas Decommissioning Projects - 20139

Project and megaproject management in the energy sector have historically focused on the planning and delivery of the construction of infrastructure. Therefore, policies are predominantly oriented to support the construction of infrastructure rather than its decommissioning. However, globally, several facilities have reached (or will soon reach) their end-of-life and need to be decommissioned. These facilities span across the energy sector, including nuclear power plants and oil and gas rigs, whose decommissioning present unprecedented techno-socio-economic challenges. Moreover, globally, their decommissioning cost is estimated to reach hundreds of billions of US dollars and this budget keeps increasing, with limited knowledge-transfer across industries to mitigate the spiraling increase of this figure. Cross-sectorial knowledge-transfer is a way to tackle this matter and improve the planning and delivery of decommissioning and waste management projects. The aim of this research is to build a road-map based on cross-sectorial industrial experience, designed to promote the sharing of good practices between projects both within and across industries dealing with major decommissioning and waste management challenges. To reach this aim, the tacit knowledge of senior practitioners on several large scale nuclear and oil and gas decommissioning projects is made explicit by the means of semi-structured interviews, systematically analyzed through directed content analysis. The results of this analysis include a list of good practices that could and should be transferred to nuclear decommissioning projects, including the need to promote the adoption of incentives explicitly linked to the project deliverables, and to develop long-term plans to support smooth transition from operations to decommissioning. Suggestions about how the knowledge-transfer can be promoted within and across organizations are also discussed. In this way, this research provides practical guidance for project managers and policymakers on how to support the planning and delivery of projects with improved performance. (authors)

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Graphite waste classification and disposal cost estimation for high temperature gas and salt reactors

As high-temperature reactor designs progress to demonstration, managing the radioactive wastes from these systems presents unique challenges. This work explores the irradiated graphite source term produced by three reactor designs: The Modular High Temperature Gas reactor (MHTGR), a pebble-bed High Temperature Gas Reactor (pb-HTGR), and a Fluoride-cooled High-temperature Reactor (FHR). We predicted a C-14 concentration of 4.3 Ci/m 3 for the MHTGR, 1.2 Ci/m 3 for the pebble bed HTGR, and 2.5 Ci/m 3 for the gFHR after 20 years of operation. The final C-14 concentration highly depended on the graphite nitrogen impurity, a major precursor for C-14. The C-14 concentration in all reactor types exceeded the 0.8 Ci/m3 threshold, resulting in a Class C waste classification. The costs associated with accepting the graphite after 20 years in a low-level waste disposal facility were projected to be 255 dollars per kWe for the MHTGR, 248 dollars per kWe for the pb-HTGR, and 56.8 dollars per kWe for the FHR.

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Geographic Information System Based Emergency Response Training Assessments for DOE Radioactive Materials Transport - 20027

Safety and security are priorities of U.S. Department of Energy (DOE) radioactive materials shipping campaigns. In the more than 70-year history of domestic transport of spent nuclear fuel (SNF), there has never been a transportation-related radiological injury. To support transportation planning, among the tools that DOE uses is the Stakeholder Tool for Assessing Radioactive Transportation (START). START contains geospatial data and transportation route analyses capabilities designed to support a range of DOE transportation planning initiatives. One of those functions is the capability to support emergency response planning and training for State and Tribal jurisdictions located on routes used for DOE shipments of radioactive materials. As part of the Department's commitment to public safety, DOE provides federally-funded radiological response training to emergency responders along DOE radioactive materials transportation corridors through its Transportation Emergency Preparedness Program (TEPP). START contains spatial data representing the locations and emergency-response capabilities of fire departments, police, hospitals, State emergency response centers, and where TEPP-trained personnel are based. The START tool supports State and Tribal users' ability to evaluate emergency-response coverage on active and potential DOE radioactive materials transport routes through their jurisdictions, provide expected response times to reach the scene of an incident, identify equipment available to support a response, and identify the number of response personnel and their respective training levels. In addition, START can be used to identify gaps in coverage along a transportation corridor where additional radiological emergency response training may be needed. This paper describes the data, features, and functionality DOE uses to provide a resource for emergency response training needs assessments for States and Tribes along active and potential routes for transporting radioactive materials, and illustrates its use. It also discusses future plans to integrate TEPP and Federal Emergency Management Agency (FEMA) radiological training data to provide a more comprehensive source of geospatial information on personnel who have received equivalent radiological response training. (authors)

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Development of Property Composition Models For RPP-WTP LAW Glasses (Final Report)

This report describes the development of property-composition models for low-activity waste (LAW) glasses for the River Protection Project Waste Treatment Plant (RPP-WTP) at the Hanford site. The RPP-WTP will separate Hanford tank wastes into LAW and high-level waste (HLW) streams and each stream will be vitrified separately. The development of LAW and HLW glass formulations for the RPP-WTP has been reported previously and is an ongoing activity. Acceptable formulations must meet a variety of processability, product quality, and waste loading requirements that are dictated either by the RPP-WTP contract or by the characteristics of the particular treatment processes that have been selected. These requirements amount to constraints on the acceptable ranges of certain glass properties. These properties are determined first and foremost by the composition of the glass or glass melt. Thus, while there is no direct way of controlling the glass properties of interest during production, there are simple and extremely effective methods of achieving the same result by instead controlling the glass composition. This basic principle is no different from that used to produce enormous volumes of commercial glass to meet exacting product specifications. An essential difference in waste vitrification, however, is that one of the raw materials (the waste itself) can be subject to considerable compositional uncertainty and variability. Thus, waste vitrification facilities and associated process control systems (of which, the operating envelope in glass formulation space is a key part) must, of course, be designed to be robust with respect to such variations. The determination of quantitative relationships between the glass properties that must be controlled and the glass composition can play an important role in the development of such facilities.

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Blending as an In-Tank Waste Processing Strategy: Year 1 Progress Report

The Hanford Site stores approximately 56 million gallons of radioactive legacy defense waste in underground tanks. Direct feed of southeast quadrant tank waste to a high-level waste treatment facility is currently being explored to accelerate tank treatment activities. This study seeks to de-risk waste transfers by favorably altering the resuspension and critical velocities of wastes through strategic waste blending. Wastes rich in aluminum- and zirconium-bearing solids pose significant risks due to their high settled strength and density. A series of tests explores how solid phase particle size and suspending phase chemistry alter waste transport properties, quantified through the just-suspended mixing speed (NJS), for pure and blended aluminum- and zirconium-bearing waste solid simulants. The study utilizes a range of gibbsites (three sizes spanning 10 to 90 µm) and zirconia solids (three sizes spanning 10 to 100 µm). As expected, the NJS of pure solids largely follows predictions based on literature correlations; blends of gibbsite and zirconium-bearing solids of differing particle sizes exhibit NJS maxima attributed to dense packing of the mixed-size settled solids. The findings of this study will assist in developing strategies to optimize the transport of waste solids for the safe and efficient delivery of tank waste.

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The Use of Silver Chloride Injection in Remediation of Iodine-129 by In Situ Capture as Silver Iodide at the F-Area Seepage Basin - 20225

The Savannah River Site (SRS) produced tritium, plutonium, and special nuclear materials for national defense, medicine, and the space programs. As part of operations, the F-Area Seepage Basins operated until 1988 for the disposition of deionized acidic wastewater from the F Separations Facility. The wastewater contained dilute nitric acid and low concentrations of non-radioactive metals, and radionuclides, with the major isotopes being Cs-137, Sr-90, U-235, U-238, Pu-239, Tc-99, I-129, and tritium. The seepage basins were closed in 1988 and backfilled and capped by 1991. The groundwater emanating from beneath the closed and capped seepage basins is acidic and contains elevated levels of both chemical and radiological contaminants. Releases from the groundwater plumes sourced from the F-Area Seepage Basins have impacted the water quality of Fourmile Branch, which is a small tributary to the Savannah River, a regional water source. A large pump-and-treat system was constructed in 1997 and operated until 2003 in an attempt to capture the releases to Fourmile Branch. The system in F Area and a similar system in H Area were expensive (∼$1.3 M/month) to operate and produced large quantities of radioactive waste, with concentrations of I- 129 too high to be disposed of at the SRS. In 2004, SRS replaced pump-and-treat with a funnel and gate system that along with operation of a base injection system at the gates reduces the flux of contaminants to the wetlands adjacent to Fourmile Branch. The alkaline solution injected into the aquifer neutralizes the acidic plume and immobilizes many of the cationic constituents. However, base injection is not effective in managing the release of iodine-129, an anionic contaminant. To address iodine-129 SRS and the Savannah River National Laboratory developed an in situ technology that uses ultra-fine ground silver chloride (AgCl) as an injectable capture medium for the sequestration of iodine-129. The AgCl amendment has a very small particle size and is designed to be injected into the contaminated aquifer to capture iodine-129. Dissolved iodine-129 forms a stable and highly insoluble solid (silver iodide) upon contact with AgCl. Laboratory studies, a field scale pilot test (2009), and three deployments (2011, 2015 and 2019) of AgCl have been successfully performed at the F-Area Seepage Basins. (authors)

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Numerical simulation of coupled THM behaviour of full-scale EBS in backfilled experimental gallery in the Horonobe URL

Bentonite-based engineered barrier system (EBS) is a key component of many repository designs for the geological disposal of high-level radioactive waste. Given the complexity and interaction of the phenomena affecting the barrier system, coupled thermo-hydro-mechanical (THM) numerical analyses are a potentially useful tool for a better understanding of their behaviour. In this context, a Task (the Horonobe EBS experiment) was undertaken to study, using numerical analyses, the thermo-hydro-mechanical (and thermo-hydro) interactions in bentonite based engineered barriers within the international cooperative project DECOVALEX 2023. One full-scale in-situ experiment and four laboratory experiments, largely complementary, were selected for modelling. The Horonobe EBS experiment is a temperature-controlled non-isothermal experiment combined with artificial groundwater injection. The Horonobe EBS experiment consists of the heating and cooling phases. Six research teams performed the THM or TH (depended on research team approach) numerical analyses using a variety of computer codes, formulations and constitutive laws. Finally, for each experiment, the basic features of the analyses are described and the comparison between calculations and laboratory experiments and field observations are presented and discussed.

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Improbability of Post-Closure Criticality in Array of Transuranic Waste Containers after Compaction by Salt Creep at Waste Isolation Pilot Plant

Based on the rationale presented, post-closure nuclear criticality is improbable when room closure compacts containers disposing transuranic (TRU) waste emplaced at the Waste Isolation Pilot Plant (WIPP), an operating repository in bedded salt in southeastern New Mexico. As described in the original WIPP certification, a qualitative estimate of the probability of post-closure criticality in TRU waste produced from atomic energy defense activities has been low either because remote-handled TRU waste canisters are neutronically isolated by the bedded salt or because the low fissile mass in an array of contact-handled TRU waste drums cannot be compacted sufficiently by room closure from salt creep. These situations are still valid for the majority of TRU waste that is disposed at WIPP without any disposal constraints, as updated herein. This report also qualitatively evaluates the probability of criticality after disposal of TRU waste in pipe overpack containers (POC) where every POC in a shipment may have the maximum 200 fissile gram equivalent of 239Pu content. The probability of criticality for a disposal room filled with POCs is estimated during four representative phases of repository evolution: (1) a large salt block falls onto POCs in the first 20 years, (2) salt creep closes a disposal room in the first 1000 years without brine seepage and subsequent gas generation, which permits maximum compaction, (3) some brine seepage occurs into the closed room, which initiates consumption of the fiberboard (cellulose) impact absorber in the POC in the second 1000 years, and (4) full brine inundation of a room and consumption of all fiberboard thereafter. Salt-block fall in the first phase does not greatly disrupt three tiers of POCs. The compacted spacing of POCs in the later three repository conditions is calculated through high-fidelity, geomechanical modeling. Criticality evaluation of compacted 200-g 239 Pu spheres at the compacted spacing shows that neither 12-inch nor 6-inch POCs are critical after the first 1000 years, the second 1000 years, or thereafter as the sea of reflector material changes to represent the three repository conditions. Specifically, fiberboard and iron isolates 239 Pu while dry, and brine reduces the reactivity when a room is partially and fully inundated. Because POC behavior bounds behavior of other standard TRU waste containers, post-closure criticality caused by room closure compacting containers is omitted in the performance assessments for the 2019 and 2026 WIPP compliance re-certification applications to the US Environmental Protection Agency.

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Improbability of Post-Closure Criticality in Compacted Criticality Control Overpacks after Room Closure at Waste Isolation Pilot Plant

As part of its periodic re-certification of the Waste Isolation Pilot Plant (WIPP), an operating repository in bedded salt for the disposal of transuranic (TRU) waste from atomic energy defense activities, the United States Environmental Protection Agency expects a re-evaluation of features, events, and processes, such as post-closure nuclear criticality. Although salt creep beneficially encapsulates the TRU waste in the closed WIPP repository, the spacing between an array of waste packages is disrupted as the salt creep closes disposal rooms and containers lose structural integrity. For most TRU waste, the possibility of post-closure criticality is exceedingly small either because the salt neutronically isolates TRU waste canisters or because closure of a disposal room from salt creep does not sufficiently compact the low mass of fissile material. The criticality evaluation was updated, however, because of the introduction of criticality control overpack (CCO) containers, which may dispose up to 380 fissile gram equivalent plutonium-239 in each container. The criticality potential is evaluated through high-fidelity geomechanical modeling of a disposal room filled with CCO containers during two representative conditions: (1) large salt block fall, and (2) gradual disposal room closure from salt creep. Geomechanical models of roof fall demonstrate three tiers of CCO containers are not greatly disrupted. Geomechanical models of gradual room closure from salt creep (without brine seepage and subsequent gas generation to permit maximum room closure) were used to predict irregular arrays of closely packed CCOs after 1000 years, when room closure has asymptotically approached maximum compaction. Models of spheres or cylinders with 380 fissile gram equivalent of plutonium (as oxide) at the predicted irregular compacted spacing demonstrate that an array of CCO containers is not critical when surrounded by salt and magnesium oxide, provided the mass of hydrogenous material shipped in CCO containers (usually plastics) is controlled or boron carbide (a neutron poison) is mixed with the fissile contents.

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Removal of Metal Oxides from Reactor Metals/Alloys by Laser Ablation

Storage of low and intermediate level radioactive waste in a deep geological repository (L and ILW DGR): - Large volumes of metallic waste = increased production of hydrogen gas. - Large fraction of the metallic wastes are out-of-core reactor components. - Carbon steel feeder pipes. - Radioactive contamination is concentrated within the surface oxides. - Removal of surface contaminants would lead to volume reduction of metallic waste that is stored in L and ILW DGR. - Current decontamination techniques generate large volumes of secondary waste. Objectives: Oxide removal from carbon steel using a ytterbium fiber laser (1064 nm) with a particular focus on identifying the key laser parameters that affect the removal efficiency. Advantages of laser ablation: Generate low volumes of waste, Flexibility of deployment, Remote operation. Grow an oxide film on carbon steel with the incorporation of cobalt. Optimize laser cleaning technology for the removal of surface oxides. Incorporate an oxide collection system for the ablated particles. Conclusions: Objective 1 - Simulated inactive oxides on carbon steel, using Co{sup 2+} as a surrogate for Co radionuclide. Objective 2 - Optimized laser parameters (200 MW/cm{sup 2}, 3 scans, 15 μm overlap); Complete removal of cobalt from surface and bulk metal with optimized cleaning settings. Objective 3 - Glass fiber media H14 HEPA filters had highest collection yield; High oxide collection yields (∼100%). Future Work: Further testing of the liquid trap will verify the efficacy of the HEPA filter and determine the amount of ablated oxides that pass through. Testing of the laser ablation system needs to be done on other reactor materials along with the incorporation of different transition metal cations. Improvements to the oxide collection system are required to improve collection yields.

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Training for Decommissioning and Waste Management - Competence in the Field of Nuclear Engineering - 20137

Due to the phase-out of nuclear power and the reorganization of the participating organizations in Germany, the demand for qualified personnel has increased. But the number of nuclear study programs has rapidly decreased in Germany, although qualified employees are needed for the upcoming tasks in nuclear decommissioning. The solution for those challenging tasks can only be achieved through a strategic approach, involving all stakeholders. AiNT supports this process by offering a tailor-made Education and Training (E and T) program in cooperation with a wide range of partners and initiating the project BEAST (Building competence - Expert knowledge - Applied techniques - Safe decommissioning - Train fellows) as part of the OECD NEA framework Nuclear Education, Skills and Technology (NEST) program. This paper describes the current situation in Germany, provides insight into the education and training program of AiNT, and presents the project BEAST. We are currently looking for international partner institutions that would like to cooperate with us within the framework of the BEAST project and expand the range of topics. (authors)

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Extraction of Ln(III) and An(III) from Alkaline Solutions by Sulfonamide Ligands and Analogs - 20216

As part of our effort to develop advanced separations involving combined extraction of cesium, strontium, and actinides from High-Level Waste (HLW) at the Savannah River Site, we are now reporting highly efficient extraction of Sm(III) from aqueous media of various alkalinity by dichloromethane solutions of several ligands, including tri-sulfonamides, o-mono-sulfonamido-phenols and o-phenylenediamine- derived disulfonamides. Distribution experiments and DFT calculations provide insights on the complexation of Samarium vs Americium by these ligands. Further optimization of these ligands for extraction into highly lipophilic solvents and application to HLW simulants are underway. (authors)

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Brine Availability Test in Salt: Thermal-Hydrological-Mechanical-Chemical Simulations of a Heated Borehole in Salt - 20239

The Brine Availability Test in Salt (BATS) is a collaboration between Los Alamos, Sandia, and Lawrence Berkeley National Laboratories that is being conducted to reduce the uncertainties associated with disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in a hypothetical bedded salt repository. Bedded salt formations may be an acceptable host rock for disposal of nuclear waste due to salt's high thermal conductivity, extremely low permeability, and self-healing capability. In this paper, we report on the first round of the BATS project, a 'shakedown test' known as Phase 1s, with an emphasis on the supporting numerical modeling. The experiment was conducted at the Waste Isolation Pilot Plant (WIPP) using previously drilled horizontal boreholes. The test included a heater borehole in which temperature, pressure, and moisture was monitored, and two nearby temperature monitoring boreholes. Numerical simulations that reflect the 36-year-long history of activity at the WIPP site are necessary to develop the appropriate initial conditions. Testing with different heater designs reveals that a 750 W radiative heater within a borehole can achieve the 120 deg. C borehole wall temperature goal. (authors)

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Uranium Mill Tailings Radiation Control Act Title II DOE Due Diligence and Lessons Learned from a Previous Site Transfer - 20352

Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) established that a government agency will provide perpetual care for closed uranium and thorium ore-processing sites that were operating under an NRC source material license in 1978 or were licensed thereafter. Commercial owners (licensees) operating under an NRC or agreement state specific license when UMTRCA was passed are responsible for conducting reclamation of any byproduct material remaining from uranium-ore processing operations in accordance with an NRC or agreement state approved reclamation plan. Reclamation includes both surface and groundwater remedies. Upon completion of reclamation and approval by NRC, the site is required to be transferred to either the host state or the DOE for long-term surveillance and maintenance. Since UMTRCA's enactment, six Title II sites have been transferred to DOE; an additional 24 Title II sites are anticipated to be transferred before 2050. DoE's role mandated under UMTRCA Title II as the long-term care custodian is to perform 'monitoring, maintenance, and emergency measures necessary to protect the public health and safety.' UMTRCA requires that the licensee pay a long-term surveillance charge 'sufficient to cover the annual costs of site surveillance.' However, at some sites such as the Bluewater, New Mexico, Disposal Site, this mandate has required additional effort and expense by DOE, beyond the originally anticipated and intended scope within UMTRCA, but within the authority of DOE under UMTRCA. In 1997, the Bluewater site became the second UMTRCA Title II site to be transferred to DOE. The site was the location of a uranium mill operated from 1953 until 1982. The specific licensee began site reclamation in 1991, and by 1995 all tailings and contaminated materials were encapsulated in two tailings disposal cells and other disposal areas. In addition to surface contamination, milling activities impacted groundwater in the two upper aquifers. In 1989, the specific licensee attempted active groundwater remediation; however, no significant reduction in contaminant concentrations was observed. As a result, the specific licensee applied to NRC for alternate concentration limits (ACLs) in 1990, which were approved in 1996 as being protective, after additional corrective actions were performed. Since transfer of the Bluewater site to DOE, unforeseen challenges have occurred, requiring additional actions. The first challenge is the occurrence of surface depressions located on the northern section of the main tailings disposal cell. The depressions were first observed during DoE's initial inspection in 1998; however, evidence of these can be observed on satellite images taken prior to transfer. Since being first observed, the depressions have continued to grow both in depth and areal extent. Due to the design of the main tailings disposal cell, the depressions impede storm water from being effectively shed off the 101- hectare (250-acre) top slope of the main tailings disposal cell. Instead, storm water accumulates in the depressions, forming a large ephemeral pond that has stored up to 16.3 x 10{sup 6} liters (4.3 million gallons) of stormwater. The ponding poses a potential risk to the integrity of the main tailings disposal cell in the case of a large storm event, with the potential to cause the pond to overtop and erode the cover material and underlying waste. DOE has taken a number of short-term actions to monitor, measure, and reduce the ponding and is currently working with the US Army Corps of Engineers to design and construct a repair. Additional challenges are associated with groundwater at the Bluewater site. Nine wells were present on the 1335-hectare (3300-acre) site upon transfer. Groundwater compliance was called into question after the State of New Mexico reduced its uranium groundwater standard from 5.0 to 0.03 milligrams per liter in 2004, and when an ACL for uranium was exceeded in a site monitoring well in 2010. Acquiring historical groundwater data and subsequent evaluations as well as additional DOE groundwater monitoring led to installing 10 new monitoring wells and performing additional site hydrogeology recharacterization. DOE continues to evaluate groundwater conditions at the site and works with the NRC to determine regulatory requirements and a path forward. As a result of lessons learned at the Bluewater site and other Title II sites, improved processes have been implemented at a programmatic level to increase due diligence before site transfer and prevent similar issues from occurring at other UMTRCA Title II sites under long-term management. DoE's due diligence process is documented in the Process for Transition of UMTRCA Title II Disposal Sites to DOE for Long-Term Surveillance and Maintenance and is designed to ensure that DOE has no technical or compliance concerns with regulatory decisions that might compromise protectiveness following site transfer to DOE. Implementation of the due diligence process has increased DoE's role prior to site transfer and has been effective in identifying potential issues. Actions by NRC and specific licensees, in response to enhanced due diligence efforts by DOE, are expected to minimize, if not totally prevent, the need for unanticipated actions by DOE pertaining to the surface and groundwater remedies after site transfer. (authors)

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Technology Development for Dry Storage of Aluminum-Clad Spent Nuclear Fuel - 20490

A candidate disposition pathway for the > 13 MTHM of aluminum-clad spent nuclear fuel (ASNF), owned and managed by the U.S. Department of Energy, is the drying and placement of the SNF into sealed-canister dry storage, with the ASNF-in- canisters 'road-ready' for transportation to and final direct disposal in a repository waste package. Technical information gaps in fuel drying, and fuel dry storage behavior, have previously challenged the declaration of technology readiness for drying and placement of this fuel into the DOE Standard Canister design for > 50 years of safe dry storage. The principal technical information gaps included: i) characterization and thermal dehydration behavior of aluminum (oxy)hydroxide films attendant on the cladding due to film formation during reactor operation and during post-discharge up to long-term wet storage histories; and ii) G-value data to enable estimation of the radiolytic gas generation from the cladding with its (oxy)hydroxide films. Thus, the oxide films on the ASNF challenged the safety of a sealed storage canister with thermal and radiolytic decomposition of the waters on the films that can lead to corrosion, pressurization, and flammability issues. These gaps in the technical information base have largely been closed. This paper discusses the investigations at the Idaho National Laboratory (INL) and the Savannah River National laboratory (SRNL), and outlines the pending technology development work for input to an engineering design to enable a road-ready dry storage system for ASNF. The ASNF inventory considered for road-ready dry storage is stored at the Savannah River Site (SRS) and at the Idaho Nuclear Technology and Engineering Center (INTEC) at the INL. The ASNF inventory in the SRS L Basin is from foreign and domestic research reactors (FRR and DRR), and is diverse in terms of design, irradiation, and post-reactor-discharge storage conditions; these factors yield a range of characteristics of cladding oxide films on ASNF. Mixed aluminum (oxy)hydroxide (boehmite and bayerite/gibbsite) films, non-uniform in thicknesses up to a maximum local thickness bounded by 25 μm, were observed on ASNF materials removed from wet storage in the L Basin and in non-sealed dry storage at the SRS following reactor service and a long-term (up to 40+ years) interim storage history. The ASNF inventory at INTEC, in both wet and vented dry storage, is predominantly from the Advanced Test Reactor (ATR), but it also includes DRR and FRR fuel. To address a profound behavior of these films, radiolysis testing of aluminum specimens with mixed type boehmite/bayerite oxide films was performed to develop basic data on G-values for production of radiolytic hydrogen under dry storage conditions with nominal relative humidity, temperature, and cover gas. Modeling and simulation of canister internal environments with postulated inventories of oxide films provides estimation of the evolution of the conditions of the canister loaded with ASNF. Simultaneous Thermal Analysis (TGA/DSC) of hydrated oxide powders, and laboratory-scale aluminum specimens with a bayerite film (∼10 μm) using TGA methods, inform drying time/temperature conditions to be used for the ASNF. A demonstration project is recommended for Verification and Validation of the drying and storage of the ASNF. Remaining major tasks leading up to the hot demonstration include scale-up radiolysis testing and scale-up drying testing. Engineering design with the information from the technology program will establish the safety basis and enable long-term (> 50 years) dry storage compatibility with ASNF in the DOE Standard Canister pending its transportation to and disposal in a repository. This full capability would show the ASNF-in-canister storage system to be road-ready. (authors)

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Cell G 7606A Mockup Measurement Analysis Methods and Results

The production of research isotopes at the Radiochemical Engineering Development Center (REDC) invariably leads to the accumulation of radioactive waste. Storage of this waste onsite at ORNL serves as an interim solution prior to shipment to long term storage facilities, however onsite capacity is limited. Proper characterization of the activity of waste products is essential for determining the appropriate waste stream and, ultimately, mitigating the cost of disposal. This is typically done via gamma spectrometry and use of the In-Situ Object Counting System (ISOCS), a software package from Mirion that serves as an accepted community standard. However, variability in the contents, density, and activity distribution in waste containers can introduce large errors in ISOCS quantification. A measurement campaign using a mockup of the proposed setup in Cell G seeks to quantify the magnitude and source of these systematic errors. Analysis of this data will guide the creation of ISOCS geometry templates and measurement methods designed to minimize overall uncertainty in reported activities.

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Radioactive Waste Sludge Washing and Demonstration of the Nitric-Glycolic Acid Flowsheet for Sludge Batch 10 Qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch (SB) is processible. During processing of SB9, DWPF will be transitioning from the Nitric-Formic Acid (NFA) flowsheet to the Nitric-Glycolic Acid (NGA) flowsheet. Thus, the qualification of SB10 was requested to only be performed using the NGA flowsheet. In order to qualify the batch for the NGA flowsheet, Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles, designated SC-19, were performed using SB10 Tank 51 sample material. SRNL received Tank 51 material in the midst of Tank Farm washing. SRNL continued the washing in the SRNL Shielded Cells. The SRNL process included the addition of Sodium Reactor Experiment (SRE) material from H Canyon, simulating the transfer of SRE from H Canyon to Tank 51 during Tank Farm washing. The washed SB10 Tank 51 material, with SRE, was characterized prior to flowsheet qualification testing.

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Proton Nuclear Magnetic Resonance (H NMR) of Glycolate in Real Waste: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

In preparation for implementing the NG acid flowsheet for the SRS LWS, analytical methods for determining glycolate at low concentration are desired to support system management and safety. Previous work documented that the IC method performed well for samples with low to moderate ionic strength. That work also included a scoping effort to determine if an alternate analytical strategy using H NMR would complement and extend the capabilities of the IC method. The H NMR scoping results indicated that the method had the potential to expand glycolate analysis in LWS samples to higher ionic strength tanks/solutions. The scope of this work is to develop innovative proton NMR techniques, including demonstrating ion exchange decontamination protocols needed for application of the technique to real waste samples. Part of radioactive waste processing at SRS uses formic acid to reduce oxidized (Hg 2+ ) to more volatile elemental Hg for steam striping, collecting, and disposal. Under acidic conditions found in the Chemical Processing Cell (CPC) at the Defense Waste Processing Facility (DWPF), formic acid has a much higher hydrogen generation rate than an alternative reductant, glycolic acid. Thus, an NG acid flowsheet has been developed utilizing glycolic acid with the benefit of easing the need for headspace monitoring requirements for hydrogen and ammonia at DWPF. Low concentrations of glycolate are conservatively assumed to be in the recycle stream, which will collect in the Recycle Collection Tank (RCT). The DWPF recycle stream collected in the RCT has a distinct pathway to the LWS waste tanks that feed the 2H and 3H Evaporator. This route involves transfer of the DWPF recycle to Tank 22 in the Concentration, Storage, and Transfer Facilities (CSTF) followed by transfer to the LWS tank farm/evaporators.

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