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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Update on Full-Scale Retrieval and Transfer System Testing for Pneumatic Conveyance of High-Level Calcine Waste - 20241

The Idaho National Laboratory processed spent nuclear fuel from 1953 to 1994. The process used solvent extraction to separate uranium, resulting in a liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive solid powder, referred to as calcine. The calcine is stored in stainless steel bins contained within concrete vaults called the Calcined Solids Storage Facilities. There are a total of seven CSSFs. The Calcine Retrieval Project is tasked with retrieving the calcine from the bins of CSSF 1 and pneumatically transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. Once the calcine is removed from CSSF1, that facility may be closed, reducing the footprint of waste storage at the INL. The same process for removing Calcine from CSSF1 may then be utilized to remove the calcine from the remaining CSSFs for final treatment and transport to a long-term storage facility. To overcome these inherent challenges with accessing CSSF 1 and extracting the calcine in a timely manner, a full-scale integrated mockup has been constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. All essential elements of the transport design are represented in the Mockup, including motive and makeup air supply (air compressor), a replica of the CSSF6 cyclone, pre-filter, HEPA filter, vacuum pump, and retrieval control devices. Construction on the Mockup was completed in 2018 and testing began in February 2019. It has already proven very beneficial to the CRP. Experience in running the Mockup have aided in the development of retrieval and transport methodologies and the selection of process parameter values. It has also uncovered some unforeseen technological challenges and will continue to be valuable to the testing team as these hurdles are overcome and the process is proven effective. Operations personnel have also benefitted from its use as a training tool and will continue to use the Mockup as a simulator to reproduce problems encountered during extraction and transfer of Calcine from CSSF 1. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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DOE OSTI2019

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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DOE OSTI2022

Flowsheet for the Neutralization of Accelerated Basin De-inventory (ABD) Material

Under the Accelerated Basin De-inventory (ABD) program H-Canyon will be dissolving aluminum spent nuclear fuel (ASNF) and then neutralizing that solution without performing head-end strike or uranium recovery operations. After dissolution in H-Canyon the material will be neutralized to a free hydroxide concentration of either 0.6 M or 1.2 M to meet the waste acceptance criteria for transfer to the SRS Concentration, Storage and Transfer Facility (CSTF). In order to ensure the material could be successfully neutralized and transferred, SRNL completed experiments with simulated H-Canyon dissolver solutions. Two bounding simulants were developed based on the expected compositions of ASNF to be dissolved, one containing only Gd as the neutron poison and one containing Fe as an additional poison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Hanford Double Shell Waste Tank Corrosion Studies (Final Report FY2023)

For fiscal year (FY) 2023, the Savannah River National Laboratory (SRNL) focused on two experimental tasks related to Hanford Double Shell Tank (DST) chemistry and integrity. The first task focused on understanding risk of corrosion due to formation of either continuous layers or discrete patches of solids on the tanks’ inner sidewalls and bottoms. Differences in the conductivity between various layers of the tank (e.g., solids, liquid, etc.) could result in differences in the electrochemical potential of the tank metal at various locations. The electrochemical potential difference may result in a corrosion current between the coupled surfaces. In FY23, SRNL investigated test configurations and protocols that could evaluate the presence of a galvanic couple between the tank bottom and the tank wall.

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DOE OSTI2025

Hanford Double Shell Waste Tank Corrosion Studies- Final Report FY2024

For fiscal year (FY) 2024, the Savannah River National Laboratory (SRNL) focused on two experimental tasks related to Hanford Double Shell Tank (DST) chemistry and integrity. The first task focused on understanding risk of corrosion due to formation of either continuous layers or discrete patches of solids on the tanks’ inner sidewalls and bottoms. The objective of this task was to determine the effect of solid deposits on the corrosion risk to the tank bottom, and whether a combination of scale, saltcake and loose solids lead to under deposit corrosion. Electrochemical testing in a single cell arrangement was employed to determine a threshold inhibitor level for the interstitial liquid in the solids, above which carbon steel was not susceptible to corrosion beneath the solids. The inhibitor levels tested were based on the previously determined probability of failure for carbon steel in a simulated waste (i.e., Pitting Factor).

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NASA NTRS1999

Effects of space flight and mixing on bacterial growth in low volume cultures

Previous investigations have shown that liquid suspension bacterial cultures grow to higher cell concentrations in spaceflight than on Earth. None of these studies included ground-control experiments designed to evaluate the fluid effects potentially responsible for the reported increases. Therefore, the emphasis of this research was to both confirm differences in final cell concentration between 1g and microgravity cultures, and to examine the effects of mixing as a partial explanation for this difference. Flight experiments were performed in the Fluid Processing Apparatus (FPA), aboard Space Shuttle Missions STS-63 and STS-69, with simultaneous 1g static and agitated controls. Additional static 1g, agitated, and clino-rotated controls were performed in 9-ml culture tubes. This research revealed that both E. coli and B. subtilis samples cultured in space flight grew to higher final cell densities (120-345% increase) than simultaneous static 1g controls. The final cell concentration of E. coli cells cultured under agitation was 43% higher than in static 1g cultures and was 102% higher with clino-rotation. However, for B. subtilis cultures grown while being agitated on a shaker or clino-rotated, the final cell concentrations were nearly identical to those of the simultaneous static 1g controls. Therefore, these data suggest that the unique fluid quiescence in the microgravity environment (lack of sedimentation, creating unique transfer of nutrients and waste products), was responsible for the enhanced bacterial proliferation reported in this and other studies.

STS-63 Shuttle Project↗
DOE OSTI2012

Treatment of Uranium Slugs at the CEA Marcoule site - 12026

The decladding units on the Marcoule nuclear site in southeast France were commissioned in the early 1960's to receive spent fuel cartridge for interim storage in pools and to prepare them for dissolution. After decladding, the uranium slugs were sent to the reprocessing plant for dissolution. The units were shut down in 1997. All the operational and process equipment was dismantled and transferred to the ANDRA waste repository between 1998 and 2001. During cleanup of the G2-G3 interim storage pool all the internal components (baskets and machinery) mounted on the civil engineering structures were cut up. In November 2007, a UNGG fuel slug in its aluminum cartridge was found at the bottom of the pool. Similarly, while iron slugs (used to control plutonium-producing reactors) were being sorted, five de-cladded uranium slugs were found between 2006 and 2008. As the processing facilities had already been dismantled, it was decided to send these fuel cartridges to a laboratory for dissolution. The project was to recover and condition the cartridge and slugs and transport them safely to the laboratory. This required the design of dedicated equipment due in particular to the high dose rate and pyro-phoricity hazard. The procedure also required safety clearance by the regulatory authorities for the operations within the facility and for the transport of radioactive materials. The cartridge was dissolved in May 2009, 18 months after it was discovered. Based on the experience acquired, the retrieval of the uranium slugs was completed in March 2011 after a year of preparation. (authors)

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DOE OSTI2015

Minimizing environmental risk through the use of evaporation - 15090

At the Hanford Site, work is being done to utilize a portable exhauster to reduce the remaining free liquid in the older underground storage tanks. Use of an exhauster allows the water in the free liquid to be evaporated reducing the possibility that mobile contaminants will leak into the soil and groundwater. Transferring the remaining liquid waste to newer tanks requires major infrastructure installation. Currently one set of tanks is undergoing the retrieval of solids and liquids. This retrieval requires substantial infrastructure in order to mobilize and pump the waste to newer tanks. Reducing the remaining liquid waste by pumping would require much of the same retrieval infrastructure and is cost prohibitive. The use of a portable exhauster will require minimal infrastructure and is much more efficient. Past results have demonstrated that this technology can be successfully used on various sizes of underground storage tanks. A portable exhauster is planned to be connected to the one tank believed to be leaking and will be used to evaporate the remaining free liquid, beginning in 2015. (authors)

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DOE OSTI2019

Different Approaches for Clearance of Metals - a Comparison - 19079

Large quantities of contaminated and potentially contaminated metals are generated during the lifetime of nuclear installations, especially when they are dismantled and demolished. In many countries, clearance for recycling is considered Best Practice for the management of metals which are contaminated or potentially contaminated. In some countries, such as Sweden, it has become a legal requirement to recycle metals and other materials when it is safe to do so and the cost is reasonable. Therefore, multiple management routes are available with the aim of establishing a graded approach. These include general (unconditional) clearance, conditional (specific) clearance of surface contaminated metals, conditional clearance of metal ingots after treatment in a licensed facility and in some cases, recycling inside the nuclear sector. Kemakta Konsult AB recently derived rounded clearance limits for the conditional clearance of scrap metal for re-melting from the decommissioning of Swedish nuclear sites, on behalf of the Swedish Radiation Safety Authority (SSM). The calculations used the RP89 exposure scenarios, but the parameters were adjusted to reflect normal practice in the Swedish scrap metal and recycling industries. The study found that, for most of the 39 radionuclides judged to be relevant, the rounded clearance levels were the same as or higher than those in RP 89. However, longer transport times in Sweden compared to the ones used for RP 89 resulted in lower rounded clearance levels for three gamma-emitting radionuclides, including Co-60, the most important radionuclide in terms of dose exposure. Cyclife Sweden AB has, through more than 40 years of operation of the metal treatment facility at the Studsvik industrial site, significant experience of the practical implementation of clearing metal ingots as well as metallic strap, in total about 50,000 tonnes. A previous study confirms very low radiological impact to the receivers of the ingots and to the general public under the current procedures and conditions for clearance. The production of steel ingots for recycling in the Cyclife plant involves pre-treatment (segmentation and decontamination) followed by melting. Certain nuclides, like Cs-137 and alpha nuclides, are transferred to the secondary waste (slag and filter dust) in the melting process. The non-separated nuclides are homogenized in the ingot. The radionuclides that are present in the ingots are expected to remain in the metal during re-melting at metal works. The general approach of [1] together with specific information for typical ingots has been used to further verify the radiological implications of applying conditional clearance of ingots. This paper provides a background, present the studies performed, give an overview of the findings and the overall result including a comparison between different approaches. (authors)

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DOE OSTI2021

Demonstration of Defense Waste Processing Facility (DWPF) Higher Fissile Content Glass

The Accelerated Basin De-inventory (ABD) Program has been proposed as an alternative for future spent nuclear fuel (SNF) and nuclear material processing at the Savannah River Site (SRS). This approach would change the baseline H-Canyon (HCAN), Concentrate, Storage, and Transfer Facility (CSTF), and Defense Waste Processing Facility (DWPF) operations. The ABD Program would require that all domestic and foreign research reactor SNF currently at SRS be dissolved, stored, and then transferred to CSTF without the recovery of uranium. Preliminary assessments in the ABD Program plan have shown that ~5000 extra SRS high-level waste (HLW) canisters would be produced if the fissile mass loading remains at the current 897 g/m 3 limit; however, increasing the limit to 2500 g/m 3 would result in ~520 extra canisters. Thus, the ABD Program plan requires an increase of the DWPF fissile mass loading limit to 2500 g/m 3 to minimize canister production. DWPF considers the following isotopes in the calculation of fissile mass loading: U-233, U-235, Pu-239 and Pu-241.

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DOE OSTI2020

The HRA/Solarium Project: Feedback Based on 20 Years of Experience in Treatment of Medium Level Waste - 20185

Since the 1980's the management of the historical site of Mol has been transferred to the National Agency for Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS). Belgoprocess is entrusted by ONDRAF/NIRAS with the operational waste management and site remediation. One of the major challenges has been the characterization, treatment and conditioning of approximately 200 m{sup 3} of medium level waste. These waste packages were stored in poor conditions in storage vaults ('HRA') or concrete containers ('Solarium'). They have been produced in various research programs and reactor operations at the Belgian nuclear energy research centre SCK.CEN, isotope production, decontamination and dismantling operations from the 1960's up to the 1980's. Despite the limited volume, this historical waste consists of a great variation of waste characteristics and waste configurations. To tackle these liabilities, a new processing facility was built in the 1990's to allow safe transfers, handling, characterization and treatment of these packages. Also some auxiliary facilities have been built to deal with the by-products, like emptied concrete containers. The engineering of the installation, safety procedures and (characterization) methodology that has been developed generically have been proved to be successful. Nevertheless, the specific nature of some items made it necessary to organize a step-by-step treatment in distinct campaigns in which some relevant extra (safety) measures had to be taken into account. This dynamic approach made it possible to process a great variety of waste types in order to to ensure that this legacy is no longer left for subsequent generations. After about 20 years of operations, there has been a great build-up of experience and feed-back concerning the waste management of these specific waste streams. Some cases will be highlighted to specify the approach followed for treatment of e.g.: - Na/NaK containing equipment; - spent radioactive sources; - fuel element residues; - medium level waste originating from activation experiments; - radium bearing medium level waste originating from Ac-227 production research. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Belgian regulatory framework for decontamination and decommissioning - lessons learned and new initiatives

Belgium can rely on significant experience in the field of decontamination and decommissioning of nuclear facilities. Several projects are ongoing and include research reactors (BR3, Thetis), uranium and MOX fuel fabrication plants (FBFC International, Belgonucleaire), fuel reprocessing facilities (Eurochemic) and radwaste processing facilities (Belgoprocess). Additional projects are expected in the coming years with the planned final shutdown of the oldest nuclear power reactor units. Two national authorities are involved in the decontamination and decommissioning process of nuclear facilities. The FANC (together with its subsidiary Bel V) is concerned for all matters related to nuclear safety and radiation protection, while NIRAS/ONDRAF is concerned for all matters related to radioactive waste and fuel management and financial provisions. These attributions ensure that all safety and material concerns are addressed and that both the licensees and the national authorities bear their own responsibilities. They rely on an existing regulatory framework covering both the procedural and the technical aspects of the decontamination and decommissioning activities. However, opportunities for regulatory improvement were raised after some recent events in Belgium, among which the bankruptcy of a nuclear company producing radioisotopes, involving numerous additional interested parties in a complex judiciary context. Amendments in the current regulations are considered to increase the prerogatives of the authorities regarding the management of radioactive waste by a licensee, the transfer of an operating license from an operator to another, and the general decommissioning strategy of a facility. Furthermore, a dedicated 'waste and decommissioning' working group within WENRA defined new reference levels applying to the decontamination and decommissioning of nuclear facilities. Belgium committed to include these requirements explicitly in its national legislation, even though most of them were already included in the existing decommissioning authorizations. Amendments will cover the safety provisions inherent to the decontamination and decommissioning phase, such as the decommissioning strategy, the qualification of techniques, the experience feedback valorization, the periodic safety reviews, the radioactive waste management, or the final characterization of the sites. Additionally, requirements regarding the authorization process will be detailed, such as the content of the licensee's application file or the structure of the safety report covering the decontamination and decommissioning phase. These changes will contribute to a more secure regulatory framework for all interested parties. (authors)

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DOE OSTI2013

Sludge Washing And Demonstration Of The DWPF Flowsheet In The SRNL Shielded Cells For Sludge Batch 8 Qualification

The current Waste Solidification Engineering (WSE) practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks to Tank 51. Tank 51 sludge is washed and transferred to Tank 40, the current Defense Waste Processing Facility (DWPF) feed tank. Prior to transfer of Tank 51 to Tank 40, the Savannah River National Laboratory (SRNL) typically simulates the Tank Farm and DWPF processes using a Tank 51 sample (referred to as the qualification sample). WSE requested the SRNL to perform characterization on a Sludge Batch 8 (SB8) sample and demonstrate the DWPF flowsheet in the SRNL shielded cells for SB8 as the final qualification process required prior to SB8 transfer from Tank 51 to Tank 40. A 3-L sample from Tank 51 (the SB8 qualification sample; Tank Farm sample HTF-51-12-80) was received by SRNL on September 20, 2012. The as-received sample was characterized prior to being washed. The washed material was further characterized and used as the material for the DWPF process simulation including a Sludge Receipt and Adjustment Tank (SRAT) cycle, a Slurry Mix Evaporator (SME) cycle, and glass fabrication and chemical durability measurements.

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DOE OSTI2015

Second phase of the OECD NEA international initiative on the preservation of records, knowledge and memory across generations - 15616

The OECD/NEA initiative on the Preservation of Records, Knowledge and Memory across Generations (RK and M) constitutes an activity within the area of transfer of information about radioactive waste disposal projects for the benefit of future generations. Launched in 2011, the initiative has now entered Phase-II, which will run through to 2017. Building on results of Phase-I, which was dedicated to exploring issues of strategic relevance for the preservation of RK and M, Phase-II of the initiative will focus on developing an understanding of the interactions between the identified issues, while continuing to examine hitherto little aspects of the same issues. This corresponds to a systemic approach to RK and M preservation consisting in employing different means and methods to preserve RK and M, which are in a complementary but also redundant relationship to each other, and in situating and analysing activities in a wider context than radioactive waste management. The strategic issues for RK and M preservation are the use of archives, cultural heritage, the so-called Key Information File, transfer of responsibilities after closure, regulations governing RK and M preservation, markers and international mechanisms that could be adapted to foster RK and M preservation. The issues were discussed extensively at 'Constructing Memory - An International Conference and Debate', which was organised by the OECD/NEA with the support of Andra (French Radioactive Waste Management Organisation) and took place in September 2014 in Verdun, France. (authors)

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DOE OSTI2025

Vitrification of Hanford Tank 241-AW-105 Waste and Equivalent Simulant

Hanford Site nuclear waste is to be vitrified at the Waste Treatment and Immobilization Plant (WTP), which is a part of the safe and efficient retrieval, treatment, and disposal mission of the U.S. Department of Energy - Hanford Field Office. A portion of Hanford tank 241-AW-105 (referred to herein as AW-105) waste was retrieved by Hanford Tank Waste Operations and Closure (H2C) and transferred to Pacific Northwest National Laboratory (PNNL). Compared to previously received and vitrified wastes (AP-107, AP-101, AN-107 and AP-105), the concentration of potassium in AW-105 was greater by an order of magnitude.

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DOE OSTI2017

Methods to estimate the transfer of contaminants into recycling products – A case study from Austria

Highlights: • Recycling of waste materials includes the risk of recycling hazardous substances. • Comparison of contaminants in recycling products to primary products and check of compliance with limit values. • Evaluation matrix to quickly scan the environmental and health risks of wastes for recycling. • Optimised sorting and quality control systems prevent transfer of contaminants into product cycles. - Abstract: Recycling of waste materials is desirable to reduce the consumption of limited primary resources, but also includes the risk of recycling unwanted, hazardous substances. In Austria, the legal framework demands secondary products must not present a higher risk than comparable products derived from primary resources. However, the act provides no definition on how to assess this risk potential. This paper describes the development of different quantitative and qualitative methods to estimate the transfer of contaminants in recycling processes. The quantitative methods comprise the comparison of concentrations of harmful substances in recycling products to corresponding primary products and to existing limit values. The developed evaluation matrix, which considers further aspects, allows for the assessment of the qualitative risk potential. The results show that, depending on the assessed waste fraction, particular contaminants can be critical. Their concentrations were higher than in comparable primary materials and did not comply with existing limit values. On the other hand, the results show that a long-term, well-established quality control system can assure compliance with the limit values. The results of the qualitative assessment obtained with the evaluation matrix support the results of the quantitative assessment. Therefore, the evaluation matrix can be suitable to quickly screen waste streams used for recycling to estimate their potential environmental and health risks. To prevent the transfer of contaminants into product cycles, improved data of relevant substances in secondary resources are necessary. In addition, regulations for material recycling are required to assure adequate quality control measures, including limit values.

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DOE OSTI2019

Melter Modeling over a Range of Scales to Support Vitrification at the WTP - 19664

Computational modeling is a powerful complement to laboratory experiments and testing of systems at various scales. When properly applied and validated, computational fluid dynamics (CFD) can provide information not readily obtainable by physical testing, especially in cases where the operating environment does not permit visual observation or the insertion of typical data acquisition instrumentation as in the case of waste glass vitrification. A series of CFD and heat transfer models over a range of scales are being developed to support melter operations at the Hanford Waste Treatment and Immobilization Plant. This paper provides a description of the models and the resulting information they provide that can be used to improve melter throughput, as well as aid in evaluating and resolving operating plant issues or upsets. This work is important for the WTP operation by providing a potential for model-predictive control of the melter operation. Increases in the rate of glass production and avoiding idling can significantly decrease the duration and cost of the vitrification campaign, potentially saving billions of dollars. (authors)

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