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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute'

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayk'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ- spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Consequence Analysis of Residual Water in a Storage Canister - 20484

Recent observations from projects sponsored under the DOE-Nuclear Energy Spent Fuel and Waste Disposition campaign show that residual free water, well above the 0.4 gm-moles that correspond to a 3 torr rebound pressure after a 30-minute hold following prototypic drying, may remain within SNF canisters used for long-term dry storage. Specifically, experimental observations in an Integrated Research Project showed subzero temperatures and a local pocket of frozen water in a mockup fuel assembly subjected to an industry practiced drying process. In gas sampling and analysis from the High Burnup (HBU) Demonstration project, using a fully loaded and dried SNF canister, indicated presence of 100 g of residual water had remained in the canister. The effects of residual water in terms of potential adverse impacts on the condition of the SNF and the canister internals during long-term dry storage has been rigorously evaluated to address inadvertent residual water contents. The impact of residual water on the fuel and canister internals, with a focus on the oxidation of the cladding and of postulated exposed (breached-clad) fuel, were evaluated using an integrated set of time-dependent environmental conditions with semi-empirical materials oxidation models. An 'integration model,' with coupled sub-models, developed and previously used by the Center for Nuclear Waste Regulatory Analysis (CNWRA), was applied and expanded with improved sub-models for water radiolysis, and for cladding and fuel oxidations. Cladding oxidation models for various alloys were developed using the rate constants for various cladding alloys in the dry storage temperature range. A burnup-dependent fuel oxidation model was developed; it accounted for inhibition effect of fission products on fuel oxidation rates. The potential for extension of the initial breach flaw (1 mm hole) was evaluated by two separate mechanics criteria. In the first case, a critical cladding strain criterion was used. In the second case, a critical stress intensity was used to evaluate stability of a planar axial flaw postulate. A failure condition exists when a sufficient amount of fuel oxidation (at the postulated breached-clad location) causes fuel pellet swelling to load the cladding to meet a mechanics criterion for flaw extension. The integration model's results provide the following insights. The radiolysis-generated oxygen is consumed with a partitioning between the cladding and the fuel. The extent of cladding oxidation is no more than 2 μm additional consumption of the cladding metal even with 10 moles of residual water, indicating that changes in cladding conditions due to water are expected to be negligible. The conditions for cladding failure (breach opening size extension) directly correlates with the extent of fuel oxidation. Fuel oxidation significant to extend the breach could occur even with 5.5 moles of residual water. The results indicate that fuel oxidation vis-a-vis cladding oxidation is more likely under a higher radiation field, i.e., when radiolytic decomposition of the residual water occurs in few years compared to several decades, and that the conditions to cause failure are increased with fuel and cladding temperatures approaching the peak storage temperature limit of 400 deg. C. (authors)

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

GrayQb Single-Faced Version 2 Open Environment Test - 16344

The GrayQb{sup TM} Single Faced Version 2 (SF2) is a non-destructive examination device developed to generate radiation contour maps showing source locations and relative radiological levels present in an area under examination. This device allows for characterization of radioactively contaminated areas such as hot cells, gloveboxes, small and large rooms, hallways, and waste tanks. The goal of GrayQb{sup TM} (pronounced 'Gray Cube') is to provide a low cost easily deployable device to aid in the process of decontamination and decommissioning (D and D). The GrayQb{sup TM} SF2 device is approximately a six inch cube weighing seven (7) pounds composed of a tungsten shield with integrated collimator, radiation imaging medium (Phosphor Storage Plate), and digital camera. This paper details the design upgrades incorporated into the new version of the GrayQb{sup TM} SF2 device and the characterization testing of this upgraded device. Device results from controlled testing performed at the Savannah River Site (SRS) Health Physics Instrument Calibration Laboratory (HPICL) are presented, as well as results from the open environment field testing performed in the E-Area Low Level Waste Storage Area. Follow on work performed deploying the GrayQb SF2 device at the Hanford Plutonium Reclamation Facility (PRF) to assist in the radiological characterization of the canyon will also be presented. Testing of the redesigned GrayQb{sup TM} SF2 device demonstrated its ability to detect low energy isotopes, e.g. Am-241 and Pu-238, not only indoors in controlled environments with well-defined source locations, but also in open environment field areas with a significant high energy background. Additionally, it was demonstrated that the device could detect higher energy isotopes (Cs-137 and Co-60) in a controlled setting and with additional testing and/or modifications to device and software, would likely be able to detect these isotopes in the open environment field testing. Follow on 2015 work performed at Hanford proved GrayQb to be a useful tool in the radiological characterization of the PRF canyon. Distribution tendencies of the radiological hot spots were identified that provided insights which will assist in more efficient planning of the continued D and D effort. Hot spots were identified which can be targeted during the future D and D effort and used as markers during the NDA assessment to quantify the radiological environment. (authors)

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

Application of the Environmentally Non-Aggressive Dissolvents to Mobilize Metals in Tailings in the Norilsk Region

Under analysis is on-site extractability of nonferrous and noble metals from tailings by leaching with natural reagents (humus acids, water, lignin). Redistribution of nonferrous and noble metals in a percolation bed made of old tailings in the Norilsk copper–nickel ore field is studied. Kinetics of dissolving of nonferrous and noble metals under capillary ascensional percolation is investigated. It is shown that directional capillary percolation of solutions is applicable to extraction of nonferrous and noble metals under long-term mining waste storage.

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

Drying Spent Magnox Fuel for Interim Dry Storage

Spent Magnox fuel is usually stored under water for shielding and cooling following use in-reactor, but the storage water can react with Magnox clad fuel and cause corrosion and forms a sludge which generates a secondary waste product. One option to prevent unwanted corrosion and avoid expensive waste management is to dry the wet fuel. This project aims to test a drying process which will convert wet and corroded Magnox fuel to a stable form to demonstrate viability for conversion from wet to dry storage. Samples of corroded Magnox supplied by National Nuclear Laboratory. Corrosion product retrieved by various methods including dry scraping, ultrasonic treatment to form sludge and sludge drying to isolate suspension. TGA undertaken on Sample 1 and Sample 2, with brucite and hydromagnesite for comparison. Both samples showed ∼10% mass loss up to 250 deg. C attributable to water held in sample. Mass loss from dehydroxylation which liberates further water in both samples. TGA data plotted alongside chemically pure brucite and hydromagnesite shows expected similarities/differences between Sample 1(mostly brucite) and Sample 2 (brucite/hydromagnesite mixture). Following treatment, dried samples observed to slowly gain mass in ambient conditions. Presumed to be hygroscopic absorption of air water vapour as effect is reversible, with similar effect observed for hydromagnesite and brucite. Etched Magnox vacuum dried for 3.5 h at 90 deg. C to test for reaction with residual water - some discoloration but no noticeable corrosion observed. Drying etched Magnox metal to observe surface effects/corrosion during drying process.

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

Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Screening Tests

More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second LAW immobilization facility will be needed for the expected volume of LAW requiring immobilization. A cementitious waste form known as Cast Stone is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. Further, the waste form must be tested to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF. The PA is needed to satisfy both Washington State IDF Permit and DOE Order requirements. Cast Stone has been selected for solidification of radioactive wastes including WTP aqueous secondary wastes treated at the Effluent Treatment Facility (ETF) at Hanford. A similar waste form called Saltstone is used at the Savannah River Site (SRS) to solidify its LAW tank wastes.

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

Development of a Universal Canister for Disposal of High-Level Waste in Deep Boreholes - 16482

The mission of the United States Department of Energy's (DOE's) Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research. Some of the wastes that must be managed have been identified as good candidates for disposal in a deep borehole in crystalline rock. In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister-based system that can be used for handling these wastes during the disposition process (i.e., storage, transfer, transportation, and disposal) could facilitate the eventual disposal of these wastes. Development of specifications for the universal canister system will consider the regulatory requirements that apply to storage, transportation, and disposal of the capsules, as well as operational requirements and limits that could affect the design of the canister (e.g., deep borehole diameter). In addition, there are risks and technical challenges that need to be recognized and addressed as Universal Canister system specifications are developed. This paper provides an approach to developing specifications for such a canister system that is integrated with the overall efforts of the DOE's Used Fuel Disposition Campaign's Deep Borehole Field Test and compatible with planned storage of potential borehole candidate wastes. (authors)

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

Using Process Knowledge to Manage Disposal Classification of Ion-Exchange Resin - 13566

It has been previously shown by EPRI [1] that Class B and C resins represent a small portion by volume of the overall generation of radioactively contaminated resins. In fact, if all of the resins were taken together the overall classification would meet Class A disposal requirements. Lowering the classification of the ion exchange resins as they are presented for disposal provides a path for minimizing the amount of waste stored. Currently there are commercial options for blending wastes from various generators for Class A disposal in development. The NRC may have by this time introduced changes and clarifications to the Branch Technical Position (BTP) on Concentration Averaging and Encapsulation [2] that may ultimately add more flexibility to what can be done at the plant level. The BTP has always maintained that mixtures of resins that are combined for ALARA purposes or operational efficiency can be classified on the basis of the mixture. This is a point often misinterpreted and misapplied. This paper will address options that can be exercised by the generator that can limit B and C waste generation by more rigorous tracking of generation and taking advantage of the normal mix of wastes. This can be achieved through the monitoring of reactor coolant chemistry data and coupled with our knowledge of radionuclide production mechanisms. This knowledge can be used to determine the overall accumulation of activity in ion-exchange resins and provides a 'real-time' waste classification determination of the resin and thereby provide a mechanism to reduce the production of waste that exceeds class A limits. It should be noted that this alternative approach, although rarely used in a nuclear power plant setting, is acknowledged in the original BTP on classification [3] as a viable option for determining radionuclide inventories for classification of waste. Also included is a discussion of an examination performed at the Byron plant to estimate radionuclide content in the final waste stream from upstream sampling of reactor coolant and fuel pool water. (authors)

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

A Study on Korea Experience in Regulating the Transport of Low and Intermediate-level Radioactive Wastes - 17182

In the Republic of Korea, the transport of Low and Intermediate-Level Radioactive Wastes (LILW) from nuclear power plants (NPP), nuclear research reactor, and radioactive isotope storage facility to Gyeongju LILW disposal facility has increased since the operation license of facility was approved in December 2014. As a result, there has been an increase in public anxiety and concerns over damage or social disorder with LILW transport. So the necessity of strengthened safety confirmation on LILW has been constantly raised. In August 2016, the Enforcement Regulations of the Nuclear Safety Act were amended in relation to the reporting and inspection of LILW transport. The first section of this paper describes the domestic transport regulations including the recently revised contents. The second section shows that the status of domestic and inspection of LILW transport were introduced and transportation regulatory experience was shared in Korea. (authors)

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

Sodium and Lithium Hydride Shields At the Department of Energy's Oak Ridge Reservation - 18412

URS-CH2M Oak Ridge LLC (UCOR) is responsible for cleanup of the U.S. Department of Energy's (DOE) East Tennessee Technology Park (ETTP) in Oak Ridge, Tennessee. Formerly known as the K-25 site, the facility enriched uranium for major defense and energy programs for more than 40 years. DOE is seeking the safest, most environmentally protective and fiscally responsible treatment for its large inventory of radioactive-contaminated metals known as 'shields'. The 66 sodium and lithium shields were constructed for use in experiments at the Tower Shielding Facility (TSF), Oak Ridge National Laboratory (ORNL). The TSF operated from 1954 to 1992 and was designed and built for radiation-shielding studies. The shield containers are constructed of aluminum or stainless steel of varying sizes and shapes that were then filled with either sodium metal or lithium hydride material. The historical records indicate the sodium shields were constructed at the Molten Salt Reactor for use in experiments in the 1971 time frame, while the lithium shields were constructed at the Y-12 site and were used in experiments as early as 1961. A subset of the sodium and lithium shields contain depleted uranium. The sodium and lithium shields were initially used at the TSF to perform in-depth measurements of the neutron transport through the shield materials. Based on process knowledge, the shields are managed as radiologically contaminated containers. Efforts have been made over the last fifteen years to find recycle markets for both the sodium and lithium hydride shields. This includes an onsite operation that was initiated at the East Tennessee Technology Park (ETTP) in 2004. After an operational exothermic reaction incident occurred soon after startup, the onsite ETTP recycling action was discontinued and a search for offsite recycling facilities was initiated. In 2016, the Y-12 National Security Complex was contacted regarding recycle/repurpose of the lithium shields; however, the lithium was determined to not be acceptable for their current repurpose process. To date, no offsite recycling outlets have been identified. Recent progress on the waste disposal actions, a shipment of 22 small sodium shields have been shipped and treated offsite. Therefore, the current shield inventory consists of 29 sodium shields located at ETTP and 15 sodium and lithium hydride shields staged at ORNL. The time period needed to reasonably identify safe disposition or recycling outlets for the shields was expected to exceed the one-year Land Disposal Restrictions storage time limit; therefore, approval was received to add the shields to the Site Treatment Plan (STP) in March, 2017. Until a disposition path is found, the shields will continue to be stored in a safe configuration that is protective of human health and the environment. This engineering evaluation identifies and screens the alternatives for dispositioning the shields at the Oak Ridge(ORR). Four alternatives were evaluated including; (1) macro-encapsulation, (2) deactivation, (3) recycle and (4) leave-in-place. An initial screening was performed, and the alternative passing the initial screening are then evaluated in more detail. The evaluation criteria used included environmental risk, technical feasibility, administrative feasibility, worker safety, nuclear safety, transportation safety, and cost. (author)

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

Characterizing 5,000 Properties to Determine which Require Radiological Remediation - 19152

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste (hLLRW) in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada. Conducting work safely, while delivering the PHAI on time and within budget, is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.7 million cubic metres of hLLRW from various locations within the community. The Port Hope Project includes a unique and important component - the Property Radiological Survey. This survey involves the interior and exterior investigation of approximately 4,800 small-scale sites (SSS) - primarily residential - in the Municipality of Port Hope for the presence or absence of hLLRW. This paper provides an overview of the characterization, delineation and design for remediation of historic LLRW for the SSS, a grouping of sites within the municipality comprised of residential properties, municipal properties including municipal road allowances, private road allowances and temporary storage sites. Comprising over 4,800 parcels of land, a robust remedial process was designed to identify, characterize, delineate and remediate interior and exterior hLLRW contamination found in soils and building materials. The process adheres to the framework of the Port Hope Project Waste Nuclear Substance Licence (License) and its associated governance frameworks and regulatory requirements [1]. The overall goal is for CNL to be able to issue the owner of each parcel of land a Compliance Letter stating that the site meets the PHAI Clean-up Criteria, either after completion of characterization (if no hLLRW is identified) or after completion of remediation and verification. (authors)

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

Strategy for Used Fuel Acquisition

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The Storage and Transportation staffs within the UFDC are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While both wet and dry storage have been shown to be safe options for storing UNF, the focus of the program is on dry storage at reactor or centralized locations. Because limited information is available on the properties of high burnup fuel (exceeding 45 gigawatt-days per metric tonne of uranium [GWd/MTU]), and because much of the fuel currently discharged from today’s reactors exceeds this burnup threshold, a particular emphasis of this program is on high burnup fuels. Since high burnup used fuels have only been loaded into dry storage systems in the past decade or so, these materials are available to the UFDC for testing in only very limited quantities. Much of what is available has come via NRC testing programs. Some of these fuels may have achieved "high burnup," but that does not mean they were designed for high burnup use (e.g. lower enrichments, smaller plenum spaces, extra reactor cycles). The handling and transfer of these materials from utility to laboratory has not always been prototypical of how used nuclear fuel is prepared for dry storage; these fuels are not subjected to the same vacuum drying conditions that can lead to changes in hydride morphology that will affect the mechanical properties of the fuel. It is recognized that sources of used high burnup fuel that can be handled in a manner consistent with how fuel is readied for dry storage is essential to the mission of the UFDC. This report documents what types of fuel are of interest to the campaign, and how those fuels could be acquired and shipped to the Idaho National Laboratory (INL) for incorporation into the campaign R&D mission. It also identifies any gaps in INL capabilities that might preclude working with one fuel type or another.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Pressure Build-Up During the Fire Test in Type B(U) Packages Containing Water - 13280

The safety assessment of packages for the transport of radioactive materials with content containing liquids requires special consideration. The main focus is on water as supplementary liquid content in Type B(U) packages. A typical content of a Type B(U) package is ion exchange resin, waste of a nuclear power plant, which is not dried, normally only drained. Besides the saturated ion exchange resin, a small amount of free water can be included in these contents. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific issues. An overview of these issues is provided. The physical and chemical compatibility of the content itself and the content compatibility with the packages materials must be demonstrated for the assessment. Regarding the mechanical resistance the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could for example be caused by radiolysis of the liquid and must be taken into account for the storage period. If the package is stressed by the total inner pressure, this pressure leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The total pressure in any calculation is the sum of partial pressures of different gases which can be caused by different effects. The pressure build-up inside the package caused by the regulatory thermal test (30 min at 800 deg. C), as part of the cumulative test scenario under accident conditions of transport is discussed primarily. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from beginning of the thermal test until cooling-down. In this case, while calculating the temperature distribution, conduction and radiation as well as evaporation and condensation during the associated process of transport have to be considered. This paper discusses limiting amounts of water inside the cask which could lead to unacceptable pressure and takes into account saturated steam as well as overheated steam. However, the difficulties of assessing casks containing wet content will be discussed. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Concrete shrinkage and creep under drying/wetting cycles

Concrete shrinkage and creep under variable hydric conditions are important factors for the safety and durability of concrete especially in nuclear reactor or nuclear waste storage background. A large (and long – more than 900 days) experimental campaign, conducted on two different concretes, has therefore been designed to study the strains of non loaded and loaded concretes submitted to drying and liquid water imbibition cycles. For the purposes of comparison, concrete strains and mass variations during drying only (50% RH) and/or following cycles of drying/rewetting were also recorded. This allowed the identification of desiccation shrinkage, basic creep and drying creep at 10 MPa of axial stress. Important results were found and have shown that the final mass and strain are not deeply modified by the introduction of a rewetting phase either for a loaded or a non loaded material.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Management of Legacy Spent Nuclear Fuel Wastes at the Chalk River Laboratories: The Challenges and Innovative Solutions Implemented - 13301

AECL's Fuel Packaging and Storage (FPS) Project was initiated in 2004 to retrieve, transfer, and stabilize an identified inventory of degraded research reactor fuel that had been emplaced within in-ground 'Tile Hole' structures in Chalk River Laboratories' Waste Management Area in the 1950's and 60's. Ongoing monitoring of the legacy fuel storage conditions had identified that moisture present in the storage structures had contributed to corrosion of both the fuel and the storage containers. This prompted the initiation of the FPS Project which has as its objective to design, construct, and commission equipment and systems that would allow for the ongoing safe storage of this fuel until a final long-term management, or disposition, pathway was available. The FPS Project provides systems and technologies to retrieve and transfer the fuel from the Waste Management Area to a new facility that will repackage, dry, safely store and monitor the fuel for a period of 50 years. All equipment and the new storage facility are designed and constructed to meet the requirements for Class 1 Nuclear Facilities in Canada. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Interaction of Sr-90 with site candidate soil for demonstration disposal facility at Serpong

Interaction of radiostrontium (Sr-90) with site candidate soil for demonstration disposal facility to be constructed in the near future at Serpong has been done. This activity is to anticipate the interim storage facility at Serpong nuclear area becomes full off condition, and show to the public how radioactive waste can be well managed with the existing technology. To ensure that the location is save, a reliability study of site candidate soil becomes very importance to be conducted through some experiments consisted some affected parameters such as contact time, effect of ionic strength, and effect of Sr{sup +} ion in solution. Radiostrontium was used as a tracer on the experiments and has role as radionuclide reference in low-level radioactive waste due to its long half-live and it's easy to associate with organism in nature. So, interaction of radiostrontium and soil samples from site becomes important to be studied. Experiment was performed in batch method, and soil sample-solution containing radionuclide was mixed in a 20 ml of PE vial. Ratio of solid: liquid was 10{sup −2} g/ml. Objective of the experiment is to collect the specific characteristics data of radionuclide sorption onto soil from site candidate. Distribution coefficient value was used as indicator where the amount of initial and final activities of radiostrontium in solution was compared. Result showed that equilibrium condition was reached after contact time 10 days with Kd values ranged from 1600-2350 ml/g. Increased in ionic strength in solution made decreased of Kd value into soil sample due to competition of background salt and radiostrontium into soil samples, and increased in Sr ion in solution caused decreased of Kd value in soil sample due to limitation of sorption capacity in soil samples. Fast condition in saturated of metal ion into soil samples was reached due to a simple reaction was occurred.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

From Radtrack to eMWaste(r) G2 process-based integration

This paper provides a process-based examination of the integration of two separate but related software applications into Sandia National Laboratories’ (SNL’s) Radioactive and Mixed Waste Management Facility’s (RMWMF) data management system. First, data collection will be upgraded through the use of a new web-based Waste Description and Disposal Request (WDDR), which SNL personnel whose work creates radioactive and hazardous waste (generators) will use to characterize dozens of specific attributes of each waste parcel. Second, a commercial software package for data storage and manipulation — eMWaste® G2 by Attention IT — will replace two existing software applications and will track all radioactive and mixed waste at Sandia. Integrating these applications into a web-based system will give the RMWMF increased system flexibility, reduced reliance on key personnel, and improved document control, traceability and efficiency. To meet these goals and ensure the applications are developed in an efficient, quality-oriented environment, it is critical to first conduct a thorough evaluation of the affected processes. This paper will analyze the processes supported by the new applications and the requirements for a successful integration. A comprehensive analysis of both project flow and data management requirements, illustrated through original flow diagrams and other charts, will provide support to the development team and programmers as they work to integrate the applications into the new system.

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

Characterisation of imperial college reactor centre legacy waste using gamma-ray spectrometry

Waste characterisation is a principal component in waste management strategy. The characterisation includes identification of chemical, physical and radiochemical parameters of radioactive waste. Failure to determine specific waste properties may result in sentencing waste packages which are not compliant with the regulation of long term storage or disposal. This project involved measurement of intensity and energy of gamma photons which may be emitted by radioactive waste generated during decommissioning of Imperial College Reactor Centre (ICRC). The measurement will use High Purity Germanium (HPGe) as Gamma-ray detector and ISOTOPIC-32 V4.1 as analyser. In order to ensure the measurements provide reliable results, two quality control (QC) measurements using difference matrices have been conducted. The results from QC measurements were used to determine the accuracy of the ISOTOPIC software.

07 ISOTOPE AND RADIATION SOURCES↗