SEARCH · Search NASA

Results for “RADIOACTIVE WASTE STORAGE”

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

SourceResearch connectionFrom yearThrough year

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

At least 1,099 records · Page 61

All sources NASA NTRS DOE OSTI

DOE OSTI2016

Inorganic tin aluminophosphate nanocomposite for reductive separation of pertechnetate

Pertechnetate (TcO4-) is the most abundant chemical form of radioactive contaminant 99Tc present in the legacy nuclear waste streams and in the subsurface of the nuclear waste storage sites.One proposed remediation approach is reductive separation of TcO4-and sequestration in low-temperature waste forms. The development of the relevant technologies has been slow due to the lack of reductive materials that retain their functionalityand are otherwise suitable for application in multicomponent or aggressive media such as highly alkaline, brine-like solutions typifying nuclear tank wastes. This research prepared a tin-based reductive material and demonstrated its potential utility for the separation of TcO4- from the alkaline nuclear wastes. This material consists of Sn(II/IV) phosphate supported by the polycrystalline aluminophosphate matrix. The aluminophosphate matrix is inert to the reaction conditions and offers the benefits of high stability and low solubility in the concentrated alkaline solutions. This Sn(II/IV)-basedmaterial exhibitshigh loading capacity for Tc and selectively removed major fraction of TcO4-from the tank waste supernatant simulants, which contains 7.8 M total sodium and 2.43 M free hydroxide concentrations. The observed Kd values for Tc are about 13,000 and 2,200 mL g-1 for the simulant solutions containing no or 33 mM Cr(VI), respectively, positioning Sn(II/IV) aluminophosphate among the best performing reductive sorbents for TcO4- developed to date. This advanced behaviour is attributed to the synergistic combination of the Sn(II/IV) aluminophosphate functionalities. The presence of Sn(II/IV)-rich fibres facilitates the reduction of TcO4- to Tc(IV), which getsimbedded along the fibre branches. Importantly, the Sn(IV)-containing inert polycrystalline matrix alsoincorporates Tc(IV)which triggers its crystallization to rutile SnO2 phase andstabilizesTc(IV) in the polycrystalline matrix.

Environmental Molecular Sciences Laboratory↗

DOE OSTI2012

Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites

The Blue Ribbon Commission on America's Nuclear Future identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Mobile Hot Cell Digital Twin: End-of-life Management of Disused High Activity Radioactive Sources – 23598

Sealed radioactive sources are utilized for a wide range of applications across nuclear facilities, universities, hospitals, and industry. When these sources reach the end of serviceable life, they become waste. As waste, this radioactive material then goes through a process of recapture and then transfer to long term storage. With the advancement of technology in conjunction with better accessibility of technology, industries are exploring the use of digital automation to enhance productivity, efficiency, and safety while minimizing operation and maintenance costs, health and environmental risks, and uncertainty in the project life cycles. One area for exploration is the use of a digital twin to help design a robust, versatile, and safe solution for recapturing spent sources. We believe this avenue can also provide further advantages in the operations aspect of end-of-life management of spent radioactive sources by reducing the deployment time, increasing operator safety, and reducing operation & maintenance cost. We present a novel digital twin framework for end-of-life management of disused high activity radioactive sources. We have designed and developed a framework that houses a digital twin for visualizing and monitoring the recapture process to inform the engineering and design of a new Mobile Hot Cell. Furthermore, we demonstrate the feasibility of the proposed framework by providing a prototypical implementation, supporting the Mobile Hot Cell and human-machine interface's virtual replication.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2015

The Bukov underground research facility as part of the Czech Republic's DGR development programme - 15644

The Czech Republic's deep geological repository (DGR) policy sets out a number of important milestones including the selection of candidate sites by 2018 and the selection of a final site by 2025. The final commissioning of the Czech DGR is planned for 2065. In 2010 a screening process was completed employing geological and other criteria that resulted in the identification of 7 potentially suitable sites, all situated in crystalline - granite host rock environments. The next step of the process will result in a reduction both in the number of sites and the surface area of remaining sites based on preliminary safety evaluations and socioeconomic, political and environmental criteria. In order to better understand the behaviour of crystalline host rock environments for DGR safety case purposes, the Radioactive Waste Repository Authority (RAWRA) made decided to enhance the extent of in-situ measurements and research at a depth corresponding to that of the future DGR. In the second half of 2013, RAWRA commenced the construction of the Bukov Underground Research Facility (Bukov URF) in the southern section of the Rozna uranium mine at a depth of 600 m below the earth's surface in a geologically well-mapped area of a Moldanubic crystalline rock formation which is composed principally of migmatised gneisses, migmatites and granulites. The Bukov URF will be, first and foremost, a research facility at which work will be conducted focusing on the taking of geophysical measurements, petrographic, mineralogic, geochemical research and the in-situ determination of the physical - mechanical, geochemical and hydrogeological properties of the host rock massif. The research results will form input data which will be used to both support and improve the safety of the future DGR. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Crystallization of Sodium Phosphate Dodecahydrate and Re-crystallization to Natrophosphate in Simulated Hanford Nuclear Waste - 16188

The nuclear waste at the Hanford site near Richland, WA, has large concentrations of phosphate in the form of the phosphate ion, sodium phosphate dodecahydrate (Na{sub 3}PO{sub 4}.12H{sub 2}O.(1/4)NaOH) and natrophosphate (Na{sub 7}F(PO{sub 4}){sub 3}.19H{sub 2}O). Sodium phosphate dodecahydrate can form a gel and natrophosphate can crystallize large particles, complicating the processing of slurries of both salts. The gel is regarded as more problematic, so natrophosphate has historically been crystallized to prevent phosphate gelling. This study determined if natrophosphate crystals can grow to large size in short time periods (a few weeks), time periods relevant to short process shutdowns. Solutions of NaOH and NaAl(OH){sub 4} were blended at different ratios with stock solutions containing NaOH, NaF, and Na{sub 3}PO{sub 4} at 50 deg. C. The mixtures were allowed to cool to 22 deg. C, and the crystal growth was monitored by Polarized Light Microscopy (PLM) for 17 days. Four of the six blends investigated gelled rapidly due to Na{sub 3}PO{sub 4}.12H{sub 2}O.(1/4)NaOH precipitation. The gel slowly dissipated over time as the solids recrystallized into natrophosphate. In one case, the natrophosphate reached sizes of greater than 1000 microns in diameter in just 4 days. This rapid gelling and crystallization kinetics is important to engineers trying to manage nuclear wastes high in phosphates. Hanford tank farm engineers are developing sampling plans to support temperature and process control strategies for preventing the formation of solid phosphates. They are also developing methods of suspending large natrophosphate crystals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Spent Nuclear Fuel Management and Disposal in Finland - Status and Lessons Learned at the End of 2015 - 16507

In November 2015 the Finnish Government approved and the Parliament ratified Posiva's 2012 License Application for Construction of a mined, deep geological repository for the spent nuclear fuel (SNF) at the Olkiluoto site in the Eurajoki municipality based upon recommendations provided by the Finnish Radiation Safety Authority (STUK), international expert groups and long-standing local support. The preceding siting process began in 1983 and it included disposal-concept selection (KBS-3V), 'consent-based' site selection and large-scale in-situ tests conducted in an underground research laboratory, the ONKALO, situated ∼440 m below the mean Sea level immediately adjacent to the planned repository at the Olkiluoto site. At the end of 2015, the Finnish SNF repository is projected to open in 2024 as the first of its kind. Four related lessons learned in Finland during the past 30 years are: 1. The selection of potentially-suitable sites for geological disposal of SNF proved to be more difficult and more time consuming than initially envisioned. 2. The siting of nuclear facilities has evolved from being mainly a radionuclide containment and isolation challenge to becoming a critical moving-target, socio-political challenge governing progress with majority local acceptance being a fundamental imperative. 3. Public and political acceptance and support are strongly linked to the competency, credibility and transparency of the involved organisations and the maturity/pedigree of the proposed disposal concept. 4. Timely sharing lessons learned and conducting selective collaborations with other countries proved to be successful keys to timely and cost-effectively obtain and maintain public and political acceptance and support. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Issue Resolution During the Development of the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20125

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high-activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance (i.e., 100 m from the SDUs). In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. A number of issues were overcome during the development of these simulations. These issues were identified as part of internal technical reviews. Simulations are developed by people and people make mistakes, so the internal technical review process is a vital step in PA development. Specific examples of resolved issues include a unit-conversion error, an inappropriate definition for a model boundary condition, a model time-stepping issue, and an error in the calculation for the buildup of contaminants in soil. Actions taken to address these issues resulted in an improved product with a better supported technical basis and more defensible results. The identification and correction of these issues are discussed. By understanding these issues, model developers and technical reviewers working on PAs in the future may avoid repeating these types of mistakes. Transparency with respect to these mistakes builds trust between waste management sites, regulators, and stakeholders. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

A Study of a Monitoring Program for Intermediate Depth Disposal of Low- Level Radioactive Waste - 20167

This study investigated a monitoring program for intermediate depth disposal of low-level radioactive waste. At first, it extracted monitoring targets, such as the status of cracks in the low diffusion layer after closure of the facility. Secondly, it focused on distributed optical fiber sensors which can minimize the amount of cable, and measurement positions and methods were specifically determined to propose the monitoring method. Finally, conformity to regulatory requirements was verified to clarify key points and issues to be addressed in performing monitoring. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

SRS Strategy for Tank 3 Salt Dissolution - Performance and Lessons Learned - 20441

The Savannah River Site (SRS) Tank Farms have 51 underground waste tanks used to store and process liquid nuclear waste materials. There are 4 different tank types, ranging in capacity from 2,840,000 to 4,920,000 L (750,000 to 1,300,000 Gal). Twenty-four of the tanks are older style and do not meet full secondary containment standards. The older style tanks are the initial focus of waste removal efforts for tank closure at SRS. Eight of these twenty-four tanks have completed waste removal and are filled with grout. Prior to salt dissolution, Tank 3 was a dry salt tank that contained 5.06 m (199.3 in) of salt and sludge waste. Additional salt waste was present on cooling coils above the salt layer up to approximately 5.59 m (220 in). Three mixing eductors were installed in Tank 3 to aid in dissolving salt waste in three tank riser access ports. A transfer pump was installed, and the transfer pump suction was located 25.4 cm (10 in) from the tank bottom. Well water was added to the tank through a downcomer until the dry bulk salt was covered with liquid. During the initial fill of Tank 3, approximately 242,000 L (64,000 Gal) of rain water were added from periodic F-Tank Farm (FTF) Catch Tank additions and approximately 17,000 L (4,500 Gal) of well water were added. Following liquid additions to cover the dry bulk salt, well water was added through the three mixing eductors in batches during each stage. Additionally, during the salt dissolution campaigns, the FTF Catch Tank was utilized to add rain water through a downcomer in the center tank riser access port as needed for volume relief in the FTF Catch Tank. Following liquid additions, the water was recirculated (internal to the tank) using the transfer pump, and a sample was pulled to confirm the target specific gravity (SpG) of the dissolved salt solution was achieved. The dissolved salt solution was then transferred to the receipt tank (Tank 7), and the mixing eductors were lowered as close to the new bulk salt layer as possible, to support subsequent dissolution campaigns. While the mixing jets were able to dissolve salt successfully in Tank 3, they did not do so in a completely uniform manner. Throughout dissolution, mounds were discovered under Riser 1, Riser 2, and the Center Riser. The mixing jets dissolved the salt around the edges of the tank well but were not as effective toward the center of the tank. FTF Catch Tank additions through the center tank riser access port downcomer were required to impact the mound under the center tank riser. FTF Catch Tank additions were sporadic as they were dependent on rainfall in the area. Additionally, all mixing jets were rarely able to be lowered to the same elevation, indicating some mounding in the bulk salt layer. Indexing of the mixing jets was also utilized to impact the salt mounds. Overall, Tank 3 salt dissolution was successful as approximately 1,476,000 L (390,000 Gal) of dissolved salt solution was transferred to Tank 7 over six stages. Slightly less than 852,000 L (225,000 Gal) of water were added to Tank 3. The original material balance prediction estimated that the bulk salt level in Tank 3 would be 3.88 m (152.8 in) after six salt dissolution stages. After six stages the actual bulk salt level was 3.38 m (133 in). This paper will discuss the salt removal strategy, each salt dissolution stage, and lessons learned for future salt dissolution. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2011

Annotated Bibliography for Drying Nuclear Fuel

Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Remote Monitoring for Stress Corrosion Cracking in Dry Cask Storage Systems

Acoustic Emission monitoring is a technique that detects and records phenomena related to the elastic waves generated by the release of strain energy during crack propagation events. The sensors used for this method are highly sensitive and do not typically need to be in relatively close proximity to the indication when compared to methods such as ultrasonic or eddy current testing. This characteristic makes acoustic emission an attractive method for monitoring locations that are not readily accessible to active inspection methods, such as with in-situ examination of dry cask storage system canisters. Acoustic emission is unique among nondestructive evaluation (NDE) methods as emission is caused by the energy released from the object inspected. Most NDE methods require that energy be introduced to the part from an external source; which is not necessary with acoustic emission. Acoustic emission signals are created during crack growth and plastic deformation. EPRI has been investigating the feasibility of using acoustic emission as a complementary NDE method for monitoring the structural condition of dry fuel storage canisters. To assess this feasibility three first of a kind field trials were conducted over a one year period. The goals of the field trials were to determine the potential path forward for the method and to assess needs for further research and development. Why this matters: The study evaluated the efficacy of structural health monitoring methods to support ISFSI license renewal commitments to support aging management of welded stainless steel canisters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Estimation of Radiation Source Terms of Hanbit Unit3 Spent Fuel with Realistic Spent Fuel Data - 19043

Besides public acceptance, the construction of the dry storage facilities requires a sophisticated safety analysis for licensing, which needs accurate estimation of radiation source terms including the spent fuel inventories. In this work, we estimated the characteristics such as heat generations, radiotoxicities, and radioactivities of spent fuels discharged from HanBit Unit 3 using the recent realistic data of the spent fuels including detailed operational histories. The estimation was performed by writing an in-house program which automatically generates ORIGEN-S input files and processes their output files. Also, we performed the evaluation of source terms using these spent fuel assemblies. In particular, these results are compared with those obtained using a simple way with a single specific power and no cooling time between successive cycles. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Panel Session 32: Panel: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (2/5)

This panel discussed the activities of bringing a first-of-a-kind treatment facility online: The Integrated Waste Treatment Unit (IWTU). The design was designed to treat 900,000 gallons of sodium-bearing liquid waste. Construction of this first-of-a-kind treatment facility was completed in 2012. A new approach was adopted in 2016 after multiple challenges were encountered during facility commissioning. This session focused on the Integrated Waste Treatment Unit (IWTU), which was constructed to convert the remaining 900,000 gallons of liquid sodium-bearing waste at the Idaho National Laboratory Site, Idaho, USA, to a granular solid. The liquid waste was generated during historic spent nuclear fuel reprocessing campaigns at the Idaho Nuclear Technology and Engineering Center and was transferred to three underground storage tanks. The waste was not first-cycle waste but was rather the liquid rinse waste from flushing waste transfer lines between reprocessing runs. IWTU will employ a steam-reforming process to treat the liquid waste. The resulting granular solid will be transferred to stainless steel canisters and then to concrete vaults, which will be stored on site until an off-site national geologic repository is available to accept the waste. Presentation proposed: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (Joel Case, US DOE (USA), Leo Thompson, Sterling Engineering and Consulting Group, LLC (USA), Joseph Giebel, Fluor Idaho, LLC (USA), Richard Boardman, Idaho National Laboratory (USA), Kevin O'Neill, US DOE (USA))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Mixed-layered bismuth-oxygen-iodine materials for capture and waste disposal of radioactive iodine

Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Mixed-layered bismuth--oxygen--iodine materials for capture and waste disposal of radioactive iodine

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Preliminary Criticality and Radiation Shielding Analysis for the Storage and Transfer of MARVEL Reactor Spent Nuclear Fuel

This report documents the results of preliminary nuclear criticality and radiation shielding assessments during transfer and dry storage of irradiated Microreactor Applications Research Validation and Evaluation (MARVEL) reactor fuel at Idaho National Laboratory (INL). The assessments focus on transfer casks and storage canisters that are currently in use at INL, which may be compatible with the irradiated MARVEL reactor fuel. The criticality assessments were performed using the radiation transport code MCNP6 with 37 MARVEL reactor fuel elements in various configurations and scenarios. All transfer and storage configurations under dry conditions were below the assumed criticality safety limit of 0.93. Some storage and transfer configurations under wet conditions exceeded the criticality safety limit. This suggests that the appropriate administrative and engineering controls, in addition to reducing the number of MARVEL reactor fuel elements per transfer cask or storage canister, can be expected to ensure criticality safety under all scenarios. The radiation shielding assessments were performed by generating conservative neutron and photon source terms using the ORIGEN module in the SCALE suite of codes. These source spectra were used to estimate the dose equivalent rates using the radiation transport code MCNP6, both on contact and 1 m away from the fuel and transfer casks. The maximum estimated dose equivalent rate of 37 unshielded MARVEL reactor fuel elements on contact is approximately 42000 R/hr. The maximum estimated dose equivalent rates on contact to the ATR transfer cask, HFEF-5 transfer cask, and high load charger were approximately 233 mR/hr, 171 mR/hr, and 201 mR/hr, respectively. This suggests that with the appropriate administrative and engineering controls, all three transfer casks analyzed can be expected to provide sufficient radiation shielding to workers during transfer of irradiated MARVEL reactor fuel. These calculations are performed to support the planning and strategy for the MARVEL project and will demonstrate the technical viability of the different configurations discussed and help identify where engineered or administrative controls may be necessary. A complete criticality safety analysis and radiation shielding analysis, including validation and contingency and accident analysis must be completed by licensed and authorized personnel before any transfer or storage of MARVEL reactor nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Drop Analysis of Department of Energy Standard Canister with Fort Saint Vrain SNF

DOE manages over 300 types of SNF, most of which are located at the INL site. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF at the INL site for ?road-ready dry storage?. ?Road-ready dry storage? is a SNF management concept where SNF is packaged into dry and sealed canisters, which are then placed in on-site storage in anticipation of later transportation. The forward-looking goal of the Demonstration is establishing the foundation for a large-scale road-ready dry storage program at the INL site. The Demonstration will first package Fort Saint Vrain SNF currently stored at INL into several DOE Standard Canisters. These Standard Canisters will then be loaded into another commercial transportation or storage containment system (e.g., storage cask or transportation package). The Standard Canister is a class of standardized canisters designed for containing the large variety of DOE-managed SNF during interim storage, transportation and/or disposal at a geological repository. One critical aspect of road-ready dry storage is the ability to license the DOE Standard Canisters and its associated transportation package to 10 CFR 71. Depending on the SNF and transportation strategy, the Standard Canisters may have to maintain structural integrity under normal conditions of transport and hypothetical accident scenarios (i.e., drop events). The Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs. However, no analysis has been completed to support the recent Demonstration. This analysis will consider the Ø0.5 m × 5.1 m Standard Canister under drop scenario(s) considered in previous INL tests and analyses, including the 9 m drop at 80 degree off vertical. However, this analysis will consider the more recent Fort Saint Vrain loading configurations proposed for the Demonstration. This analysis will performed using strain-based acceptance criteria established by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3. It will be compared to previous analyses and form the foundation of further formal calculations that will be used to support licensing efforts of the road-ready dry storage system at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Management of the Cs/Sr Capsule Project at the Hanford Site. Technology Readiness Assessment Report

The Federal Project Director (FPD) for the U.S. Department of Energy (DOE), Richland Operations Office (RL) Waste Management and D&D Division (WMD) requested a Technology Readiness Assessment (TRA) for the Management of the Cesium/Strontium Capsule Storage Project (MCSCP) at the Waste Encapsulation and Storage Facility (WESF) on the Hanford Site in Washington State. The MCSCP CD-1 TRA was performed by a team selected in collaboration between the Office of Environmental Management (EM) Chief Engineer (EM-3.3) and RL, WMD FPD. The TRA Team included subject matter and technical experts having experience in cask storage, process engineering, and system design who were independent of the MCSCP, and the team was led by the Director of Operations and Processes from the EM Chief Engineer's Office (EM-3.32). Movement of the Cs/Sr capsules to dry storage, based on information from the conceptual design, involves (1) capsule packaging, (2) capsule transfer, and (3) capsule storage. The project has developed a conceptual process, described in 30059-R-02, "NAC Conceptual Design Report for the Management of the Cesium and Strontium Capsules Project", which identifies the five major activities in the process to complete the transfer from storage pool to pad-mounted cask storage. The process, shown schematically in Figure 1, is comprised of the following process steps: (1) loading capsules into the UCS; (2) UCS processing; (3) UCS insertion into the TSC Basket; (4) cask transport from WESF to CSA and (5) extended storage at the CSA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗