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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Making the Case For Safe Storage of Used Nuclear Fuel For Extended Periods of Time: Combining Near-Term Experiments and Analyses with Longer-Term Confirmatory Demonstrations

The need for extended storage of used nuclear fuel is increasing globally as disposition schedules for used fuel are pushed further into the future. This is creating a situation where dry storage of used fuel may need to be extended beyond normal regulatory licensing periods. While it is generally accepted that used fuel in dry storage will remain in a safe condition, there is little data that demonstrate used fuel performance in dry storage environments for long periods of time. This is especially true for high burnup used fuel.

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

An Experimental Feasibility Study on Applying Neutron Tomography to Encapsulated Spent Nuclear Fuel - 20050

Visual inspection makes easier to ensure the integrity and safety of spent nuclear fuel (SNF) than any classical techniques. Various classical techniques have been applied but there are no reliable methods to qualitatively and quantitatively verify spent fuel in dry storage. Thus, the present authors have developed the prototype safeguards apparatus for dry storage employing the array of He-4 gas scintillation detectors (S670E, Arktis Radiation Detectors Ltd., Switzerland), newly designed to simultaneously measure thermal and fast neutrons without any moderators. The S670E detector has a cylindrical shape with a diameter of 52 mm and active length of 600 mm (total length: 875 mm). The detector is filled by He-4 gas with an approximate pressure of 180 bar for fast neutron detection, and its inner wall is coated by Li-6 for thermal neutron detection. The scintillation lights generated via Li-6 nuclear reaction and elastic scattering are collected by 24 SiPMs linearly paired at the center of the detector. The detector delivers a TTL (Transistor-Transistor Logic) output for pulse readout and UART (Universal Asynchronous Receiver Transmitter) for device control. In order to assess feasibility of the apparatus, an experimental system has been designed, built, and optimized via computational studies. Cf-252 neutron sources and linearly arrayed detectors, working as a single detector, were occupied for this study due to the difficulties in working with SNF. The laboratory scale cask (diameter: 0.67 m, height: 1.5 m), minimized by a factor of 10 compared to the actual thickness of a commercial TN-32 cask, was also manufactured. The detector array was designed to rotate the lab-scale cask and obtain 36 image profiles at every 10 degrees. All profiles were aligned in single frame image called a sinogram, and the cross-sectional image was then fabricated by the inverse radon transform algorithm. These experiments have been repeated with different configurations and numbers of sources. Some gamma-ray sources were also measured with neutron sources in order to distinguish between neutron and gamma-ray pulses. Basically, a He-4 detector is designed to run on Linux OS so it is difficult to directly apply to Windows-based equipment widely used in S. Korea. Therefore, a new data acquisition board working on Windows OS was designed and built. The board mainly consists of FPGA (Field Programmable Gate Array) and SoC (System on Chip) for TTL pulse readout, sorting measured data, and transferring data to a user interface. In conclusion, the tomographic system at lab-scale has shown considerable potential to detect a partial or gross defect of encapsulated assemblies in dry storage. Next steps of this study will be to 1) repeatedly carry out experiments to demonstrate scientific reliability and validity, and 2) numerically integrate signals with weight factors to enhance the image quality since the suggested system based on passive interrogation method requires longer measurement time. Finally, the system will apply to a commercial dry storage phased out soon in S. Korea. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Resource Conservation and Recovery Act General Contingency Plan for Hazardous Waste Treatment, Storage, and Disposal Units at the Y-12 National Security Complex. Revision 12

This document, Y/TS-1015/R12, Resource Conservation and Recovery Act General Contingency Plan for Hazardous Waste Treatment, Storage, and Disposal Units at the Y-12 National Security Complex, is for reference purposes. It is a consolidation of the August 10, 2018, versions of Attachment 6, “Contingency Plan,” from Tennessee Hazardous Waste Management Permits TNHW-122 and TNHW-127. It is not intended to supersede any requirements in these permits and any differences are unintentional. The decision to implement this contingency plan, as part of the Emergency Plan, depends upon whether an imminent or actual incident involving a RCRA treatment, storage, or disposal unit could threaten human health or the environment.

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DOE OSTI2013

Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Evaluation of Electrical Resistivity Tomography to Monitor the Transport of Past Releases Beneath Tank Farms

Underground storage tanks at the Hanford Site, in southeastern Washington State, hold radioactive waste generated from four decades of plutonium production. The 149 single-shell tanks and the 28 double-shell tanks have all exceeded their initial design life of approximately 25 years. At least 67 tanks are assumed to have leaked in the past, resulting in radioactive releases into the vadose zone. Gamma ray logging within dry monitoring wells is currently the primary method for tracking the migration of leaked tank waste through the vadose zone. While this approach provides an accurate assessment of radioactive contamination, that information is only provided near (within ~1m) the borehole, leaving most of the vadose zone unmonitored, particularly the important region directly beneath the tank. This report describes a numerical study that investigates the feasibility and performance of time-lapse 3D electrical resistivity tomography (ERT) for long-term monitoring of a hypothetical tank waste location and migration through the vadose zone. ERT is a method of remotely imaging the bulk electrical conductivity (EC) of the subsurface, which is significantly impacted by the presence of conductive solid and liquid tank waste. The release of liquid tank wastes increases subsurface fluid conductivity and saturation over time, creating a target to use time-lapse ERT for long-term monitoring. Although the presence of metallic infrastructure can cause ERT interference, recent advancements in ERT data processing enable the deleterious effects of buried metallic infrastructure (e.g. pipes, wellbore casings, tanks) to be removed to better determine the liquid tank waste migration over time. Three hypothetical realistic scenarios were simulated in the ERT evaluation. The first two scenarios assume the same leak amount and rate (i.e., between 1/1/1951 and 12/31/1951 at the rate of 347 m 3 per year) but different leaky tanks. Scenario 1 assumes leaks under tank B-102, which is located on the edge of the B-tank farm and surrounded by a few metallic infrastructure including cased pipes/wells/tanks. Scenario 2 assumes leaks under tank B-108, which is located near the center of the B-tank farm and surrounded by larger amount of metallic infrastructure than B-102. Scenario 3 assumes the same metallic infrastructure as B-102, with a more recent contaminant leak that was simulated to have occurred between 1/1/2018 and 12/31/2023 at a rate of 1.89 m 3 per year. The leak time in Scenarios 1 and 2 corresponds to a historical overfill event in 1951 and Scenario 3 corresponds to a recent found tank leak in 2019. In each scenario, a “true” bulk EC model vs. time reflecting contaminant migration was generated. ERT data was simulated from these “true” bulk EC models and a time-lapse ERT inversion produced “imaged” bulk EC vs. time. Three electrode configurations in two, four and eight boreholes surrounding the leak tank were used in the ERT simulations in each scenario. These borehole configurations were considered logistically feasible and cost-effective for monitoring. The hypothetical ERT boreholes are assumed to have non-metallic casing. By comparing the “imaged” bulk EC with the “true” bulk EC, it was demonstrated that the three configurations of wells used (two, four, and eight wells) were able to successfully monitor the migration of tank leaks through the vadose zone, with bulk EC resolution increasing with the number of down borehole ERT arrays for all scenarios. Therefore, the use of eight boreholes to perform ERT monitoring beneath the tanks provided the best spatiotemporal information.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Assessment of Molybdate Based Corrosion Inhibitor for use at the Savannah River Tank Farm Cooling Water System

The Savannah River Site (SRS) has relied on 51 underground storage tanks, many dating back to the 1950s, to store radioactive liquid waste generated from nuclear processing and radionuclide production. These tanks use carbon steel cooling coils, cooled by soft, acidic water, to dissipate the decayed heat produced by radioactive waste. For decades, following historical industry practices, chromate has been added to the cooling water (~450 ppm at pH 9-11) to effectively inhibit corrosion, ensuring cooling coil longevity. Inspections of failed coils reveal that chromate-treated surfaces are generally well-preserved.

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

Bare Fiber Bragg Gratings embedded into concrete buffer Supercontainer concept for nuclear waste storage [ANIMMA--2015-IO-337]

Nuclear power plants have been generating electricity for more than 50 years. In Belgium, 55% of the current energy supply comes from nuclear power. Providing for the safe storage of nuclear waste, including spent fuel (SF) and vitrified high level radioactive waste (HLW), remains an important challenge in the life cycle of nuclear fuel. In this context, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS) is investigating a reference conceptual design called the Supercontainer (SC) for the packaging of SF and HLW. This conceptual design is based on a multiple-barrier system consisting of a hermetically-sealed carbon steel overpack and a surrounding highly-alkaline concrete buffer. The first one is developed to retain the radionuclides. The two main functions of the buffer are (a) to create a high pH environment around the carbon steel overpack in order to passivate the metal surface and so to slow down the corrosion propagation during the thermal phase and (b) to provide a radiological shielding during the construction and the handling of the Supercontainer. A recent test has been performed to investigate the feasibility to construct the SC. This test incorporated several kinds of sensors including Digital Image Correlation (DIC), Acoustic Emission (AE), corrosion sensing techniques and optical fibers with and without fiber Bragg gratings (FBGs). In particular, several single-mode optical fibers with 4 mm long FBGs with different Bragg wavelengths and distributed along the optical fibers were used. For casting and curing condition monitoring, a number of gratings were incorporated inside the concrete buffer during the first stage of construction. Then other sensors were embedded near a heat source installed in the second stage to simulate the effects of heat generated by radioactive waste. The FBGs were designed to measure both temperature and strain effects in the concrete. To discriminate between these effects special packaging was used for some sensors that were installed very close to the unpackaged ones. Sensors placed in plastic tubes have reduced sensitivity to strain, while the ones inserted in metal tubes are only temperature sensitive and their readings can be directly compared with those obtained from thermocouples located nearby. In addition to monitoring temperature and strain behaviour, embedding also had as objective to determine the impact of the high alkaline environment on the silica fibers over a very long time. This article presents the preliminary results obtained with the different FBGs and provides recommendations for future improvement. (authors)

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DOE OSTI2015

The decommissioning imperative - motivating progress in us used fuel management - 15185

When the Yucca Mountain project was terminated in 2010 there was approximately 3500 metric tons of used fuel stored at Independent Spent Fuel Storage Installations (ISFSIs) at which there was no operating reactor (10 shutdown plant sites and the GE Morris facility). By 2015, four additional sites with five reactors will have shut down, causing the amount of used fuel 'stranded' at sites where there is no operating reactor to nearly double to approximately 6900 metric tons which will be stored at a total of 15 sites. At each of these sites, once the reactors are decommissioned, the only thing preventing the land from being returned to communities for other uses is the remaining presence of the used fuel. The cost of maintaining 15 stand-alone ISFSIs is born by electric ratepayers and taxpayers. Going forward, the costs of maintaining stand-alone ISFSIs is certain to grow even if the number of stand-alone ISFSIs does not grow beyond the already announced shutdown of the Oyster Creek reactor in 2019 (which would increase the amount of stranded used fuel by another 10% and create a 16. stand-alone ISFSI). One reason for this cost escalation will be the need for ISFSI owners to renew storage licenses, and implement aging management programs to support these renewals, due to the continued delay in the federal program for removing the used fuel. Storage licenses at 10 of 16 stand-alone ISFSIs will expire by 2021, which is the earliest projected date at which the Department of Energy (DOE) has indicated it could have a pilot interim storage facility available - and nearly 3 decades before DOE projects a repository will be available. The burden and cost of continuing to renew storage licenses - and manage aging storage facilities - at 16 stand-alone sites instead of moving stranded used fuel to a single consolidated storage location or repository will likely be significant. Industry resources are increasingly being channeled into dry storage aging management activities as public and regulator interest in potential long-term degradation mechanism continues to grow. Whether the ultimate destination for used fuel is Yucca Mountain or some other location, the current stalemate preventing progress in the federal program to move it must be broken - and the increasing prominence of the decommissioned reactor quandary may very well be a pivotal factor in breaking that stalemate. (authors)

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

Potential Cost Implications of an Interim Storage Facility for Commercial SNF+ - 17132

The question of whether or not consolidated interim storage of commercial spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. This paper summarizes an evaluation of the cost implications of incorporating a consolidated interim storage facility (ISF) into the waste management system (WMS). In this study, the order-of-magnitude estimates of total system costs were calculated and tabulated. The analyzed scenarios involve shipment of SNF from reactors in dual-purpose storage and transportation canisters (DPCs) currently being used by utilities for dry storage at reactor sites. A number of pertinent conclusions can be drawn from this evaluation, including: - Delay in repository availability increases total system costs. Any delay in opening a repository increases total system costs, regardless of whether the system has an ISF or not. This is due to the increased cost associated with an extended duration for storage, whether at multiple independent spent fuel storage installations (ISFSIs) or at an ISF, until waste can be disposed of.- There is a potential total system lifecycle cost avoidance in realistic scenarios with an ISF when compared to scenarios with no ISF. However, most of the cost avoidance occurs several decades after the ISF is opened. The total WMS cost differential over the long term is mainly attributed to the reduced operational costs of storing the fuel in a consolidated facility rather than at individual reactor sites. - Earlier establishment of an ISF allows for more avoidance of post-shutdown at-reactor storage costs for any repository opening date. An ISF allows earlier acceptance of fuel from reactors, which reduces at-reactor costs from a total system perspective. - Transportation costs have little impact on a WMS with or without an ISF. These impacts range from 3-11% of the total cost in all scenarios. Therefore, transporting the fuel twice does not appear to be a significant cost concern driver relative to other system costs. In conclusion, designing, licensing, and constructing an ISF will require a large near-term financial investment. However, reduced annual operating and maintenance costs for an integrated waste management system with an ISF, compared to a system without an ISF, could eventually result in system-wide (integrated) cost avoidance due to reductions in at-reactor storage costs in the long term. The largest cost avoidances would not occur for several decades, so assumptions about inflation, escalation, and discount rates will have a significant effect on potential economic impacts of an ISF. It should be noted that other benefits associated with the investment of an ISF-such as earlier removal of SNF from shutdown reactor sites, earlier prospects for site re-utilization, and additional overall integrated waste management system flexibility-were not addressed in this study. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Spent Nuclear Fuel Receipt Rate Analysis within an Integrated Waste Manage-ment System Architecture that Includes Consolidated Interim Storage

A key parameter in analyzing the performance of an integrated waste management system (IWMS) architecture for the disposition of spent nuclear fuel (SNF) is the SNF receipt rate from reactor and other custodian sites. The introduction of one or more federal consolidated interim storage facilities (CISFs) into the IWMS architecture can enable the receipt rate profile as a function of time to be accelerated relative to system architectures without a CISF. The question then arises as to what an optimal SNF receipt rate profile for an IWMS architecture might be in view of practical constraints and desired system performance attributes and associated metrics. This paper describes a sensitivity study on SNF receipt rates and the associated results for a selected set of IWMS scenarios aimed an in-forming near-term planning for interim storage capabilities and transportation assets. Two different strategies are compared, one that fills an initial CISF quickly and then idles the transportation system while a disposal system is prepared, and a second strategy that aims to provide a more continuous use of transportation assets and receipt capabilities at the IWMS while the disposal system is readied for SNF receipt. Cost considerations and other factors such as impact on timing of clearing reactor sites of SNF, efficient use of capital assets, and other metrics, including those which may be important to a CISF host community, are examined. Based on the analysis, an initial approach is presented targeting a continuous receipt strategy while having the flexibility to step up receipt capabilities to a reasonable degree when needed and beneficial within overall system constraints.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Design and Validation Testing of TruckScan to Assay Large Sacks of Fukushima Radioactive Debris on a Truck

As a result of the March 2011 earthquake and resulting tsunami in Japan, there was a serious accident at the Fukushima Dai-ichi Nuclear Power Plant. This accident has contaminated soil and vegetation in a wide area around the plant. Decontamination projects over the last 4 years have resulted in large numbers of 1 cubic meter canvas bags of debris, commonly called Super Sacks [SS]. These are currently stored nearby where they were generated, but starting in 2015, they will be moved to various Interim Storage Facilities [ISF]. Trucks will typically carry 8-20 of these SSs. When the trucks arrive at the ISF they need to be rapidly sorted into groups according to radioactivity level, for efficient subsequent processing. Canberra Industries, Inc. [CI] has designed a new truck monitoring system 'TruckScan' for use at these ISFs. The TruckScan system must measure the entire truck loaded with multiple closely packed SSs, and generate a nuclide specific assay report showing the radioactivity in each individual SS. The Canberra-Japan office, along with Obayashi Corporation have performed validation testing to demonstrate to the regulatory authorities that the proposed technique was sufficiently accurate. These validation tests were conducted at a temporary storage area in Fukushima prefecture. Decontaminated waste of various representative types and of various levels of radioactivity was gathered and mixed to create homogeneous volumes. These volumes were sampled multiple times and assayed with laboratory HPGe detectors to determine the reference concentration of each pile. Multiple SSs were loaded from each pile. Some of the SSs were filled 50% full, others 75% full, and others 100% full, to represent the typical loading configuration of the existing SSs in the field. The content of the SSs are either sand, soil, or vegetation with densities ranging from 0.3 g/cc - 1.6 g/cc. These SSs with known concentrations of Cs-134 and Cs-137 were then loaded onto trucks in a variety of configurations, typical of how they might be on the real trucks. A partial system was installed at the site and used to assay these trucks with the various loading configurations. Whereas the full system will have 8 collimated 3 x 3'' NaI detectors, the test system only had two detectors; therefore the truck was moved and counted 4 times. The data were acquired and analyzed with the Canberra Genie software to determine the peak counts for both Cesium nuclides. That data was then analyzed with a prototype version of a Maximum Entropy algorithm, to determine the individual SS activity. The goal of the validation tests was to demonstrate that the system could detect 1000 Bq/kg in 15 seconds, and to determine how accurately it could quantify individual SSs. The validation tests demonstrated that the product would perform as predicted. The TruckScan results were consistent with the sample assay results [y = 1.0029 x, R{sup 2} = 0.914]. The Total Propagated Uncertainty, including both uncertainties from these tests and others that were estimated but not tested was 16.6% percent. (authors)

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

Progress in Defining the UK Highly Active Storage Tanks POCO Strategy - 16134

In the UK Highly Active (HA) waste is stored in stainless steel tanks prior to encapsulation in glass in one of three operating vitrification plants. Significant progress has been made in recent years towards the reduction in the volume of HA Liquor (HAL) stored in these tanks and a number of the tanks are now operating at heel levels. As a result, a strategy is being developed to carry out Post Operational Clean Out (POCO) of some tanks while reprocessing operations continue, thus enabling accelerated and more cost effective clean-up of these HA facilities. This paper, and subsequent conference presentation, will highlight some of the key aspects of this strategy, including a description of the challenges and progress made so far to help facilitate the clean-up mission without having a detrimental effect on ongoing reprocessing operations. (authors)

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

Extended Storage of High Burn-up Fuel Cask R and D Project - 16591

Under contract by EPRI and DOE, AREVA Federal Services LLC (AFS) is leading a team consisting of AREVA Inc. and Dominion Virginia Power (Dominion) to develop and conduct a full-scale confirmatory Research project with HBU fuel. A test plan was developed using an existing AREVA Transnuclear (TN) TN-32 bolted storage cask with a modified, specially instrumented lid to be engineered and then loaded with multiple cladding types of HBU pressurized water reactor (PWR) UNF from Dominion's North Anna Power Station (this modified cask is identified as TN-32B). Seven specially designed thermocouple lances, each containing nine detectors, will be inserted into the designated fuel assembly guide tubes through the modified cask lid. The loaded cask with the monitoring system will be placed into dry storage on the existing North Anna Independent Spent Fuel Storage Installation (ISFSI) pad for a period of extended monitoring. In addition, design modifications to the lid have been incorporated to permit periodic gas sampling of the cask internals during the storage period on the ISFSI pad. The fuel selected for this research project ranges in average assembly burnup between 50 GWD/MTU to 67 GWD/MTU. The instrumentation used to collect the thermal data is very similar to in-core thermocouple detectors used at nuclear power plants and is considered very reliable for this application. Prior to cask loading at North Anna, a series of 'fit up' tests will be performed at the fabricator to ensure proper alignment, fit and function of the modified cask and related instrumentation and equipment. (authors)

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

Finite Element Structural Analysis Evaluating New Retrieval Strategies in the Hanford Waste Tanks 1 - 19125

Several waste retrieval methods being considered for the Hanford underground single- and double-shell nuclear waste storage tanks would require changes to the equipment loads above the tanks or structural modifications to the tanks themselves. Detailed structural analysis has been used in the planning phase to ensure that the reinforced concrete tanks continue to meet structural design codes in the modified configuration of these waste retrieval options. The feasibility of installing new single and multiple large risers penetrating the tank dome was evaluated using this approach. This paper summarizes the structural analysis methods and design codes used to evaluate the introduction of new unreinforced penetrations in a single-shell tank dome. The anticipated equipment loads during current and future waste retrieval operations are evaluated to ensure the continued structural integrity of the tank under thermal, dead weight, and seismic loads. (authors)

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

Integrated salt waste processing at the Savannah River Site - 15301

Treatment and disposition of salt waste is the critical path to completion of the Savannah River Site Liquid Waste (LW) Disposition and Cleanup Program. The new Salt Waste Processing Facility (SWPF) is a critical link in the pre-treatment chain and will process high curie salt waste removed from various SRS liquid waste storage tanks. Significant infrastructure modifications are needed in the existing LW facilities in order to integrate the SWPF into the LW system. (authors)

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

Case Study of Urban Residential Remediation and Restoration in Port Hope, Canada - 13250

The Canadian Municipality of Port Hope, Ontario, is located some 100 km east of Toronto and has been the location of radium and/or uranium refining since the 1930's. Historically, these activities involved materials containing radium-226, uranium, arsenic and other contaminants generated by the refining process. In years past, properties and sites in Port Hope became contaminated from spillage during transportation, unrecorded, un-monitored or unauthorized diversion of contaminated fill and materials, wind and water erosion and spread from residue storage areas. Residential properties in Port Hope impacted by radioactive materials are being addressed by the Canadian federal government under programs administered by the Low-Level Radioactive Waste Management Office (LLRWMO) and the Port Hope Area Initiative Management Office (PHAIMO). Issues that currently arise at these properties are addressed by the LLRWMO's Interim Waste Management Program (IWM). In the future, these sites will be included in the PHAIMO's Small Scale Sites (SSS) remedial program. The LLRWMO has recently completed a remediation and restoration program at a residential property in Port Hope that has provided learnings that will be applicable to the PHAIMO's upcoming SSS remedial effort. The work scope at this property involved remediating contaminated refinery materials that had been re-used in the original construction of the residence. Following removal of the contaminated materials, the property was restored for continued residential use. This kind of property represents a relatively small, but potentially challenging subset of the portfolio of sites that will eventually be addressed by the SSS program. (authors)

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

Instrumentation. Nondestructive Examination for Verification of Canister and Cladding Integrity - FY2013 Status Update

This report documents FY13 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) characterize hydrogen effects in UNF cladding to facilitate safe storage and retrieval.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Computational scheme for burnup credit applied to long term waste disposal - 14448

The application of burnup credit (BUC) to criticality safety evaluations (CSE) of geological repositories for long term disposal of spent nuclear fuel (SNF) is under consideration in Switzerland. Therefore, in collaboration with the national waste management organization, PSI is conducting research towards the development of a BUC methodology tailored for Swiss specific applications. As part of this research, a computational sequence from the in-core burnup phase, through the interim storage, until final disposal is being established. The tools selected for this methodology, referred to as BUCSS-R (Burnup Credit System for Swiss reactors and Repositories), are CASMO for the in-core fuel depletion, the decay module of SERPENT for the decay calculations, and MCNP for the criticality calculations. In connection to BUCSS-R, a computational interface to validated reference core models of the Swiss reactors is being developed. The purpose is to apply reconstructed depletion histories based on real core operation conditions with CASMO assembly calculations, to transfer the pin wise nuclide compositions at discharge to the subsequent SERPENT decay calculations, and afterwards to the final MCNP criticality calculations. An objective of this paper is to present the elements of the BUCSS-R methodology, and to show a first application to so-called bounding analyses of a preliminary Swiss disposal canister design loaded with fuel assembly types operated in a Swiss PWR reactor. The rationale for these bounding analyses is that before starting the integration of methodologies for bias and uncertainty quantifications, the span of the k-eff variation between fresh and burnt fuel loadings should first be quantified in a realistic manner. The obtained results are reported in this paper together with studies performed in order to evaluate the effects from burnup levels and nuclides credit at flooded conditions on the k-eff values. (authors)

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