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DOE OSTI2016

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Sixth National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

The United States of America (U.S.) ratified the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention) on April 9, 2003. The Joint Convention establishes an international peer review process among Contracting Parties and provides incentives for nations to take appropriate steps to bring their spent fuel and radioactive waste management activities into compliance with general safety standards and practices. An extensive U.S. legal and regulatory structure ensures the safety of spent fuel and radioactive waste management. This report describes radioactive waste management in the U.S. in both the commercial and government sectors, and provides annexes with information on spent fuel and waste management facilities, inventories, and ongoing decommissioning projects. It also provides detailed information on spent fuel and radioactive waste management safety, as well as transboundary movements (imports/exports) and disused sealed sources, as required by the Joint Convention.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Head-end process for the reprocessing of HTGR spent fuel

The reprocessing of HTGR spent fuels is in favor of the sustainable development of nuclear energy to realize the maximal use of nuclear resource and the minimum disposal of nuclear waste. The head-end of HTGR spent fuels reprocessing is different from that of the LWR spent fuels reprocessing because of the difference of spent fuel structure. The dismantling of the graphite spent fuel element and the highly effective dissolution of fuel kernel is the most difficult process in the head end of the reprocessing. Recently, some work on the head-end has been done in China. First, the electrochemical method with nitrate salt as electrolyte was studied to disintegrate the graphite matrix from HTGR fuel elements and release the coated fuel particles, to provide an option for the head-end technology of reprocessing. The results show that the graphite matrix can be effectively separated from the coated particle without any damage to the SiC layer. Secondly, the microwave-assisted heating was applied to dissolve the UO{sub 2} kernel from the crashed coated fuel particles. The ceramic UO{sub 2} as the solute has a good ability to absorb the microwave energy. The results of UO{sub 2} kernel dissolution from crushed coated particles by microwave heating show that the total dissolution percentage of UO{sub 2} is more than 99.99% after 3 times cross-flow dissolution with the following parameters: 8 mol/L HNO{sub 3}, temperature 100 Celsius degrees, initial ratio of solid to liquid 1.2 g/ml. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Tools Developed to Prepare and Stabilize Reactor Spent Fuel for Retrieval from Tile Holes - 12251

Spent fuel from the Chalk River Laboratories (CRL) nuclear reactors is stored in the waste management areas on site. This fuel is contained within carbon steel spent fuel cans that are stored inside vertical carbon steel lined concrete pipes in the ground known as tile holes. The fuel cans have been stored in the tile holes for greater than 30 years. Some of the fuel cans have experienced corrosion which may have affected their structural integrity as well as the potential to form hydrogen gas. In addition to these potential hazards, there was a need to clean contaminated surfaces inside of and around the exposed upper surface of the tile holes. As part of the site waste management remediation plan spent fuel will be retrieved from degraded tile holes, dried, and relocated to a new purpose built above ground storage facility. There have been a number of tools that are required to be developed to ensure spent fuel cans are in a safe condition prior to retrieval and re-location. A series of special purpose tools have been designed and constructed to stabilize the contents of the tile holes, to determine the integrity of the fuel containers and to decontaminate inside and around the tile holes. Described herein are the methods and types of tools used. Tools that have been presented here have been used, or will be used in the near future, in the waste management areas of the CRL Site in preparation for storage of spent fuel in a new above ground facility. The stabilization tools have been demonstrated on mock-up facilities prior to successful use in the field to remove hydrogen gas and uranium hydrides from the fuel cans. A lifting tool has been developed and used successfully in the field to confirm the integrity of the fuel cans for future relocation. A tool using a commercial dry ice blaster has been developed and is ready to start mock-up trials and is scheduled to be used in the field during the summer of 2012. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

The Recommendation of Public Engagement Commission on Spent Nuclear Fuel in Korea - 16074

The Public Engagement Commission on Spent Nuclear Fuel Management was lunched and embarked on public engagement activities with an aim of presenting recommendations on the method for managing spent fuel to the government by June 2015. Therefore, a national policy, strategy and management plan for spent fuel will be presented in detail in the Basic Plan for Radioactive Waste Management where the results of public engagement activities will be incorporated. In consideration of Korea's circumstances, permanent disposal facilities should be constructed and operated with the goal of 2051. For this reason, Underground research laboratory (URL) in the same area and on-site and site conditions of disposal facilities by selecting the site and construction process in 2020 should be started and the empirical study should be initiated in 2030. This study is expected to transfer experience of and lessons from determining national policy to countries that have not established their final management policy for spent fuel, based on Korea's solutions to shortage of spent fuel pools and the formation of national consensus. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Quantity and management of spent fuel from prototype and research reactors in Germany

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on the information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)

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DOE OSTI2014

An approach to determine a defensible spent fuel ratio.

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. However, the validity of these large-scale results remain in question due to the lack of a defensible Spent Fuel Ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980 s and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000 s. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort is urgent in order to secure the only known available SNF samples with a clearly defined path to disposal.

Durbin, Samuel G.↗

DOE OSTI2015

Multilateral Spent Fuel Management Efforts: Nonproliferation and Energy Implications in the Asia Pacific

Management of spent nuclear fuel in the Asia Pacific will be a growing challenge in coming decades as the region continues its rapid economic and nuclear energy expansion. Accompanying this growth has been skyrocketing energy demand. The capacity of nuclear energy in the Asia Pacific is expected to increase by approximately 50% by 2035. In fact, of the 66 reactors under construction worldwide, 47 are being built in the Asia Pacific region alone. In a sense, the area is experiencing the so-called ‘Nuclear Renaissance’, which could simultaneously increase proliferation risks while fueling economic growth and political stability. And because ‘Asia Pacific’ includes North America, Oceania, East Asia, Southeast Asia and, by some definitions, South and Central Asia as well as the Middle East, the demand for nuclear energy in particular is expected to continue. Consequently, the Asia Pacific region will soon face important decisions about how to support and implement the necessary infrastructure for nuclear energy expansion and deal with the waste, while also meeting all nonproliferation obligations under the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). One option for dealing with spent nuclear fuel that should be considered is multilateral spent fuel management. Such an effort could help reduce risks associated with safety, proliferation and security of nuclear energy. While the original idea of multilateral spent nuclear fuel management is a decades-old concept, no official plan has been developed because of challenges related to cost, liability, transportation, export controls, international commerce, and nonproliferation, to name a few. Nonetheless, as more and more newcomer countries decide to develop nuclear energy, but do not yet have a strategic plan or the financial means to deal with the waste, the international community may need to take a more proactive approach in developing a multilateral effort, particularly with respect to geological repositories and interim dry storage. Risks and benefits will need to be analyzed, while the role of such an effort in the revival of the Nuclear Renaissance will need to be evaluated. This paper provides an overview of spent nuclear fuel management efforts in the Asia Pacific region and attempts to analyze the possible benefits and feasibility of a multilateral approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Numerical Estimation of the Spent Fuel Ratio

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Integrating Management of Spent Nuclear Fuel from Generation to Disposal

The current management system in the United States for commercial spent nuclear fuel does not emphasize integration among storage, transportation, and disposal. Unless a path can be implemented that addresses the long-term needs for integration, the United States could end up leaving substantial quantities of stranded commercial spent nuclear fuel stored at decommissioned reactor sites in an increasingly wide variety of containers. This lack of integration does not cause safety issues, but may delay transporting the spent fuel and complicate options for permanent disposal. The large containers now in use for dry storage remain at high temperatures for decades, thereby delaying transportation from decommissioned reactors. The large containers also have no easy path to disposal unless (1) disposal is further delayed (up to 150 years or more for some mined repository concepts); (2) the contents are repackaged into smaller, cooler packages; or (3) the high temperatures are used as de facto site-selection and design criteria for a repository. Implementing consolidated interim storage could address many issues that exist because of this lack of integration. A consolidated interim storage facility that includes appropriate capabilities can allow existing disparate parts to integrate as a system. Previous agencies and commissions have noted this theme before as a way to provide flexibility in the waste management system. This report uses the rationale for such an approach as a framework to discuss the complexities of reconfiguring the waste management system to include consolidated storage. However, concerns that increased storage capacity will reduce the national urgency for a repository are unavoidable, and continued effort will be necessary in public dialogues on the societal aspects of moving commercial spent nuclear fuel into consolidated interim storage. A single optimal solution for integrating current storage and planned transportation with disposal is unlikely. Rather, efforts to integrate various phases of spent fuel management should begin promptly and continue throughout the remaining life of the current fuel cycle. These efforts will need to adapt continuously to evolving circumstances.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Thermal Modeling of the TN-32B Cask for the High Burnup Spent Fuel Data Project

This report describes thermal modeling done for the High Burnup Dry Storage Research Project that is being performed by the U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) research and development (R&D) program. The purpose of this project to investigate the performance of high-burnup spent nuclear fuel in dry storage. Part of this project is a demonstration test being performed with a storage module in the North Anna Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an Orano TN-32B High Burnup cask (Fig. S-1) which has been loaded with fuel assemblies from the North Anna spent fuel pool. The main goals of this test are to provide confirmatory data for model validation and potential improvement, support license renewals and new licenses for ISFSIs, and support transportation licensing for high burnup spent nuclear fuel (EPRI 2014).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Validation of Spent Fuel Dose Rate Calculations Using NDA Measurement Campaign Data

The predicted radiation gamma dose rate is a key parameter to determine if a spent nuclear fuel assembly is self-protected from misuse as well as an indicator for worker/inspector safety. Recent studies showed that using state-of-the-art techniques for dose rate calculations, including the one adopted for this study, the numerical calculations of the gamma dose rates are three times lower than those from the most commonly-cited reference values for spent LWR fuel used by domestic and international agencies. With only a few publically available experimental data points worldwide for this type of benchmark study, the gamma flux measurements from the Spent Fuel Non Destructive Assay (NDA) measurement campaign can be used to help validate the physics approach used in calculating gamma dose rates. Specifically, this data would validate the gamma flux calculation from spent fuel, which is a key component of these multi-physics techniques for dose rate assessments; dose rate measurements were not part of the NDA campaign. These data are primarily the passive gamma measurements of the spent nuclear fuel assemblies from the Swedish Nuclear Fuel and Waste Management Company (SKB).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Corrosion-Resistant Coatings on Spent Nuclear Fuel Canisters to Mitigate and Repair Potential Stress Corrosion Cracking (FY22 Status)

This report summarizes the activities performed by Sandia National Laboratories in FY22 to identify and test coating materials for the prevention, mitigation, and/or repair of potential chloride-induced stress corrosion cracking in spent nuclear fuel dry storage canisters. This work continues efforts by Sandia National Laboratories that are summarized in previous reports in FY20 and FY21 on the same topic. The previous work detailed the specific coating properties desired for application and implementation to spent nuclear fuel canisters (FY20) and identified several potential coatings for evaluation (FY21). In FY22, Sandia National Laboratories, in collaboration with four industry partners through a Memorandum of Understanding, started evaluating the physical, mechanical, and corrosion-resistance properties of 6 different coating systems (11 total coating variants) to develop a baseline understanding of the viability of each coating type for use to prevent, mitigate, and/or repair potential stress corrosion on cracking on spent nuclear fuel canisters. This collaborative R&D program leverages the analytical and laboratory capabilities at Sandia National Laboratories and the material design and synthesis capabilities of the industry collaborators. The coating systems include organic (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin), organic/inorganic ceramic hybrids (silane-based polyurethane hybrid and a quasi-ceramic sol-gel polyurethane hybrid), and hybrid systems in conjuncture with a Zn-rich primer. These coatings were applied to stainless steel coupons (the same coupons were supplied to all vendors by SNL for direct comparison) and have undergone several physical, mechanical, and electrochemical tests. The results and implications of these tests are summarized in this report. These analyses will be used to identify the most effective coatings for potential use on spent nuclear fuel dry storage canisters, and also to identify specific needs for further optimization of coating technologies for their application on spent nuclear fuel canisters. In FY22, Sandia National Laboratories performed baseline testing and atmospheric exposure tests of the coating samples supplied by the vendors in accordance with the scope of work defined in the Memorandum of Understanding. In FY23, Sandia National Laboratories will continue evaluating coating performance with a focus on thermal and radiolytic stability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2014

Effects of Spent Cooling and Swirler Angle on a 9-Point Swirl-Venturi Injector

This paper presents multipoint Lean-Direct-Injection (LDI) emissions results for flame tube combustion tests at an inlet pressure of 1034 kPa and inlet temperatures between 835 and 865 K; these are the combustor inlet conditions that the High Speed Research (HSR) program used for supersonic cruise. It focuses on one class of LDI geometry, 9-point swirl-venturi LDI (SV-LDI). Two parameters are compared in this paper: the use of dome cooling air and the swirler blade angle. Dome cooling air is called "spent cooling" and is at combustor inlet conditions. Three cooling variations are studied: cooling at the venturi throat, cooling at the dome face, and no cooling at all. Two swirler blade angles are studied: 45deg and 60deg. The HSR 9-point SV-LDI emissions are also compared to a similar 9-point SV-LDI design which was used in the later ultra-efficient engine technology (UEET) program. The HSR and UEET designs cannot be compared directly due to different UEET combustor conditions. Therefore, this paper uses previously published UEET correlation equations to make comparisons. Results show that using a 45deg swirler produces lower NOx emissions than using a 60deg swirler. This is consistent with the later UEET results. The effects of spent cooling depend on swirler angle, spent cooling location, and the test conditions. For the configuration with 45deg swirlers, spent cooling delivers lower NOx emissions when it is injected at the throat. For the 60deg swirler, spent cooling does not have much effect on NOx emissions. These results might be caused by the location and the intensity of the flame recirculation zone.

He, Zhuohui J.↗

NASA NTRS2013

Effects of Spent Cooling and Swirler Angle on a 9-point Swirl-Venturi Injector

This paper presents multipoint lean-direct-injection (LDI) emissions results for flame tube combustion tests at an inlet pressure of 1034 kPa and inlet temperatures between 835 and 865 K; these are the combustor inlet conditions that the High Speed Research (HSR) program used for supersonic cruise. It focuses on one class of LDI geometry, 9-point swirl-venturi LDI (SV-LDI). Two parameters are compared in this paper: the use of dome cooling air and the swirler blade angle. Dome cooling air is called 'spent cooling' and is at combustor inlet conditions. Three cooling variations are studied: cooling at the venturi throat, cooling at the dome face, and no cooling at all. Two swirler blade angles are studied: 45 deg and 60 deg. The HSR 9-point SV-LDI emissions are also compared to a similar 9-point SV-LDI design which was used in the later ultra-efficient engine technology (UEET) program. The HSR and UEET designs cannot be compared directly due to different UEET combustor conditions. Therefore, this paper uses previously published UEET correlation equations to make comparisons. Results show that using a 45 deg swirler produces lower NOx emissions than using a 60 deg swirler. This is consistent with the later UEET results. The effects of spent cooling depend on swirler angle, spent cooling location, and the test conditions. For the configuration with 45 deg swirlers, spent cooling delivers lower NOx emissions when it is injected at the throat. For the 60 deg swirler, spent cooling does not have much effect on NOx emissions. These results might be caused by the location and the intensity of the flame recirculation zone.

He, ZH., Joe↗

DOE OSTI2020

Analysis of spent nuclear fuel multipurpose canister dynamics during rail transportation

The primary mode of spent nuclear fuel transportation within the United States will be by railcar. One such system is the Atlas railcar, which is designed to transport 17 different spent nuclear fuel cask systems, including bare fuel systems and canister fuel systems. In the latter configuration, multipurpose canisters containing spent nuclear fuel may be placed within an overpack for storage, or within a cask for transportation. Compared to bare fuel systems, canister fuel systems have additional degrees of freedom for motion during transportation, because clearance between the cask and canister allows for some motion of the canister to occur relative to the cask. This work investigates the effect of canister motion on the shock and vibration imparted to the spent nuclear fuel within. Structural dynamic analyses have been conducted to identify the effects of canister to cask clearance, presence and type of dunnage, and loading direction and frequency. This modeling study calculates anticipated cask motion, canister motion, and spent nuclear fuel structural dynamic response to normal conditions of transportation railcar motion using finite element analysis methods that were developed to model the rail segment of the ENSA/DOE (Equipos Nucleares S.A., U.S. Department of Energy) multimodal transportation test of 2017.

SNF, UNF, LS-DYNA, dynamics, kinematics, shock, vi↗

DOE OSTI2015

The challenge: Safe management on spent and disused sealed radioactive source in Thailand - 15654

When sealed radioactive sources are no longer in use, they should be returned back to the country of origin. However, most of them could not be returned to the country of origin, therefore the safety management of spent and disused radioactive source is need. Thailand Institute of Nuclear Technology (TINT) where is as the national waste management center, has the duty to ensure on the safety and the security for long term storage of spent and disused sealed sources before final disposal. The Radioactive Waste Management Center at TINT, is authorized to operate treatment, conditioning and storage of radioactive waste. This paper will describe the proper operational procedures on characterization technique, identification of unknown sources, volume reduction technique, re-packaging, registration and record keeping of spent and disused radioactive source. The successful results exhibit that the record keeping of spent and disused source is developed, and the national inventory of spent and disused source is up to date. This paper would confirm the safety management of radioactive waste in Thailand. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Waste Management Analyses for Spent Structures at SNS - Paper 106

The Spallation Neutron Source (SNS) in Oak Ridge, Tennessee, is an accelerator driven neutron scattering facility for materials research. Presently, the SNS operates at 1.2 megawatt (MW) proton beam power incident on a mercury target with a proton beam energy of 1 GeV and 60 Hz repetition rate. The SNS consists of an accelerator facility, target facility, and a world-class suite of neutron scattering instruments supporting material, life-science and fundamental physics research. According to the SNS operations plan, some components are replaced when they reach their end-of-life due to radiation induced material damage or burn-up, or because of mechanical failure. During operation, these components, especially those in proximity to the target, are exposed to a radiation environment and build up significant activity during their service lifetime. All these components must be safely removed, placed in a container/package for storage, and ultimately transported offsite for disposal. An accurate estimate of the radionuclide inventory is performed assuming realistic irradiation history and decay time in order to classify a spent component. On the basis of a calculated radionuclide inventory, the spent component is classified, and an appropriate cask/package for transport and storage is suggested. After the choice for the container is made, radiation transport calculations are performed for the facility component placed inside the container, to ensure that the container is compliant with waste management regulations. When necessary, additional shielding is added. For each spent component before it goes out of the SNS site, supporting documentation is produced with a radionuclide inventory prediction, dose rate analyses and, if required, additional shielding. A dose rate analyses for the exposure prediction of personnel during components change out are performed as well. Most of the effort is concentrated on the target facility components such as target, proton beam window (PBW) and core vessel insert (CVI) plugs because of severe radiation conditions. Additionally, analyses are performed for the accelerator facility components, which are removed from the facility. Activation analyses and residual dose rates analyses are performed to support safe handling of SNS spent components at the SNS site and during transport to the waste disposal site. On the basis of the calculated radionuclide inventories, spent components characterized and classified with respect to waste management guidelines, and activity-based determinations are made regarding appropriate packaging for transport. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗