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DOE OSTI2013

Extended Sleeve Products Allow Control and Monitoring of Process Fluid Flows Inside Shielding, Behind Walls and Beneath Floors - 13041

Throughout power generation, delivery and waste remediation, the ability to control process streams in difficult or impossible locations becomes increasingly necessary as the complexity of processes increases. Example applications include radioactive environments, inside concrete installations, buried in dirt, or inside a shielded or insulated pipe. In these situations, it is necessary to implement innovative solutions to tackle such issues as valve maintenance, valve control from remote locations, equipment cleaning in hazardous environments, and flow stream analysis. The Extended Sleeve family of products provides a scalable solution to tackle some of the most challenging applications in hazardous environments which require flow stream control and monitoring. The Extended Sleeve family of products is defined in three groups: Extended Sleeve (ESV), Extended Bonnet (EBV) and Instrument Enclosure (IE). Each of the products provides a variation on the same requirements: to provide access to the internals of a valve, or to monitor the fluid passing through the pipeline through shielding around the process pipe. The shielding can be as simple as a grout filled pipe covering a process pipe or as complex as a concrete deck protecting a room in which the valves and pipes pass through at varying elevations. Extended Sleeves are available between roughly 30 inches and 18 feet of distance between the pipeline centerline and the top of the surface to which it mounts. The Extended Sleeve provides features such as ± 1.5 inches of adjustment between the pipeline and deck location, internal flush capabilities, automatic alignment of the internal components during assembly and integrated actuator mounting pads. The Extended Bonnet is a shorter fixed height version of the Extended Sleeve which has a removable deck flange to facilitate installation through walls, and is delivered fully assembled. The Instrument Enclosure utilizes many of the same components as an Extended Sleeve, yet allows the installation of process monitoring instruments, such as a turbidity meter to be placed in the flow stream. The basis of the design is a valve body, which, rather than having a directly mounted bonnet has lengths of concentric pipe added, which move the bonnet away from the valve body. The pipe is conceptually similar to an oil field well, with the various strings of casing, and tubing installed. Each concentric pipe provides a required function, such as the outermost pipes, the valve sleeve and penetration sleeve, which provide structural support to the deck flange. For plug valve based designs, the next inner pipe provides compression on the environmental seals at the top of the body to bonnet joint, followed by the innermost pipe which provides rotation of the plug, in the same manner as an extended stem. Ball valve ESVs have an additional pipe to provide compressive loading on the stem packing. Due to the availability of standard pipe grades and weights, the product can be configured to fit a wide array of valve sizes, and application lengths, with current designs as short as seven inches and as tall as 18 feet. Central to the design is the requirement for no special tools or downhole tools to remove parts or configure the product. Off the shelf wrenches, sockets or other hand tools are all that is required. Compared to other products historically available, this design offers a lightweight option, which, while not as rigidly stiff, can deflect compliantly under extreme seismic loading, rather than break. Application conditions vary widely, as the base product is 316 and 304 stainless steel, but utilizes 17-4PH, and other allows as needed based on the temperature range and mechanical requirements. Existing designs are installed in applications as hot as 1400 deg. F, at low pressure, and separately in highly radioactive environments. The selection of plug versus ball valve, metal versus soft seats, and the material of the seals and seats is all dependent on the application requirements. The design of the Extended Sleeve family of products provides a platform which solves a variety of accessibility problems associated with controlling process flow streams in remote, hard to reach locations in harsh environments. Installation of the equipment described has shown to allow access to flow streams that otherwise would require exceptional means to access and control. The Extended Sleeve family of products provides a scalable solution to both control and monitor process fluid flow through shielding, walls or floors when direct connection is advantageous. (authors)

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DOE OSTI2023

Machine Learning Surrogates of a Fuel Matrix Degradation Process Model for Performance Assessment of a Nuclear Waste Repository

Spent nuclear fuel repository simulations are currently not able to incorporate detailed fuel matrix degradation (FMD) process models due to their computational cost, especially when large numbers of waste packages breach. The current paper uses machine learning to develop artificial neural network and k-nearest neighbor regression surrogate models that approximate the detailed FMD process model while being computationally much faster to evaluate. Further, using fuel cask temperature, dose rate, and the environmental concentrations of CO 3 2- , O 2 , Fe 2+ , and H 2 as inputs, these surrogates show good agreement with the FMD process model predictions of the UO 2 degradation rate for conditions within the range of the training data. A demonstration in a full-scale shale repository reference case simulation shows that the incorporation of the surrogate models captures local and temporal environmental effects on fuel degradation rates while retaining good computational efficiency.

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DOE OSTI2013

Development of an Automated Remedy Performance Evaluation Program - 13622

Performance monitoring is a vital element of groundwater remediation. Unfortunately, the enormous efforts and costs that are expended procuring, managing, processing and storing monitoring data are often not subject to correspondingly rigorous evaluation. This is despite the fact that many steps in the process are predictable and are repeated many times over the remedy life cycle. At the Department of Energy (DOE) Hanford Site, a program is underway to automate many of these steps - processing, formatting any analyzing large volumes of heterogeneous data associated with the operation of several groundwater pump-and-treat (P and T) and in-situ remedies. The Automated Remedy Performance Evaluation Program (ARPEP) was developed from a work-flow process designed to deliver (a) monthly data summaries and preliminary analysis, (b) quarterly performance assessments, and (c) annual roll-up analyses that detect changes in long-term monitoring datasets and support remedy optimization. The intent of the ARPEP is to provide detailed, systematic and traceable data summaries, depictions and analyses that can be used by project scientists to complete their evaluation of remedy performance. The ARPEP work-flow was formalized following extensive review of applicable guidance, regulation and industry standards. The ARPEP incorporates disparate data types collected over different frequencies, such as water levels and pumping rates recorded every minute, and groundwater sample results obtained on quarterly, annual or irregular intervals. The data are processed, reduced to frequencies suitable for assessment, and combined in various ways leading to performance indicators such as (a) pumped well capacities and system downtime that reflect operational performance; (b) hydraulic gradients and areas of hydraulic containment that reflect hydraulic performance; and (c) time-series (longitudinal) and geo-statistical (spatial) trend analyses that reflect progress toward attainment of Remedial Action Objectives (RAOs). The entire sequence of data processing and analysis is undertaken within the R programming environment, an open-source environment for statistical computing and graphics, and is summarized in a comprehensive PDF project report. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Summary of FY11 Sulfate Retention Studies for Defense Waste Processing Facility Glass

This report describes the results of studies related to the incorporation of sulfate in high level waste (HLW) borosilicate glass produced at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). A group of simulated HLW glasses produced for earlier sulfate retention studies was selected for full chemical composition measurements to determine whether there is any clear link between composition and sulfate retention over the compositional region evaluated. In addition, the viscosity of several glasses was measured to support future efforts in modeling sulfate solubility as a function of predicted viscosity. The intent of these studies was to develop a better understanding of sulfate retention in borosilicate HLW glass to allow for higher loadings of sulfate containing waste. Based on the results of these and other studies, the ability to improve sulfate solubility in DWPF borosilicate glasses lies in reducing the connectivity of the glass network structure. This can be achieved, as an example, by increasing the concentration of alkali species in the glass. However, this must be balanced with other effects of reduced network connectivity, such as reduced viscosity, potentially lower chemical durability, and in the case of higher sodium and aluminum concentrations, the propensity for nepheline crystallization. Future DWPF processing is likely to target higher waste loadings and higher sludge sodium concentrations, meaning that alkali concentrations in the glass will already be relatively high. It is therefore unlikely that there will be the ability to target significantly higher total alkali concentrations in the glass solely to support increased sulfate solubility without the increased alkali concentration causing failure of other Product Composition Control System (PCCS) constraints, such as low viscosity and durability. No individual components were found to provide a significant improvement in sulfate retention (i.e., an increase of the magnitude necessary to have a dramatic impact on blending, washing, or waste loading strategies for DWPF) for the glasses studied here. In general, the concentrations of those species that significantly improve sulfate solubility in a borosilicate glass must be added in relatively large concentrations (e.g., 13 to 38 wt % or more of the frit) in order to have a substantial impact. For DWPF, these concentrations would constitute too large of a portion of the frit to be practical. Therefore, it is unlikely that specific additives may be introduced into the DWPF glass via the frit to significantly improve sulfate solubility. The results presented here continue to show that sulfate solubility or retention is a function of individual glass compositions, rather than a property of a broad glass composition region. It would therefore be inappropriate to set a single sulfate concentration limit for a range of DWPF glass compositions. Sulfate concentration limits should continue to be identified and implemented for each sludge batch. The current PCCS limit is 0.4 wt % SO 4 2- in glass, although frit development efforts have led to an increased limit of 0.6 wt % for recent sludge batches. Slightly higher limits (perhaps 0.7-0.8 wt %) may be possible for future sludge batches. An opportunity for allowing a higher sulfate concentration limit at DWPF may lay lie in improving the laboratory experiments used to set this limit. That is, there are several differences between the crucible-scale testing currently used to define a limit for DWPF operation and the actual conditions within the DWPF melter. In particular, no allowance is currently made for sulfur partitioning (volatility versus retention) during melter processing as the sulfate limit is set for a specific sludge batch. A better understanding of the partitioning of sulfur in a bubbled melter operating with a cold cap as well as the impacts of sulfur on the off-gas system may allow a higher sulfate concentration limit to be established for the melter feed. This approach would have to be taken carefully to ensure that a sulfur salt layer is not formed on top of the melt pool while allowing higher sulfur based feeds to be processed through DWPF.

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

International Implementation of IAEA's Borehole Disposal Concept for Sealed Radioactive Sources - 18545

With funding from the Government of Canada, and support from the U.S. Nuclear Regulatory Commission's (NRC's) Office of International Program and many others, the International Atomic Energy Agency (IAEA) is supporting the field deployment of the Borehole Disposal Concept (BDC) for secure disposal of radioactive sources in Ghana, the Philippines and Malaysia. This is a first-of-its-kind implementation. Though small in volume, disused sealed radioactive sources can be intensively radioactive and create a significant safety and security liability for the majority of the 168 Member States in the IAEA. Millions of sealed radioactive sources have been manufactured since 1901, and it is estimated that hundreds of thousands are now unwanted. The irony is that some of the countries with the smallest inventories have the greatest difficulties managing their radioactive wastes. The BDC was originally conceptualized in an IAEA-study undertaken by the South African Nuclear Energy Corporation (NECSA) for countries with small inventories of disused sealed radioactive sources. There are multiple, passive engineered and natural barriers in the BDC design, including an inner stainless-steel capsule (which holds the wastes), an inner cement containment barrier, an outer stainless steel container and outer backfill cement, as well as the surrounding geosphere. In common subsurface geochemical environments, these barriers will isolate the wastes from the biosphere, provide effective containment and ensure long-term safety. To greatly limit the likelihood of inadvertent human intrusion (safety), and deliberate human intrusion (security), these waste packages will be disposed of in a narrow diameter borehole (26 cm) at depths greater than 30 m. This type of borehole disposal creates an exceptionally small footprint, removes the wastes from normal human surficial activities, and isolates the waste packages from near-surface processes. Over the past 20 years, the IAEA, NECSA and many others have advanced the BDC on the shoulders of the IAEA's broader work supporting safe and secure disposal of radioactive wastes and management of sealed radioactive sources. With recent funding and support from many others, the IAEA has assisted Ghana, Malaysia and the Philippines in assessing their inventories and conditioning their disused sealed radioactive sources, in drilling investigation boreholes and characterizing their proposed sites, and in designing their disposal systems. These countries plan to place the uppermost waste package at depths greater than 100 m. While activities in the Philippines remain focused on-site characterization, Ghana is in the latter stages of developing their Safety Case for operational and 'post-closure' safety and Malaysia has completed their Safety Case and submitted a license application to construct and operate a BDC facility. Though this paper identifies the support provided to Member States by the IAEA, it is the Ghana Atomic Energy Commission and the Malaysian Nuclear Agency that are responsible for the safety and security of the disposals and that have been responsible for carrying-out their disposal programs. Once the BDC is licensed by the relevant national regulatory bodies, these first-of-a-kind disposal approaches will provide a template for other countries to safely dispose of their disused sealed radioactive sources - permanently eliminating their safety and security liabilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

In-Situ Radiological Surveys to Address Nuclear Criticality Safety Requirements During Remediation Activities at the Shallow Land Disposal Area, Armstrong County, Pennsylvania - 12268

Cabrera Services Inc. (CABRERA) is the remedial contractor for the Shallow Land Disposal Area (SLDA) Site in Armstrong County Pennsylvania, a United States (US) Army Corps of Engineers - Buffalo District (USACE) contract. The remediation is being completed under the USACE's Formerly Utilized Sites Remedial Action Program (FUSRAP) which was established to identify, investigate, and clean up or control sites previously used by the Atomic Energy Commission (AEC) and its predecessor, the Manhattan Engineer District (MED). As part of the management of the FUSRAP, the USACE is overseeing investigation and remediation of radiological contamination at the SLDA Site in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), 42 US Code (USC), Section 9601 et. seq, as amended and, the National Oil and Hazardous Substance Pollution Contingency Plan (NCP), Title 40 of the Code of Federal Regulations (CFR) Section 300.430(f) (2). The objective of this project is to clean up radioactive waste at SLDA. The radioactive waste contains special nuclear material (SNM), primarily U-235, in 10 burial trenches, Cabrera duties include processing, packaging and transporting the waste to an offsite disposal facility in accordance with the selected remedial alternative as defined in the Final Record of Decision (USACE, 2007). Of particular importance during the remediation is the need to address nuclear criticality safety (NCS) controls for the safe exhumation and management of waste containing fissile materials. The partnership between Cabrera Services, Inc. and Measutronics Corporation led to the development of a valuable survey tool and operating procedure that are essential components of the SLDA Criticality Safety and Material Control and Accountability programs. Using proven existing technologies in the design and manufacture of the Mobile Survey Cart, the continued deployment of the Cart will allow for an efficient and reliable methodology to allow for the safe exhumation of the Special Nuclear Material in existing SLDA trenches. (authors)

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DOE OSTI2012

The multiple market-exposure of waste management companies: A case study of two Swedish municipally owned companies

Highlights: Black-Right-Pointing-Pointer Swedish municipally owned waste management companies are active on political, material, technical, and commercial markets. Black-Right-Pointing-Pointer These markets differ in kind and their demands follow different logics. Black-Right-Pointing-Pointer These markets affect the public service, processing, and marketing of Swedish waste management. Black-Right-Pointing-Pointer Articulating these markets is a strategic challenge for Swedish municipally owned waste management. - Abstract: This paper describes how the business model of two leading Swedish municipally owned solid waste management companies exposes them to four different but related markets: a political market in which their legitimacy as an organization is determined; a waste-as-material market that determines their access to waste as a process input; a technical market in which these companies choose what waste processing technique to use; and a commercial market in which they market their products. Each of these markets has a logic of its own. Managing these logics and articulating the interrelationships between these markets is a key strategic challenge for these companies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Removal of Radiocesium from Food by Processing: Data Collected after the Fukushima Daiichi Nuclear Power Plant Accident - 13167

Removal of radiocesium from food by processing is of great concern following the accident of TEPCO's Fukushima Daiichi Nuclear Power Plant accident. Foods in markets are monitored and recent monitoring results have shown that almost all food materials were under the standard limit concentration levels for radiocesium (Cs-134+137), that is, 100 Bq kg{sup -1} in raw foods, 50 Bq kg{sup -1} in baby foods, and 10 Bq kg{sup -1} in drinking water; those food materials above the limit cannot be sold. However, one of the most frequently asked questions from the public is how much radiocesium in food would be removed by processing. Hence, information about radioactivity removal by processing of food crops native to Japan is actively sought by consumers. In this study, the food processing retention factor, F{sub r}, which is expressed as total activity in processed food divided by total activity in raw food, is reported for various types of corps. For white rice at a typical polishing yield of 90-92% from brown rice, the F{sub r} value range was 0.42-0.47. For leafy vegetable (indirect contamination), the average F{sub r} values were 0.92 (range: 0.27-1.2) after washing and 0.55 (range: 0.22-0.93) after washing and boiling. The data for some fruits are also reported. (authors)

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

The OECD-NEA RK and M Initiative (2011-2018) - a 7-year step in a 10,000-year journey: What next? - 19432

The OECD Nuclear Energy Agency (NEA) has coordinated an international initiative to develop understanding and guidelines related to the long-term preservation of records, knowledge and memory (RK and M) for radioactive waste repositories. After seven years of work, it has come to a close in 2018. Currently, the 'Final Report of the RK and M Initiative' is in the final process of publication [the status/sentence will be updated in early 2019]. It includes a historical review, sets out the RK and M fundamentals, presents a menu of approaches and mechanisms of RK and M preservation, and finally provides suggestions on how to develop a systemic strategy aimed at embedding radioactive waste repositories in society over extended periods of time. This paper highlights the 'multi-dimensionality' of RK and M preservation: Multiple time frames need to be considered, multiple transmission modes explored, and multiple actors brought together. The latter seems to be the primary aspect to be considered in the discussion about future (international) cooperation in the RK and M preservation field. (authors)

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

Germany's First Steps in the Restarted Site Selection Process - 18015

After the Fukushima event in 2011 Germany has decided to phase out nuclear power. Subsequently, it seemed to be politically attractive also to achieve a broad consensus with respect to final disposal of the high-active heat-generating waste after a long societal conflict lasting for decades. Germany has then decided to start a new site selection program for the determination of a site for high-active and heat-generating waste. In 2013, an overall parliamentary consensus could be achieved to agree on a new Repository Site Selection Act. This law has then been evaluated by a commission and the Repository Site Selection Act became amended. Besides, also organization and responsibilities became adapted as well as the financing system. A fund has been installed, which was financed by the utilities, to secure the availability of money to finance the whole process. In the meanwhile, the organizations for licensing of interim storage and final disposal facilities (BfE), for the operation of interim storage facilities for spent fuel, high-active waste and low and intermediate level waste (BGZ) and for site investigation, construction, operation and decommissioning of repositories for spent fuel and high-active waste as well as for low and intermediate level waste (BGE) had been founded and the site investigation process has just started. (authors)

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

Challenges in Interim and Final Storage of HLW in Germany - 19060

After the Fukushima event in 2011 Germany has decided to phase out nuclear power. Subsequently, it seemed to be politically attractive also to achieve a broad consensus with respect to final disposal of the high-active heat-generating waste after a long societal conflict lasting for decades. Germany has then decided to start a new site selection program for the determination of a site for high-active and heat-generating waste. In 2013, an overall parliamentary consensus could be achieved to agree on a new Repository Site Selection Act. This law has then been evaluated by a commission and the Repository Site Selection Act became amended. Besides, also organization and responsibilities became adapted as well as the financing system. A fund has been installed, which was financed by the utilities, to secure the availability of money to finance the whole process. In the meanwhile, the organizations for licensing of interim storage and final disposal facilities (BfE), for the operation of interim storage facilities for spent fuel, high-active waste and low and intermediate level waste (BGZ) and for site investigation, construction, operation and decommissioning of repositories for spent fuel and high-active waste as well as for low and intermediate level waste (BGE) had been founded and the site investigation process has just started. (authors)

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DOE OSTI2016

Free-standing Liquids in a Resin Disposal Container - How Unidentified Legacy Issues and Decisions Can Affect Current Radwaste Processing Activities - 16605

On Monday, October 27, 2014 a radwaste and transportation specialist was informed that a condensate filter demineralizer (CFD) radwaste disposal container exceeded the radioactive waste disposal facility free-standing liquid requirement for acceptance for disposal (limit is 0.5% of waste volume or 5.6 gallons); 8.5 gallons was measured). As a result, on October 28, the plant was informed that our disposal site use permit was suspended, pending a point-of-origin inspection by the agreement state regulatory authority. This paper details the investigation conducted to identify causes, especially legacy issues one may not necessarily identify, and documents corrective actions taken to assure similar issues are not repeated. Actions taken to prevent recurrence have been successful. The five containers shipped, including 14-033, were verified dewatered and dried to comply with the disposal facility waste acceptance criteria. Although discoloration is viewed in the dewatering sight glass, milky color has not been viewed and the dewatering process has continued. Isolation of the lowest set of dewatering filter laterals for use at the end of the dewatering process has worked well and use of the dewatering verification filters has provided evidence that each CRF resin liner was successfully dewatered and dried. To date, 37 CFD liners have been successfully dewatered, dried, verified, transported to the disposal facility, and buried without incident. If similar issues are encountered, you should keep an open mind regarding causes including, but not limited to discussions with your vendor and conduct a search of past procedure revisions to gain additional insight regarding program/procedure changes and how they could have an effect on issues, which occur today.

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

Gas Retention, Gas Release, and Fluidization of Spherical Resorcinol-Formaldehyde (sRF) Ion Exchange Resin

The Low-Activity Waste Pretreatment System (LAWPS) is being developed to provide treated supernatant liquid from the Hanford tank farms directly to the Low-Activity Waste (LAW) Vitrification Facility at the Hanford Tank Waste Treatment and Immobilization Plant. The design and development of the LAWPS is being conducted by Washington River Protection Solutions, LLC. A key process in LAWPS is the removal of radioactive Cs from LAW liquids using ion exchange (IX) columns filled with spherical resorcinol-formaldehyde (sRF) resin. When loaded with radioactive Cs, radiolysis of water in the LAW liquid will generate hydrogen gas. In normal operations, the generated hydrogen is expected to remain dissolved in the liquid and be continuously removed by liquid flow. One accident scenario being evaluated is the loss of liquid flow through the sRF resin bed after it has been loaded with radioactive Cs and hydrogen gas is being generated by radiolysis. For an accident scenario with a loss of flow, hydrogen gas can be retained within the IX column both in the sRF resin bed and below the bottom screen that supports the resin within the column, which creates a hydrogen flammability hazard. Because there is a potential for a large fraction of the retained hydrogen to be released over a short duration as a gas release event, there is a need to quantify the size and rate of potential gas release events. Due to the potential for a large, rapid gas release event, an evaluation of mitigation methods to eliminate the hydrogen hazard is also needed. One method being considered for mitigating the hydrogen hazard during a loss of flow accident is to have a secondary flow system, with two redundant pumps operating in series, that re-circulates liquid upwards through the bed and into a vented break tank where hydrogen gas is released from the liquid and removed by venting the headspace of the break tank. The mechanism for inducing release of gas from the sRF bed is to fluidize the bed, which should allow retained bubbles to rise and be carried to the break tank. Currently, neither the ability of fluidizing a settled sRF bed to release gas nor the quantity of hydrogen gas that can be retained and the corresponding potential for large, rapid gas release events has been tested and evaluated. Accordingly, the purpose of this study was to conduct experiments and develop models that evaluate the ability of a re-circulating flow system to fluidize an sRF resin bed as a method to release retained hydrogen gas and maintain hydrogen gas retention at low levels, thereby avoiding conditions where hydrogen gas poses a flammability hazard. The purpose of the study was also to determine if effective hydrogen gas release would occur using a single constant fluidization velocity, because this is the most simple safety system to implement, for the range of fluid and sRF resin conditions anticipated in the column (because a velocity that is effective for one fluid/resin pair can be ineffective for different resin/fluid pairs). Finally, the study also included quantifying the volume of gas that can be retained in an sRF bed and the behavior of spontaneous, rather than fluidization-induced, gas release events.

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

Mitigation of Radioactive Waste By Sampling and Analyzing The Affected Area and Efficient Use of Different Detectors - 17400

The paper summarizes the way by which the mitigation of the radioactive waste during events like decommissioning of nuclear power plants, or during a nuclear disaster can be done so that the rehabilitation process for flora and fauna can begin in a natural manner. The paper describes the problems faced during ER operations in the developing countries and need of international committees in these areas. The paper suggests marking of the radioactive areas which can be used for controlled personnel access and which can be useful for monitoring purposes, while also focusing on the need of proper knowledge while handling detectors in ER operations for successful site characterization. It also highlights the importance of bio-indicators in these operations and also gives a description of various detectors used widely and their related significance for ER operations. The paper gives an overview of the planning and site characterization stages that follow in the environmental remediation. Planning stage should consider: 1. The impact on the ecosystem and human beings and the risk of spreading of the radionuclides. 2. The financial aid in hand for detectors, machinery and skilled labor. As the planning stages develop, the very first thing is the survey of the site. Site investigation should be carried out in all aspects including meteorology, geology, and hydrology, seismicity, geomorphology and the other aspects depending on site for a certain period of time. A site characterization should take: 1. The amount of damage occurred to the environment in the nearby region. 2. The distribution, character and the extent to which radioactivity has been dispersed. 3. The amount of risk that will be involved for personnel working in the field. 4. The precision of data collected that is to be used as input for further processes. In the developing countries, the central governmental bodies play a pivotal role in providing funds internally or from international bodies. The paper shows the effect how political issues effect the work progress of ER operations, that hamper the process and make people frustrated, who really work on the field. Paper provides the suggestion for the proper use of detectors in these countries where surveyors lack knowledge on what kind of detectors to be used while taking precision as a parameter. The table beside gives just a small picture of the large variety of detectors available today and tries to summarizes the main detectors with applications. There should also be a proper marking of the areas into different levels of radiation that will be helpful for work access control and monitoring personnel health. The paper also suggests of using bio-indicators in ER operation more often. The paper proposes to use site-specific bio-indicators which are very helpful in knowing the proper effect of the event in the environment. There is no single paved path for every situation that can lead to a successful ER. The planning should be done in such a way that the future work is not disturbed due to technical problems. In developing countries, political issues can hamper the progress of work and time remains a critical issue in such areas. The lack of proper knowledge for the use of detectors can affect the outcome of site characterization. Areas of the survey should be marked with different levels of activity for efficient monitoring of personnel's. The use of bio-indicators should be used more often and should be chosen as per the site. As they are a helpful tool in evaluating the work progress. (authors)

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

Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Degradation of Dome Cutting Minerals in Hanford Waste - 13100

At the Hanford Tank Farms, recent changes in retrieval technology require cutting new risers in several single-shell tanks. The Hanford Tank Farm Operator is using water jet technology with abrasive silicate minerals such as garnet or olivine to cut through the concrete and rebar dome. The abrasiveness of these minerals, which become part of the high-level waste stream, may enhance the erosion of waste processing equipment. However, garnet and olivine are not thermodynamically stable in Hanford waste, slowly degrading over time. How likely these materials are to dissolve completely in the waste before the waste is processed in the Waste Treatment and Immobilization Plant can be evaluated using theoretical analysis for olivine and collected direct experimental evidence for garnet. Based on an extensive literature study, a large number of primary silicates decompose into sodalite and cancrinite when exposed to Hanford waste. Given sufficient time, the sodalite also degrades into cancrinite. Even though cancrinite has not been directly added to any Hanford tanks during process times, it is the most common silicate observed in current Hanford waste. By analogy, olivine and garnet are expected to ultimately also decompose into cancrinite. Garnet used in a concrete cutting demonstration was immersed in a simulated supernate representing the estimated composition of the liquid retrieving waste from Hanford tank 241-C-107 at both ambient and elevated temperatures. This simulant was amended with extra NaOH to determine if adding caustic would help enhance the degradation rate of garnet. The results showed that the garnet degradation rate was highest at the highest NaOH concentration and temperature. At the end of 12 weeks, however, the garnet grains were mostly intact, even when immersed in 2 molar NaOH at 80 deg. C. Cancrinite was identified as the degradation product on the surface of the garnet grains. In the case of olivine, the rate of degradation in the high-pH regimes of a waste tank is expected to depend on two main parameters: carbonate is expected to slow olivine degradation rates, whereas hydroxide is expected to enhance olivine dissolution rates. Which of these two competing dissolution drivers will have a larger impact on the dissolution rate in the specific environment of a waste tank is currently not identifiable. In general, cancrinite is much smaller and less hard than either olivine or garnet, so would be expected to be less erosive to processing equipment. Complete degradation of either garnet or olivine prior to being processed at the Waste Treatment and Immobilization Plant cannot be confirmed, however. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Proton nuclear magnetic resonance (1H NMR) of flammable organic chemicals in radioactive high–level supernatant waste at the Savannah River Site (SRS)

The Savannah River Site stores approximately 36 million gallons of radioactive and hazardous waste that contains approximately 245 million curies. The waste is sent through various chemical processes to reduce its volume and to separate various components. The facility plans to replace formic acid (a chemical used to reduce soluble mercury) with glycolic acid. Recycle solution with glycolate may flow back to the tank farm, where the glycolate can generate hydrogen gas by thermal and radiolytic mechanisms. The current analytical method for detecting glycolate (ion chromatography) in supernatant requires a large dilution to reduce interference from the nitrate anions. Hydrogen nuclear magnetic resonance is an analytical method that requires less sample dilution. It takes advantage of the CH 2 group in glycolate. Liquid samples were spiked with four different levels of glycolate to build a calibration line, as it is recommended in the standard addition method. The detection and quantitation limits determined were 1 and 5 ppm, respectively, for 32 scans, which is well below the process limit of 10 ppm. In one test, 800 scans of a supernatant spiked with 1 ppm glycolate resulted in a -CH 2 peak with a signal-to-noise ratio of 36.

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