Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “CONCRETE”

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

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

At least 1,387 records · Page 77
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Completion of the Operational Closure of Tank 18F and Tank 19F at the Savannah River Site by Grouting - 13236

Radioactive waste is stored in underground waste tanks at the Savannah River Site (SRS). The low-level fraction of the waste is immobilized in a grout waste form, and the high level fraction is disposed of in a glass waste form. Once the waste is removed, the tanks are prepared for closure. Operational closure of the tanks consists of filling with grout for the purpose of chemically stabilizing residual material, filling the tank void space for long-term structural stability, and discouraging future intrusion. Two of the old-style single-shell tanks at the SRS have received regulatory approval confirming waste removal had been completed, and have been stabilized with grout as part of completing operational closure and removal from service. Consistent with the regulatory framework, two types of grout were used for the filling of Tanks 18F and 19F. Reducing grout was used to fill the entire volume of Tanks 18F and 19F (bulk fill grout) and a more flowable grout was used to fill equipment that was left in the tank (equipment fill grout). The reducing grout was added to the tanks using portable grout pumps filled from concrete trucks, and delivered the grout through slick lines to the center riser of each tank. Filling of the two tanks has been completed, and all equipment has been filled. The final capping of riser penetrations brings the operation closure of Tanks 18F and 19F to completion. (authors)

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

The 100-C-7 Remediation Project. An Overview of One of DOE's Largest Remediation Projects - 13260

The U.S. Department of Energy Richland Operations Office (RL), U.S. Environmental Protection Agency (EPA) and Washington Closure Hanford LLC (WCH) completed remediation of one of the largest waste sites in the U.S. Department of Energy complex. The waste site, 100-C-7, covers approximately 15 football fields and was excavated to a depth of 85 feet (groundwater). The project team removed a total of 2.3 million tons of clean and contaminated soil, concrete debris, and scrap metal. 100-C-7 lies in Hanford's 100 B/C Area, home to historic B and C Reactors. The waste site was excavated in two parts as 100-C-7 and 100-C-7:1. The pair of excavations appear like pit mines. Mining engineers were hired to design their tiered sides, with safety benches every 17 feet and service ramps which allowed equipment access to the bottom of the excavations. The overall cleanup project was conducted over a span of almost 10 years. A variety of site characterization, excavation, load-out and sampling methodologies were employed at various stages of remediation. Alternative technologies were screened and evaluated during the project. A new method for cost effectively treating soils was implemented - resulting in significant cost savings. Additional opportunities for minimizing waste streams and recycling were identified and effectively implemented by the project team. During the final phase of cleanup the project team applied lessons learned throughout the entire project to address the final, remaining source of chromium contamination. The C-7 cleanup now serves as a model for remediating extensive deep zone contamination sites at Hanford. (authors)

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

The Creation of a French Basic Nuclear Installation - Description of the Regulatory Process - 13293

CEA is a French government-funded technological research organization. It has to build a medium-level waste interim storage facility because the geological repository will not be available until 2025. This interim storage facility, called DIADEM, has to be available in 2017. These wastes are coming from the research facilities for spent fuel reprocessing and the dismantling of the most radioactive parts of nuclear facilities. The CEA handles the waste management by inventorying the needs and updating them regularly. The conception of the facility is mainly based on this inventory. It provides quantity and characteristics of wastes and it gives the production schedule until 2035. Beyond mass and volume, main characteristics of these radioactive wastes are chemical nature, radioisotopes, radioactivity, radiation dose, the heat emitted, corrosive or explosive gas production, etc. These characteristics provide information to study the repository safety. DIADEM mainly consists of a concrete cell, isolated from the outside, wherein stainless steel welded containers are stored, stacked in a vertical position in the racks. DIADEM is scheduled to store three types of 8 mm-thick, stainless steel cylindrical containers with an outside diameter 498 mm and height from 620 to 2120 mm. DIADEM will be a basic nuclear installation (INB in French) because of overall activity of radioactive substances stored. The creation of a French basic nuclear installation is subject to authorization according to the French law No. 2006-686 of 13 June 2006 on Transparency and Security in the Nuclear Field. The authorization takes into account the technical and financial capacities of the licensee which must allow him to conduct his project in compliance with these interests, especially to cover the costs of decommissioning the installation and conduct remediation work, and to monitor and maintain its location site or, for radioactive waste disposal installations, to cover the definitive shut-down, maintenance and surveillance expenditure. The authorization is issued by a decree adopted upon advice of the French Nuclear Safety Authority and after a public enquiry. In accordance with Decree No. 2007-1557 of November 2, 2007, the application is filed with the ministries responsible for nuclear safety and the Nuclear Safety Authority. It consists of twelve files and four records information. The favorable opinion of the Nuclear Safety Authority on the folder is required to start the public inquiry. Once the public inquiry is completed, the building permit is issued by the prefect. (authors)

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

Savannah River Site Footprint Reduction Results under the American Recovery and Reinvestment Act - 13302

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, managed and operated by Savannah River Nuclear Solutions. Construction of SRS began in the early 1950's to enhance the nation's nuclear weapons capability. Nuclear weapons material production began in the early 1950's, eventually utilizing five production reactors constructed to support the national defense mission. Past operations have resulted in releases of hazardous constituents and substances to soil and groundwater, resulting in 515 waste sites with contamination exceeding regulatory thresholds. More than 1,000 facilities were constructed onsite with approximately 300 of them considered radiological, nuclear or industrial in nature. In 2003, SRS entered into a Memorandum of Agreement with its regulators to accelerate the cleanup using an Area Completion strategy. The strategy was designed to focus cleanup efforts on the 14 large industrial areas of the site to realize efficiencies of scale in the characterization, assessment, and remediation activities. This strategy focuses on addressing the contaminated surface units and the vadose zone and addressing groundwater plumes subsequently. This approach streamlines characterization and remediation efforts as well as the required regulatory documentation, while enhancing the ability to make large-scale cleanup decisions. In February 2009, Congress approved the American Reinvestment and Recovery Act (ARRA) to create jobs and promote economic recovery. At SRS, ARRA funding was established in part to accelerate the completion of environmental remediation and facility deactivation and decommissioning (D and D). By late 2012, SRS achieved 85 percent footprint reduction utilizing ARRA funding by accelerating and coupling waste unit remediation with D and D of remnant facilities. Facility D and D activities were sequenced and permitted with waste unit remediation activities to streamline regulatory approval and execution. Achieving footprint reduction fulfills the Government's responsibility to address legacy contamination; allows earlier completion of legally enforceable compliance agreement milestones; and enables future potential reuse of DOE resources, including land and infrastructure for other missions. Over the last 3.5 years significant achievements were met that contributed to footprint reduction, including the closure of 41 waste units (including 20 miles of radiologically contaminated stream) and decommissioning of 30 facilities (including the precedent setting in situ closure of two former production reactors, the first in the DOE Complex). Other notable achievements included the removal of over 39,750 cubic meters of debris and 68,810 cubic meters of contaminated soils, including 9175 cubic meters of lead-contaminated soil from a former site small arms testing range and treatment of 1,262 cubic meters of tritium-laden soils and concrete using a thermal treatment system. (authors)

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

Handling Radioactive Waste from the Proton Accelerator Facility at the Paul Scherrer Institut (PSI) - Always Surprising? - 13320

The Paul Scherrer Institut (PSI) is the largest national research centre in Switzerland. Its multidisciplinary research is dedicated to a wide field in natural science and technology as well as particle physics. In this context, PSI is operating, amongst others, a large proton accelerator facility since more than 30 years. In two cyclotrons, protons are accelerated to high speeds and then guided along roughly 100 m of beam line to three different target stations to produce secondary particles like mesons and neutrons for experiments and a separately beam line for UCN. The protons induce spallation processes in the target materials, and also at other beam loss points along the way, with emission of protons, neutrons, hydrogen, tritium, helium, heavier fragments and fission processes. In particular the produced neutrons, due to their large penetration depth, will then interact also with the surrounding materials. These interactions of radiation with matter lead to activation and partly to contamination of machine components and the surrounding infrastructures. Maintenance, operation and decommissioning of installations generate inevitably substantial amounts of radioactive operational and dismantling waste like targets, magnets, collimators, shielding (concrete, steel) and of course secondary waste. To achieve an optimal waste management strategy for interim storage or final disposal, radioactive waste has to be characterized, sorted and treated. This strategy is based on radiation protection demands, raw waste properties (size, material, etc.), and requirements to reduce the volume of waste, mainly for legal and economical reasons. In addition, the radiological limitations for transportation of the waste packages to a future disposal site have to be taken into account, as well as special regulatory demands. The characterization is a task of the waste producer. The conditioning processes and quality checks for radioactive waste packages are part of an accredited waste management process of PSI, especially of the Section Dismantling and Waste Management. Strictly proven and accepted methods needed to be developed and enhanced for safe treatment, transport, conditioning and storage. But in the field of waste from research activities, individual and new solutions have to be found in an increasingly growing administrative environment. Furthermore, a wide variety of components, with a really large inventory of radioactive nuclides, has to be handled. And there are always surprising challenges concerning the unusual materials or the nuclide inventory. In case of the operational and dismantling radioactive accelerator waste, the existing conditioning methods are in the process of a continuous enhancement - technically and administratively. The existing authorized specifications of conditioning processes have to be extended to optimize and fully describe the treatment of the inevitably occurring radioactive waste from the accelerator facility. Additional challenges are the changes with time concerning the legal and regulatory requirements - or do we have to consider it as business as usual? This paper gives an overview of the current practices in radioactive waste management and decommissioning of the existing operational accelerator waste. (authors)

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

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

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

US NRC-Sponsored Research on Stress Corrosion Cracking Susceptibility of Dry Storage Canister Materials in Marine Environments - 13344

At a number of locations in the U.S., spent nuclear fuel (SNF) is maintained at independent spent fuel storage installations (ISFSIs). These ISFSIs, which include operating and decommissioned reactor sites, Department of Energy facilities in Idaho, and others, are licensed by the U.S. Nuclear Regulatory Commission (NRC) under Title 10 of the Code of Federal Regulations, Part 72. The SNF is stored in dry cask storage systems, which most commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack vented to the external atmosphere to allow airflow for cooling. Some ISFSIs are located in marine environments where there may be high concentrations of airborne chloride salts. If salts were to deposit on the canisters via the external vents, a chloride-rich brine could form by deliquescence. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC), particularly in the presence of residual tensile stresses from welding or other fabrication processes. SCC could allow helium to leak out of a canister if the wall is breached or otherwise compromise its structural integrity. There is currently limited understanding of the conditions that will affect the SCC susceptibility of austenitic stainless steel exposed to marine salts. NRC previously conducted a scoping study of this phenomenon, reported in NUREG/CR-7030 in 2010. Given apparent conservatisms and limitations in this study, NRC has sponsored a follow-on research program to more systematically investigate various factors that may affect SCC including temperature, humidity, salt concentration, and stress level. The activities within this research program include: (1) measurement of relative humidity (RH) for deliquescence of sea salt, (2) SCC testing within the range of natural absolute humidity, (3) SCC testing at elevated temperatures, (4) SCC testing at high humidity conditions, and (5) SCC testing with various applied stresses. Results to date indicate that the deliquescence RH for sea salt is close to that of MgCl{sub 2} pure salt. SCC is observed between 35 and 80 deg. C when the ambient (RH) is close to or higher than this level, even for a low surface salt concentration. (authors)

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

Facility Utilization and Risk Analysis for Remediation of Legacy Transuranic Waste at the Savannah River Site - 13572

Savannah River Nuclear Solutions (SRNS) completed the Accelerated TRU Project for remediating legacy waste at the Savannah River Site with significant cost and schedule efficiencies due to early identification of resources and utilization of risk matrices. Initial project planning included identification of existing facilities that could be modified to meet the technical requirements needed for repackaging and remediating the waste. The project schedule was then optimized by utilization of risk matrices that identified alternate strategies and parallel processing paths which drove the overall success of the project. Early completion of the Accelerated TRU Project allowed SRNS to pursue stretch goals associated with remediating very difficult TRU waste such as concrete casks from the hot cells in the Savannah River National Laboratory. Project planning for stretch goals also utilized existing facilities and the risk matrices. The Accelerated TRU project and stretch goals were funded under the American Recovery and Reinvestment Act (ARRA). (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Development and Testing of Techniques for In-Ground Stabilization, Size Reduction and Safe Removal of Radioactive Wastes Stored in Large Containments in Burial Grounds - 13591

Radioactive waste materials, including Transuranic (TRU) wastes from laboratories have been stored below ground in large containments at a number of sites in the US DOE Complex, and at nuclear sites in Europe. These containments are generally referred to as caissons or shafts. The containments are in a range of sizes and depths below grade. The caissons at the DOE's Hanford site are cylindrical, of the order of 2,500 mm in diameter, 3,050 mm in height and are buried about 6,000 mm below grade. One type of caisson is made out of corrugated pipe, whereas others are made of concrete with standard re-bar. However, the larger shafts in the UK are of the order of 4,600 mm in diameter, 53,500 mm deep, and 12,000 below grade. This paper describes the R and D work and testing activities performed to date to evaluate the concept of in-ground size reduction and stabilization of the contents of large containments similar to those at Hanford. In practice, the height of the Test Facility provided for a test cell that was approximately 22' deep. That prevented a 'full scale mockup' test in the sense that the Hanford Caisson configuration would be an identical replication. Therefore, the project was conducted in two phases. The first phase tested a simulated Caisson with surrogate contents, and part of a Chute section, and the second phase tested a full chute section. These tests were performed at VJ Technologies Test Facility located in East Haven, CT, as part of the Proof of Design Concept program for studying the feasibility of an in-situ grout/grind/mix/stabilize technology for the remediation of four caissons at the 618-11 Burial Ground at US Department of Energy Hanford Site. The test site was constructed such that multiple testing areas were provided for the evaluation of various tools, equipment and procedures under conditions that simulated the Hanford site, with representative soils and layout dimensions. (authors)

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

Successful Characterization and Remedial Contour of Highly Contaminated Mercury Soil at the Y-12 National Security Complex - 13593

An area known as the 81-10 pad within the footprint of the Y-12 National Security Complex, suspected to be heavily contaminated with mercury, was slated for characterization in support of a Federal Facilities Agreement (FFA) milestone to be accomplished by September 30, 2012. A full remedial design report (RDR) required the soil in Exposure Unit -9 (EU-9) to be fully characterized for a number of contaminates of concern including mercury. The goal of this characterization effort was to determine what soil, if any, would need to be removed for the protection of industrial workers and impacts to the surface and ground water. Funding for this project was made available using buy-back scope under the American Recovery and Reinvestment Act (ARRA). The EU-9 soil unit involved 3 different classifications which were determined as follows: Class 1: Known to have been impacted, contamination is likely; Class 2: Suspected to have been impacted, contamination is unknown; Class 3: Area not known to have been impacted, contamination unlikely. Due to various sampling and analysis events since the 1980's, significant mercury contamination was expected under the concrete pad of an area known as 81-10. Mercury contamination outside of the boundary of this pad within the EU-9 footprint was not known and therefore an original planned estimate of 1,461 cubic meters of material were expected to be heavily contaminated with mercury requiring removal, treatment and disposal. Through the use of a highly effective nature and extent sampling and analysis design that involved a hybrid of statistically-based and judgmental sampling, the actual remedial contour requiring removal was approximately 717 cubic meters, roughly 12% of the original estimate. This characterization approach was executed in full compliance with the Record of Decision (ROD) [1] documents that were agreed upon by the U.S. Department of Energy, Environmental Protection Agency and Tennessee Department of Environment and Conservation. In addition, the RDR was completed ahead of the FFA milestone date of September 30, 2012. (authors)

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

Progress in Decommissioning the Humboldt Bay Power Plant - 13604

Decommissioning of the Pacific Gas and Electric (PG and E) Company Humboldt Bay Power Plant (HBPP) Unit 3 nuclear facility has now, after more than three decades of SAFSTOR and initial decommissioning work, transitioned to full-scale decommissioning. Decommissioning activities to date have been well orchestrated and executed in spite of an extremely small work site with space constricted even more by other concurrent on-site major construction projects including the demolition of four fossil units, construction of a new generating station and 60 KV switchyard upgrade. Full-scale decommissioning activities - now transitioning from Plant Systems Removal (PG and E self-perform) to Civil Works Projects (contractor performed) - are proceeding in a safe, timely, and cost effective manner. As a result of the successful decommissioning work to date (approximately fifty percent completed) and the intense planning and preparations for the remaining work, there is a high level of confidence for completion of all HBPP Unit 3 decommissions activities in 2018. Strategic planning and preparations to transition into full-scale decommissioning was carried out in 2008 by a small, highly focused project team. This planning was conducted concurrent with other critical planning requirements such as the loading of spent nuclear fuel into dry storage at the Independent Spent Fuel Storage Installation (ISFSI) finishing December 2008. Over the past four years, 2009 through 2012, the majority of decommissioning work has been installation of site infrastructure and removal of systems and components, known as the Plant System Removal Phase, where work scope was dynamic with significant uncertainty, and it was self-performed by PG and E. As HBPP Decommissioning transitions from the Plant System Removal Phase to the Civil Works Projects Phase, where work scope is well defined, a contracting plan similar to that used for Fossil Decommissioning will be implemented. Award of five major work scopes in various stages of development are planned as they include: Turbine Building Demolition, Nuclear Facilities Demolition and Excavation, Intake and Discharge Canal Remediation, Office Facility Demobilization, and Final Site Restoration. Benefits realized by transitioning to the Civil Works Projects Phase with predominant firm fixed-price/fixed unit price contracting include single civil works contractor who can coordinate concrete shaving, liner removal, structural removal, and other demolition activities; streamline financial control; reduce PG and E overhead staffing; and provide a specialized Bidder Team with experience from other similar projects. (authors)

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

Removal of the Plutonium Recycle Test Reactor - 13031

The 309 Facility housed the Plutonium Recycle Test Reactor (PRTR), an operating test reactor in the 300 Area at Hanford, Washington. The reactor first went critical in 1960 and was originally used for experiments under the Hanford Site Plutonium Fuels Utilization Program. The facility was decontaminated and decommissioned in 1988-1989, and the facility was deactivated in 1994. The 309 facility was added to Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) response actions as established in an Interim Record of Decision (IROD) and Action Memorandum (AM). The IROD directs a remedial action for the 309 facility, associated waste sites, associated underground piping and contaminated soils resulting from past unplanned releases. The AM directs a removal action through physical demolition of the facility, including removal of the reactor. Both CERCLA actions are implemented in accordance with U.S. EPA approved Remedial Action Work Plan, and the Remedial Design Report / Remedial Action Report associated with the Hanford 300-FF-2 Operable Unit. The selected method for remedy was to conventionally demolish above grade structures including the easily distinguished containment vessel dome, remove the PRTR and a minimum of 300 mm (12 in) of shielding as a single 560 Ton unit, and conventionally demolish the below grade structure. Initial sample core drilling in the Bio-Shield for radiological surveys showed evidence that the Bio-Shield was of sound structure. Core drills for the separation process of the PRTR from the 309 structure began at the deck level and revealed substantial thermal degradation of at least the top 1.2 m (4LF) of Bio-Shield structure. The degraded structure combined with the original materials used in the Bio-Shield would not allow for a stable structure to be extracted. The water used in the core drilling process proved to erode the sand mixture of the Bio-Shield leaving the steel aggregate to act as ball bearings against the core drill bit. A redesign is being completed to extract the 309 PRTR and entire Bio-Shield structure together as one monolith weighing 1100 Ton by cutting structural concrete supports. In addition, the PRTR has hundreds of contaminated process tubes and pipes that have to be severed to allow for a uniformly flush fit with a lower lifting frame. Thirty-two 50 mm (2 in) core drills must be connected with thirty-two wire saw cuts to allow for lifting columns to be inserted. Then eight primary saw cuts must be completed to severe the PRTR from the 309 Facility. Once the weight of the PRTR is transferred to the lifting frame, then the PRTR may be lifted out of the facility. The critical lift will be executed using four 450 Ton strand jacks mounted on a 9 m (30 LF) tall mobile lifting frame that will allow the PRTR to be transported by eight 600 mm (24 in) Slide Shoes. The PRTR will then be placed on a twenty-four line, double wide, self powered Goldhofer for transfer to the onsite CERCLA Disposal Cell (ERDF Facility), approximately 33 km (20 miles) away. (authors)

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

Characterizations of the radioactive waste by the remotely-controlled collimated spectrometric system

Decontamination and decommissioning of the research reactors MR (Testing Reactor) and RFT (Reactor of Physics and Technology) has recently been initiated in the National Research Center (NRC) 'Kurchatov institute', Moscow. In the building, neighboring to the reactor, the storage of HLRW is located. The storage is made of monolithic concrete in which steel cells depth 4 m are located. In cells of storage the HLRW packed into cases are placed. These the radioactive waste are also subject to export on long storage in the specialized organization. For characterization of the radioactive waste in cases the remote-controlled collimated spectrometer system was used. The system consists of a spectrometric collimated gamma-ray detector, a color video camera and a control unit, mounted on a rotator, which are mounted on a tripod with the host computer. For determination of specific activity of radionuclides in cases, it is developed programs of calculation of coefficients of proportionality of specific activity to the corresponding speeds of the account in peaks of full absorption at single specific activity of radionuclides in cases. For determination of these coefficients the mathematical model of spectrometer system based on the Monte-Carlo method was used. Dependences of calibration coefficients for various radionuclides from distance between the detector and a case at various values of the radioactive waste density in cases are given. Measurements of specific activity in cases are taken and are discussed. By results of measurements decisions on the appeal of the radioactive waste being in cases are made. (authors)

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

Preliminary Study on Recycling and Reuse of Metals from Decommissioning in Construction of Disposal Vaults in Slovakia - 17217

The decommissioning of nuclear installations represents a complex process resulting in the generation of large amounts of various waste materials. Majority of them are non-contaminated, however some of them may be hazardous as well (e.g. chemicals, asbestos, oil products, etc.). During decommissioning of NPP, significant quantities of materials with various levels of radionuclides concentration arise as well. All of these materials should be managed appropriately keeping in mind safety principles, legislative requirements, lessons learned and other relevant factors. Therefore, selection of an appropriate strategy for management of the materials arising from decommissioning is crucial and strongly influences the overall effectiveness of decommissioning process. In line with international incentives for optimization of radioactive material management, concepts of recycling and reuse of slightly radioactive materials are widely discussed and applications of these concepts are analyzed. Therefore, the paper is devoted to the preliminary study of potential for recycling of re-melted metals from decommissioning in form of reinforcement bars or mesh used in the construction of ferro-concrete disposal vaults within the near surface repository. The paper presents the preliminary results and discusses the technical and non-technical aspects related to the recycling scenario. Moreover, comparison of various scenarios related to the scrap metal clearance and recycling within nuclear industry is included. Key findings and lessons learned from study may be useful for the international community dealing with waste management optimization and in particular with recycling of materials within nuclear industry. (authors)

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

Disposing of waste package with respect to physico-chemical characteristics - 15041

The waste inventory to be taken into account for the design of the Cigeo project is defined in the industrial program for waste management established in cooperation between nuclear operators (EDF, AREVA, CEA) and Andra. This waste inventory has been constituted with a list of different families of radioactive waste packages from High Level Waste (HLW) to Intermediate Level Long-lived Waste (IL-LLW). IL-LLW include a large variety of items such as structural elements from fuel assemblies (cladding from the fuel rods called 'hulls', end pieces called 'end caps' and assembly spacer grids, etc.), sludge from effluent treatment or miscellaneous equipment (filters, pumps, etc.). Most ILLW (except solidified sludges) are basically metallic but organic and inorganic components such as plastics (cellulose, PVC...) and salts may also be included. The rules used for designing Cigeo are based on the separation of waste having different physico-chemical characteristics in different cells. However it could be possible to dispose of waste packages with different physico-chemical characteristics in the same cell if we can demonstrate that a type of waste does not generate significant physico-chemical perturbations on other co-stored families. The first step of our analysis has been the identification of categories of materials present in waste likely to interact with each other. The second step consisted in analysing qualitatively the physico-chemical interactions between these materials. In the third step, we defined criteria allowing evaluation of the consequences of these interactions on (i) gas production (mainly H{sub 2}), (ii) radionuclide release and transfer and (iii) degradation of concrete components of the disposal cell. We built then a table allowing qualitative comparison of the consequences of the interactions between the materials types defined in the first step. The conclusion of this study shows that it is necessary to pay a very careful attention before disposing of waste packages containing organic components nearby waste packages containing actinides radionuclides. The geochemical phenomena coupling must also be quantitatively evaluated so as to analyse possible interaction processes between the released organic compounds and the cementitious components of the cells, the near-field clay-stone and the radionuclides migration. (authors)

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

The Cires facility in Morvilliers: from a VLL waste disposal facility to the development of industrial activities - 15252

The disposal facility of CIRES for very low level wastes was designed to provide safe and cost effective services to waste generators, in particular for decommissioning works. The facility could significantly increase waste annual deliveries to meet waste generators' needs and accommodate large disused components as PWR steam generators after decontamination. However an anticipated saturation of its capacity is foreseen as needs for disposal has doubled since its start-up. The geometry of the cells was optimized and the technical capacity of the repository is presently by 40% above its licensed capacity. Saturation has been delayed; however a new facility will be necessary by 2030 and other waste management options are to be considered as recycling of metallic wastes or crushed concrete rubble. CIRES facility also became Andra's industrial platform for the management of institutional wastes. A grouping and a storage facility have been built and a new processing facility should start up in 2016. (authors)

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

High-level waste relocation requires use of new technology in aging nuclear facility - 15269

The West Valley Demonstration Project (WVDP) is preparing to relocate 278 canisters of vitrified high-level waste (HLW) from their current location inside the Main Plant Process Building (MPPB) to an outdoor pad for long-term, on-site storage. The waste was produced at the WVDP between 1996 and 2002 and was originally placed in interim storage inside the MPPB in anticipation of off-site shipment for disposal. Relocation of the canisters to a new on-site location is now necessary as part of WVDP Phase 1 Site Decommissioning work to enable pre-demolition activities inside the MPPB to proceed. The logistics of packaging and moving the HLW canisters within an aging facility to another on-site location has provided numerous technical challenges with subsequent innovations. Significant preparations have taken place to resolve structural inadequacies, including sub-surface soil subsidence issues, antiquated equipment, and facility reconfigurations to support up to 134 metric ton (148 ton) loads. To perform the transfer and storage of the HLW canisters, new equipment and fabrications were required including a robust storage system consisting of stainless steel canister overpacks within steel-reinforced concrete storage casks, three specialty cask transporters, a remotely operated welding system and construction of an outdoor engineered storage pad to store the 56 casks containing the relocated HLW canisters. Packaging and relocation of vitrified canisters is scheduled to begin in 2015. (authors)

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

Mega salt-stone Disposal Units at the Savannah River site - 15322

The Liquid Waste (LW) System at Savannah River Site (SRS) is a highly integrated operation involving safely storing radioactive liquid waste in underground storage tanks; vitrifying the higher activity waste; and storing the vitrified waste in stainless steel canisters until permanent disposition is provided; and removing, treating, and dispositioning the low-level waste (LLW) fraction in concrete with final disposition being the salt-stone Disposal Units (SDUs) at the SRS salt-stone Disposal Facility (SDF). Treatment and disposition of salt waste is the critical path to completion of the SRS Liquid Waste (LW) Disposition Program. Successful and timely salt waste removal and disposal is integral to efforts by SRS to proceed with all aspects of tank cleanup and removal from service. The construction and implementation of the first-of-a-kind Mega SDU (MSDU) is underway to provide the most environmentally safe and cost-effective solution for salt processing and long term storage of the LLW at SRS. The MSDU design is a duplicate of the Water Treatment Tanks located in Syracuse, New York. This commercial design, developed by an affiliate of Savannah River Remediation LLC (SRR), was slightly modified to meet the requirements of the nuclear industry. The MSDU is a critical link in salt disposition. In comparison to the current SDU design, approximately 46 meters (150 feet) in diameter and 7 meters (22 feet) tall with a capacity of approximately 11,000 m{sup 3} (3.0 million gallons), the MSDU is 114 meters (375 feet) in diameter and 13 meters (43 feet) tall which will provide a capacity of approximately 121,000 m{sup 3} (32.0 million gallons) to receive the LLW fraction embedded in a grout-like form. The MSDU will provide the taxpayers an estimated $300 M in life cycle savings at SRS by reducing the number of SDUs from the estimated 72 smaller vessels to 7 of the new larger vessels. Additionally, based on the reduced number of SDUs, the environmental footprint is significantly reduced as well. Due to the similarities between the existing SDU design and the MSDU design, efforts to integrate the MSDU into the LW system are minimal. Required infrastructure modifications will consist of tying into the existing grout and drain lines, as well as the installation of new electrical and instrumentation services for the MSDU. Grout flow testing has been successfully completed, with results showing the MSDU will be a safe and cost-effective replacement for the existing SDUs. The MSDU design is complete, Critical Decision (CD)-3 was approved on June 17, 2014, and the first MSDU, salt-stone Disposal Unit n.6 (SDU 6), is being constructed in the existing LW SDF operated by SRR as a Capital Line Item project. As an added benefit, once constructed, lessons learned will be available for future MSDUs at SRS and throughout the complex as necessary. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗