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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Experimental Challenges and Successes in Measuring Aerosol Concentrations at Prototypic Spray Conditions Encountered at the Hanford Waste Treatment and Immobilization Plant - 13327

To date, majority of the work done on measuring aerosol releases from failure of process piping was done using simple Newtonian fluids and small engineered-nozzles that do not accurately represent the fluids and breaches postulated during accident analysis at the Hanford Waste Treatment and Immobilization Plant (WTP). In addition, the majority of the work conducted in this area relies on in-spray measurements that neglect the effect of splatter and do not yield any information regarding aerosol generation rates from this additional mechanism. In order to estimate aerosol generation rates as well as reduce the uncertainties in estimating the aerosol release fractions over a broad range of breaches, fluid properties and operating conditions encountered at the WTP, the Pacific Northwest National Laboratory (PNNL) has designed, commissioned, and tested two experimental test stands. The first test stand, referred to as the large-scale test stand, was designed specifically to measure aerosol concentrations and release fractions under prototypic conditions of flow and pressure for a range of breaches postulated in the hazard analysis for 0.076 m (3-inch) process pipes. However, the size of the large-scale test stand, anticipated fluid loss during a breach, experimental risks, and costs associated with hazardous chemical simulant testing limited the large-scale test stand utility to water and a few non-hazardous physical simulants that did not fully span the particle size and rheological properties of the fluids encountered at the WTP. Overcoming these limitations and extending the range of simulants used, required designing and building a smaller test stand, which was installed and operated in a fume hood. This paper presents some of the features of both test stands, the experimental challenges encountered, and successes in measuring aerosol concentration in both test stands over a range of test conditions. (authors)

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DOE OSTI2018

Cementation of Nuclear Wastes: From Past to Future in a Sustainable Approach - 18266

Cementation of nuclear wastes is a widespread technology due to several factors such as excellent feedback from past experience, simplicity of implementation and versatility. Indeed, as nuclear wastes come in many physical states (bulk wastes, powders, sludges), cementation is the immobilisation process that can easily accommodate this large range of wastes provided the appropriate mixing technology is used: heterogenous cementation (grouting of bulky wastes) or homogenous cementation (intermixing process between the powder or the sludge and the cementitious material). However, nuclear wastes equally present an important chemical variability. This is especially true as New Areva's range of nuclear activities covers the entire fuel cycle: mining, conversion, enrichment and recycling and dismantling and decommissioning. The chemical interactions between the waste and the cement matrix are carefully studied by New Areva's R and D teams in order to provide comprehensive solutions by developing cement formulations which are adapted to wastes characteristics. The aim of the studies is to obtain cement waste-forms that meet the waste acceptance criteria, the final product being thus in perfect adequacy with the disposal requirements. For over 30 years New Areva has been developing cement formulations adapted to the physical and chemical specificities of the wastes accounting for constrains imposed by waste characteristics. The overall approach includes the definition of the appropriate cement composition at laboratory scale, engineering design, equipment manufacturing, testing and commissioning of the cementation facilities. With this strong technical background, a specific methodology has been developed by New Areva teams in order to allow a rapid screening of possible solutions for a given waste (in terms of cement chemistry adapted to the waste) in a learn-fast approach that is based on theoretical analysis and small-scale tests. The methodology was applied for wastes presenting a complex nature (such as reactive metals, ion exchange resins, oils and solvents, mercury, etc), known to interact with Portland cement from a chemical point of view and leading to durability issues of the wasteform (swelling, strength reduction, leaching) and thus reducing the final waste loading. Once the screening process is over, the main drivers for choosing the most appropriate solution to be further developed at pilot scale are: reducing the overall volume and the impact of the final waste package, ensure the waste package performance as defined by the disposal requirements and control the long-term behaviour in final disposal. Moreover, a modelling approach is addressed, aiming to develop a thermo-chemo-mechanical (THM) concrete models that are able to predict the behaviour of the structure (waste form) in time. The aim of the present paper is to describe the main principles within the developed methodology and illustrate it by using a few specific examples. Several distinct cement formulations are developed for complex wastes, limiting the chemical interaction between the cement matrix and the given wastes, obtaining better waste loadings and better overall durability behaviour which facilitates long term behaviour demonstration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Developing a Signature Based Safeguards Approach for the Electrorefiner and Salt Cleanup Unit Operations in Pyroprocessing Facilities

Traditional nuclear materials accounting does not work well for safeguards when applied to pyroprocessing. Alternate methods such as Signature Based Safeguards (SBS) are being investigated. The goal of SBS is real-time/near-real-time detection of anomalous events in the pyroprocessing facility as they could indicate loss of special nuclear material. In high-throughput reprocessing facilities, metric tons of separated material are processed that must be accounted for. Even with very low uncertainties of accountancy measurements (<0.1%) the uncertainty of the material balances is still greater than the desired level. Novel contributions of this work are as follows: (1) significant enhancement of SBS development for the salt cleanup process by creating a new gas sparging process model, selecting sensors to monitor normal operation, identifying safeguards-significant off-normal scenarios, and simulating those off-normal events and generating sensor output; (2) further enhancement of SBS development for the electrorefiner by simulating off-normal events caused by changes in salt concentration and identifying which conditions lead to Pu and Cm not tracking throughout the rest of the system; and (3) new contribution in applying statistical techniques to analyze the signatures gained from these two models to help draw real-time conclusions on anomalous events.

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

Effects of Blending and In-Tank Pretreatment on Direct Feed Hanford High Level Waste Vitrification - 18434

A direct feed high level waste (DFHLW) option is under consideration for early operations of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). In this option, the Pretreatment Facility would be bypassed in order to support an earlier start-up of the HLW Vitrification Facility. For HLW, this would mean that ultrafiltration and caustic leaching that would otherwise have been performed either would not be performed or would be replaced by interim pretreatment functions such as in-tank washings. These changes could result in higher aluminum contents, higher chromium contents, higher fractions of supernate, and lower solids contents in the HLW feed. As a result, glass formulations and downstream vitrification operations will be impacted. In the present work, we investigate the effects of various waste blending and in-tank pretreatment scenarios on DFHLW glass waste loadings, canister counts, and processing rates. Eleven source tanks and five waste groups were selected for the studies of waste blending. In addition, the five waste groups selected were the subjects of an in-tank pretreatment study. Glass property-composition models and melter test data were used to calculate waste loadings, canister counts, processing rates, and treatment durations for a range of DFHLW scenarios. The processing scenarios were defined by the types of waste blending and in-tank pretreatment that were assumed. Blending cases were evaluated that ranged between the 'No Blend' and 'Total Blend' extremes and included pairwise blending (blending of two source tanks, either completely or in various proportions) and macro-batches formed from various combinations of the source tanks. Over 36,000 possible pairwise blending scenarios were evaluated and an optimal retrieval/blending sequence was identified. The waste loading of this optimal sequence is within 0.03 wt% of the 'Total Blend' case. Further, a total of 38 retrieval sequences were identified that have waste loadings within 0.1 wt% of the 'Total Blend' case. To evaluate the effects of in-tank pretreatment, an in-tank wash/settle/decant process was assumed and tank-specific wash factors were employed. Washing of the macro-batch blends gave a general increase in HLW loading, decrease in canister count, and decrease in glass yield. Glass yields are towards the low end of the values expected for the WTP, which could lead to lower than nominal glass production rates. The waste loading increases and canister count reductions with washing were small for two of the wastes indicating that direct treatment of the unwashed waste may be practical. For wastes that are relatively low in sulfate and halides, excessive washing is counterproductive. The results highlight the importance of maintaining high solids content in the HLW feed to maintain glass production rates. The results also illustrate that the optimum washing strategy will need to be evaluated on a tank-to-tank basis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Synroc Waste-forms for Immobilization of Advanced Reprocessing Wastes - 16167

EURO-GANEX, an advanced reprocessing technology, aims to separate both major (U, Pu) and minor (Np, Am, and Cm) actinides together from fission products. As the final waste composition is free from actinides - as opposed to current process waste, e.g. from PUREX - adapted immobilization matrices need to be developed. Synroc is a potential wasteform that has proven itself to be efficient to immobilize high-level nuclear wastes (HLW). In this study, a new composition of Synroc, 'Synroc-Z', is designed and characterized. The primary modification is to reduce the amount of zirconolite phase, which acts as the main host for actinides. The optimum hot-press conditions for Synroc-Z samples with a waste loading of 20 wt% were determined to be: 1200 deg. C and 20 MPa for 3 h. The durability of Synroc-Z was examined via the MCC-1 test at 90 deg. C. It shows similar leach resistance to Synroc-C, which is much better than current glass waste-forms. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Crystallization in high level waste (HLW) glass melters: Savannah River Site operational experience

This paper provides a review of the scaled melter testing that was completed for design input to the Defense Waste Processing Facility (DWPF) melter. Testing with prototype melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by refractory corrosion versus spinels that precipitated from the HLW glass melt pool. A review of the crystallization observed with the prototype melters and the full-scale DWPF melters (DWPF Melter 1 and DWPF Melter 2) is included. Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for a waste treatment and immobilization plant.

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DOE OSTI2018

The Strategic Benefit of Self Shielded Boxes for Storage and Disposal of Metallic Uranic Material-18141

Sellafield Ltd, on behalf of the Nuclear Decommissioning Authority (NDA), is responsible for the management of a number of facilities whose inventories include a range of metallic uranic material. Additional fuel and fuel bearing materials exist in the NDA estate, some of which are similar to the Sellafield-managed inventory and others present different management challenges. Whilst UK Government policy directs that some of these materials are assets and are therefore managed with eventual future use in mind, there remains a large quantity of material, principally either out-of-scope for reprocessing or which derives from legacy facilities, which is likely to be disposed of to a future Geological Disposal Facility (GDF). The majority of this material is stored under wet conditions to facilitate dose shielding and heat dissipation and will therefore require removal from wet storage, consolidation and conditioning, prior to transfer to GDF. In order to accelerate high hazard and risk reduction by sooner retrieval of legacy material from wetted storage, one of the legacy storage facilities has led technical work to develop the design of Self Shielded Boxes (SSBs) and to substantiate their use for interim storage of the materials in question. Use of the SSBs would represent a change of strategy on the Sellafield Site: historically, timescales for retrieval of material have been reliant upon a waste treatment and storage/disposal route being available. SSBs provide the opportunity to retrieve uranic materials from aging legacy storage facilities for immediate buffer storage, by decoupling the timing of retrieval from the availability of an onward route. An understanding was required of the scope for defining optimal routes for each metallic uranic fuel and fuel bearing material group for safe transport and storage commensurate with current final disposal assumptions and risks, taking into account international best practice. A study is underway to understand options for transfer of material into SSBs, buffer/interim storage locations, treatment/conditioning processes and transport to the GDF. This work forms the basis of a prioritised road-map for the development of routes for study material groups aligned with ongoing and existing planned technical work, which is a pre-requisite to strategy development. The ultimate output of the study will be identification of full life cycle management strategies for metallic uranic material from an enterprise perspective. The combination of this strategic work and ongoing technical work will underpin a series of decisions to divert sub-groups of uranic material away from currently assumed baseline routes in support of earliest practicable hazard and risk reduction. Hence, each decision will reduce the scope of a placeholder future uranic fuel treatment capability, which will diminish the associated British Pound 1 bn capital cost. To date, a technical case for interim storage of uranic material in SSBs has been substantiated for around two thirds of the study materials. The requirement for a new strategy at Sellafield was driven by technical development of SSBs in response to the needs of one programme, but this prompted the study into the potential utilisation of SSBs in order to derive benefit for the wider enterprise and across the NDA estate. (authors)

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DOE OSTI2017

Separable Organic Evaluation: Waste Feed Delivery Impacts - 17563

The potential for separable organic materials to separate out from Hanford tank waste requires consideration for future Waste Feed Delivery operations. Separable organic materials are organic compounds of very limited solubility in bulk waste that can form a separate phase or layer. If the organic materials bound to waste separates during mixing and transfer operations, the feed batch would be disqualified under the Hanford Tank Waste Treatment and Immobilization Plant (WTP) waste acceptance criteria for 'no visible immiscible layer.' Sample data and process history were used to identify tanks with the potential to contain separable organic materials. Centrifuge testing was performed on available archived tank samples and on waste simulants developed to mimic characteristics of the archived tank samples. A separable layer formed during centrifuge testing of archived samples 241-C-204, 241-SY-102, and for all developed simulants tested. Due to the formation of a separable organic layer during the centrifugation studies, an impact assessment was performed. This impact assessment identified forty-four potential tanks of concern regarding the release of organic material bound to solids during waste disturbing activities. A Hanford Tank Waste Operations Simulator model run was modified to trace the origin of each batch to the WTP. The outputs of the model run indicated that a separable organic layer would form during High Level Waste Feed Delivery operations and would cascade to all tanks containing or receiving significant amounts of organic solvents/diluents. Hence, there is a potential for a ripple effect across the feed batches. (authors)

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

Focus on Different Binders for Concentrated Alkali and Magnesium Hydroxide Waste Conditioning - 18122

As part of a nuclear dismantling process, an effluent highly concentrated in alkali hydroxides and magnesium hydroxide could be produced and should be conditioned in a matrix capable of ensuring the adequacy between the hydraulic binder performances and the conditioning requirements. First, several matrices were chosen according to the compatibility of their interstitial pore water pH with our highly concentrated in alkali and magnesium hydroxide waste. Then, we observed their behavior from a macroscopic point of view when an increasing amount of alkali hydroxide is added in the mixing water. The compressive strength - an important requirement for the storage acceptance - along with some microstructural comprehension allowed us to discriminate the different binders. Only one was chose to continue the study with the integration of magnesium hydroxide in the formula. Finally, a validation of the early ages performance acquired without magnesium hydroxide was performed and a durability study is undergoing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

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

Development and Testing of a High Level Waste Slurry Sampling Technique to Support Hanford Waste Processing - 17164

The Hanford Tank Operations Contractor must demonstrate the capability to adequately sample staged high-level waste feed. The sampler employed must have accuracy and precision performance that allow the Tank Operations Contractor (TOC) to meet the Waste Treatment and Immobilization Plant (WTP) Waste Acceptance Criteria Data Quality Objectives (DQO). A full-scale sampling loop was used at a test facility to develop and demonstrate sampler capability. This paper will discuss the modifications to a sampler configuration previously tested - modifications focused on the improvement of the accuracy of test results. This work targets one of the remaining technical issues with the high-level waste treatment mission at Hanford. The sampling method employed must support both Tank Operations and Waste Treatment and Immobilization Plant requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the TOC's Chief Technology Office in conjunction with the One System Integrated Project Team. Results of the testing showed that performance of the Isolok{sup R} sampler, chosen for implementation, can be greatly improved by modifying the systems configuration to provide accurate, repeatable results for Hanford's simulated high level waste. (authors)

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DOE OSTI2016

Characterization of U(VI)-Bearing Precipitates Produced by Ammonia Gas Injection Technology for Unsaturated Sediments - 16600

The processing of irradiated fuel to extract plutonium for the nuclear arsenal has resulted in a legacy of uranium contamination in the vadose zone (VZ) of the Hanford 200 Area. This contamination is a potential source for groundwater contamination, posing a risk to receptors through water uptake from contaminated wells or discharges to surface water. The injection of reactive gases is an innovative technology being evaluated for reducing the mobility of radionuclide contaminants in the Hanford vadose zone. Ammonia (NH{sub 3}) gas in particular has been used to reduce uranium mobility in laboratory scale studies and is being evaluated for field scale application. The sediment of the Hanford vadose zone is composed of a complex of mineral phases including quartz, feldspar, mica, and calcium carbonates. The silica content in the 200 Area subsurface markedly outweighs the majority of elements, with aluminum following closely. The injection of ammonia gas into the vadose zone is associated with an increase in pH conditions as the gas partitions into the pore water. The resulting alkaline conditions significantly increase the solubility of some aluminum and silicate-containing minerals within the system. As pH decreases to natural conditions, the aluminosilicate minerals will precipitate, potentially coating and sequestering the uranium contaminants. Although previous laboratory scale tests have shown decreased uranium mobility, there are many questions regarding the application of the remediation method to 200 Area vadose zone conditions. In particular, additional testing is required to understand the impact of various environmental factors on the formation of uranium mineral phases with the application of the NH{sub 3} injection method. In order to answer some of the unknowns regarding this remediation technology for the 200 Area vadose zone, an effort was made to characterize the uranium-bearing precipitates being formed with the injection of ammonia gas. Rather than a sediment-water system, a synthetic pore water solution was prepared using the ion concentrations relevant to the 200 Area vadose zone. These solutions were limited to select constituents believed to play a major role in the remediation process. The use of this carefully composed solution allows for the thorough evaluation of the contribution the variable components make to the formation of the solid analyte. The characterization of solid samples was attempted using a host of complementary analytical techniques including scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS) and X-ray diffraction (XRD). Though characterization of the samples is ongoing, the data produced thus far has allowed for the imaging and X-ray spectroscopy of the solid uranium phases produced as well as a tentative identification of cejkaite (Na{sub 4}(UO{sub 2})(CO{sub 3}){sub 3}) as the major uranium mineral phase. (authors)

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DOE OSTI2013

Distribution of Radioactive Materials in the Absheron Peninsula, Azerbaijan - 13567

The Absheron Peninsula forms the extreme Eastern part of Azerbaijan and juts into the Caspian Sea. The region has a long history of oil and gas exploration, transport, and processing and includes a number of abandoned chemical plants that were used in the separation of iodine from formation waters. As a result of lax environmental standards during the Soviet era, the industrial activity has led to serious contamination from oils residues, heavy metals and naturally occurring radioactive materials (NORM). Radiometric surveys performed over a wide range of the Absheron Peninsula showed generally low NORM concentrations. However, radiation levels two to three orders of magnitude above background levels were detected at two abandoned iodine separation plants near the capital city, Baku. These elevated radiation levels are mainly due to Ra-226 and U-238 with lower contributions from Ra-228 and U-235. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Removal of Trace Amounts of Co-60 from NPP Waste Water Using UV-C Irradiation, TiO{sub 2} and Co-Treat Inorganic Ion Exchange Media - 16025

The removal of Co-60 from an aqueous nuclear liquid waste obtained from a BWR NPP was tested using a two-step process. In the first phase the liquid waste was oxidized with ultraviolet irradiation at 254 nm (UV-C irradiation) together with a photo-catalyst, TiO{sub 2} P25 AEROXIDE, to degrade the organic complexing agents present in the solution. During some of the experiments it was noticed that TiO{sub 2} P25 AEROXIDE also removed Co-60 from the solution. In the second phase the UV-C irradiated solution was mixed with inorganic ion exchange media, Co-Treat, to remove the rest of Co-60 from the solution. The effect of pH and contact time with UV-C irradiation were evaluated and both had a noticeable effect on the results. Best results were attained when pH of the initial test solution was lowered to 3.5 ± 0.2 before the UV-C irradiation. From the initial Co-60 activity, 2.2 Bq/ml, the activity decreased to 0.03 ± 0.01 Bq/ml after 24 hours of UV-C irradiation together with the photo-catalyst and mixing with Co-Treat. (authors)

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

Innovative Process for Comprehensive Treatment of Liquid Radioactive Waste - 12551

This paper presents the results of research activities aimed at creation of a principally new LRW distilling treatment method. The new process is based on the instantaneous evaporation method widely used in distillation units. The main difference of the proposed process is that the vapor condensation is conducted without using heat exchangers in practically ideal mode by way of direct contacting in a vapor-liquid system. This process is conducted in a specially designed ejector unit in supersonic mode. Further recuperation of excess heat of vaporization is carried out in a standard heat exchanger. Such an arrangement of the process, together with use of the barometric height principle, allows to carry out LRW evaporation under low temperatures, which enables to use excess heat from NPS for heating initial LRW. Thermal calculations and model experiments have revealed that, in this case, the expenditure of energy for LRW treatment by distilling will not exceed 3 kilowatt-hour/m{sup 3}, which is comparable with the reverse-osmosis desalination method. Besides, the proposed devices are 4 to 5 times less metal-intensive than standard evaporation units. These devices are also characterized by versatility. Experiments have revealed that the new method can be used for evaporation of practically any types of LRW, including those containing a considerable amount of oil products. Owing to arrangement of the evaporation process at low temperatures, the new devices are not sensitive to 'scale formation'. This is why, they can be used for concentrating brines of up to 500-600 g/l. New types of such evaporating devices can be required both for LRW treatment processes at nuclear-power plants under design and for treating 'non-standard' LRW with complex physicochemical and radionuclide composition resulting from the disaster at the Fukushima I Nuclear Power Plant.) As a result of accidents at nuclear energy objects, as it has recently happened at NPP 'Fukushima-1', personnel faces the necessity to take emergency measures and to use marine water for cooling of reactor zone in contravention of the technological regulations. In these cases significant amount of liquid radioactive wastes of complex physicochemical composition is being generated, the purification of which by traditional methods is close to impossible. According to the practice of elimination of the accident after-effects at NPP 'Fukushima' there are still no technical means for the efficient purification of liquid radioactive wastes of complex composition like marine water from radionuclides. Therefore development of state-of-the-art highly efficient facilities capable of fast and safe purification of big amounts of liquid radioactive wastes of complex physicochemical composition from radionuclides turns to be utterly topical problem. Cesium radionuclides, being extremely dangerous for the environment, present over 90% of total radioactivity contained in liquid radioactive wastes left as a result of accidents at nuclear power objects. For the purpose of radiation accidents aftereffects liquidation VNIIHT proposes to create a plant for LRW reprocessing, consisting of 4 major technological modules: Module of LRW pretreatment to remove mechanical and organic impurities including oil products; Module of sorption purification of LWR by means of selective inorganic sorbents; Module of reverse osmotic purification and desalination; Module of deep evaporation of LRW concentrates. The first free modules are based on completed technological and designing concepts implemented by VNIIHT in the framework of LLRW Project in the period of 2000-2001 in Russia for comprehensive treatment of LWR of atomic fleet. These industrial plants proved to be highly efficient and secure during their long operation life. Module of deep evaporation is a new technological development. It will ensure conduction of evaporation and purification of LRW of different physicochemical composition, including those containing hardness salts, resulted in generation of LRW concentrate 300-600 g/l. The method is based on utilization of supersonic ejector for intensification of thermal physic processes and performance of evaporation in brine recycling mode. All proposed technological solutions are totally based on patented Russian developments. Proposed work will allow to construct modular plants, which will be totally prepared for efficient purification of any types of liquid radioactive wastes from radionuclides in case of force majeure. According to proposed scheme concentration level of cesium radionuclides in safe-for-storage form will make up not less than 5000. With respect to purification from cesium radionuclides of liquid radioactive wastes stored at NPP 'Fukushima' about 10 t of inorganic sorbents, loaded in 160 protective filter-containers, will be required for solving this problem. The amount of secondary wastes will be reduced approximately in 5 times in comparison with traditional schemes, applied in purification of secondary LRW of Fukushima-1 by Areva (France) and Kurion (USA) companies. All units of modular plants will be constructed and manufactured as totally automated, providing their twenty-four-hour safe operation. Modular design will ensure efficiency and let optimize the costs of secondary LRW treatment. In order to ensure off-line operation in emergency conditions the plant should be equipped with auxiliary modules: energy and ventilation ones. Under normal conditions these modules can be stored in 'mothballed' condition at special warehouses under the authority of federal bodies. It will be reasonable to choose required transport facilities, the most suitable for transportation of modules to target destination beforehand, using vessel classification list.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Summary Report of Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

The inadvertent creation of transuranic waste carrying hazardous waste codes D001 and D002 requires the treatment of the material to eliminate the hazardous characteristics and allow its eventual shipment and disposal at the Waste Isolation Pilot Plant (WIPP). This report documents the effectiveness of two treatment methods proposed to stabilize both the unremediated and remediated nitrate salt waste streams (UNS and RNS, respectively). The two technologies include the addition of zeolite (with and without the addition of water as a processing aid) and cementation. Surrogates were developed to evaluate both the solid and liquid fractions expected from parent waste containers, and both the solid and liquid fractions were tested. Both technologies are shown to be effective at eliminating the characteristic of ignitability (D001), and the addition of zeolite was determined to be effective at eliminating corrosivity (D002), with the preferred option1 of zeolite addition currently planned for implementation at the Waste Characterization, Reduction, and Repackaging Facility. During the course of this work, we established the need to evaluate and demonstrate the effectiveness of the proposed remedy for debris material, if required. The evaluation determined that Wypalls absorbed with saturated nitrate salt solutions exhibit the ignitability characteristic (all other expected debris is not classified as ignitable). Follow-on studies will be developed to demonstrate the effectiveness of stabilization for ignitable Wypall debris. Finally, liquid surrogates containing saturated nitrate salts did not exhibit the characteristic of ignitability in their pure form (those neutralized with Kolorsafe and mixed with sWheat did exhibit D001). As a result, additional nitrate salt solutions (those exhibiting the oxidizer characteristic) will be tested to demonstrate the effectiveness of the remedy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Direct Feed Low Activity Waste (DFLAW) Program: Development, Progression and Risk Management - 16476

This Direct Feed Low-Activity Waste (DFLAW) Program paper describes the key programmatic elements to enable the startup of treatment and disposal of Hanford tank waste through a phased startup of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Specifically the DFLAW Program consists of the Low- Activity Waste (LAW) vitrification facility, and the direct feed to LAW of treated tank farm supernatant from the Low-Activity Waste Pretreatment System (LAWPS). This document identifies the overall strategy, objectives, and processes of the U.S. Department of Energy (DOE), Office of River Protection (ORP) and all supporting DOE subcontractors, and describes the scope of the program elements. In addition gaps, risks, and opportunities to successful integration for commissioning of all needed systems and facilities to produce immobilized LAW product (ILAW) are described. The scope of the DFLAW Program extends up to the point that LAW Facility hot commissioning is complete and all DFLAW systems are fully operational. The DFLAW Program coordinates, integrates, and monitors all the activities necessary to successfully commission and start up the WTP LAW facility. This involves integrating the readiness of all the plant, personnel and program/procedures that support startup of hot commissioning of tank farms systems, WTP systems, waste treatment and disposal systems, and site utility, infrastructure and services. Multiple Hanford contractors have responsibilities to execute and perform scope under the umbrella of the DFLAW Program. The primary focus of the One System organization, comprised of WTP Project and Tank Operations Contract (TOC) personnel, is to manage the coordination and integration across the interfaces of the work scope described above. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

DOE Hanford Tank Side Cesium Removal System (TSCR) Update: Fabrication and Factory Acceptance Testing - 20300

The Tank Side Cesium Removal (TSCR) System is a modular, quickly-deployable, and passively safe system for efficiently removing solids and cesium from alkaline liquid waste raffinate created by fuel processing on the Hanford site. This paper provides an overview of the project with particular emphasis on completion of fabrication and factory acceptance testing (FAT). All TSCR System components are inside three enclosures known as the Process Enclosure, Ancillary Enclosure, and Control Enclosure. The Process Enclosure is seismic design category II (SDC-II) and built to withstand the rigors of transportation and lifting, high winds, volcanic ash fall, and heavy precipitation (snow) over a temperature range of -25 to 115 deg. F. The Process Enclosure and Control Enclosure were fabricated in AVANTech's Columbia, SC facility, while the Ancillary Enclosure was fabricated in AVANTech's Richland, WA facility. The design and preliminary documented safety analysis were completed in December 2019, with all work being performed in accordance with ASME NQA-1 quality assurance requirements. The Process Enclosure, Ion Exchange Columns (IXCs), and several other TSCR components have a nuclear safety function and are designated as safety significant structures, systems, and components (SS-SSCs). The NQA-1 Commercial Grade Dedication (CGD) process was used to identify critical characteristics in order to accept materials and items used for fabrication of several safety significant TSCR components. This paper also addresses unique aspects and challenges of implementing CGD in DOE facilities and systems. After fabrication and preliminary testing, the Process and Control Enclosures were shipped to AVANTech's Richland facility where the complete TSCR System was setup and connected in a represented layout and configuration that will be used on the Hanford site. Test tanks and pumps were put in place and connected to the TSCR Process Enclosure in a manner that mimicked interfacing Double Shell Tanks in the 241-AP Tank Farm. The TSCR System was then tested/operated under conditions equivalent to those planned for onsite operation. Testing was separated into five separate phases as defined below: - Phase 1: Confirmation of TSCR System readiness for the FAT; - Phase 2: Verification of TSCR System operation under standard conditions; - Phase 3: Demonstration of TSCR System operation under non-standard conditions; - Phase 4: Demonstration of ion exchange media removal from an IXC; and - Phase 5: Completion of TSCR System lift plans and site mobilization. The Hanford Tank Operations Contractor, Washington River Protection Solutions (WRPS), participated in all phases of testing. Testing at AVANTech's Richland Facility allowed WRPS personnel to familiarize themselves with TSCR operations in a radiologically clean facility prior to onsite mobilization. Engineering and operations personnel gained valuable experience that allowed them to better refine onsite operational protocols and procedures - including handling and movement of IXCs with a forklift. This paper describes the technologies used by the TSCR System as well as how these technologies are deployed and operated to advance the Hanford mission of safely, efficiently, and effectively treating tank waste. It also provides detailed information on fabrication and FAT, including operational data from simulated tank waste runs. A project schedule for WRPS' planned construction acceptance testing, operational acceptance testing, and actual tank waste treatment is provided. This presentation should be of great interest to parties responsible for the design, testing, and mobilization of modular NQA-1 technologies for the treatment of raffinate and similar liquid wastes containing high concentrations of radioactive cesium. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗