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At least 289 records · Page 16
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DOE OSTI2013

Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for the selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail routes for hazardous materials transport. The paper concludes that while the HM 164 regime is sufficient for certain applications, it does not provide an adequate basis for a national plan to ship spent nuclear fuel and high-level radioactive waste to centralized storage and disposal facilities over a period of 30 to 50 years. (authors)

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DOE OSTI2013

The Radioactivity Characteristics of the NPP Charcoal Sample Contaminated by Carbon-14 - 13531

The radioactivity of {sup 14}C-contaminated charcoal sample was analyzed by using a high temperature oxidation and liquid scintillation counting method. The radioactivity of the sample was monotonically increased according to the increase of the combustion time at each temperature where the experimental uncertainty was calculated in the 95 % confidence level. It showed that the {sup 14}C radioactivity was not completely extracted from the sample by simply increasing the combustion time unless the combustion temperature was high enough. The higher the combustion temperature was, the higher the recovery during the first 30 minutes was. The first 30 minute recoveries were 100 % at a temperature equal to or greater than 450 deg. C. The ratios of the recovery during the first 30 minutes to the total recovery during whole duration were more than 90 % at each experiment temperature. It was understood that the temperature was a critical factor for the complete removal of the {sup 14}C from the waste sample. (authors)

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DOE OSTI2013

Technology for Treatment of Liquid Radioactive Waste Generated during Uranium and Plutonium Chemical and Metallurgical Manufacturing in FSUE PO Mayak - 13616

Created technological scheme for treatment of liquid radioactive waste generated while uranium and plutonium chemical and metallurgical manufacturing consists of: - Liquid radioactive waste (LRW) purification from radionuclides and its transfer into category of manufacturing waste; - Concentration of suspensions containing alpha-nuclides and their further conversion to safe dry state (calcinate) and moving to long controlled storage. The following technologies are implemented in LRW treatment complex: - Settling and filtering technology for treatment of liquid intermediate-level waste (ILW) with volume about 1500m{sup 3}/year and alpha-activity from 10{sup 6} to 10{sup 8} Bq/dm{sup 3} - Membrane and sorption technology for processing of low-level waste (LLW) of radioactive drain waters with volume about 150 000 m{sup 3}/year and alpha-activity from 10{sup 3} to 10{sup 4} Bq/dm{sup 3}. Settling and filtering technology includes two stages of ILW immobilization accompanied with primary settling of radionuclides on transition metal hydroxides with the following flushing and drying of the pulp generated; secondary deep after settling of radionuclides on transition metal hydroxides with the following solid phase concentration by the method of tangential flow ultrafiltration. Besides, the installation capacity on permeate is not less than 3 m{sup 3}/h. Concentrates generated are sent to calcination on microwave drying (MW drying) unit. Membrane and sorption technology includes processing of averaged sewage flux by the method of tangential flow ultrafiltration with total capacity of installations on permeate not less than 18 m{sup 3}/h and sorption extraction of uranium from permeate on anionite. According to radionuclide contamination level purified solution refers to general industrial waste. Concentrates generated during suspension filtering are evaporated in rotary film evaporator (RFE) in order to remove excess water, thereafter they are dried on infrared heating facility. Solid concentrate produced is sent for long controlled storage. Complex of the procedures carried out makes it possible to solve problems on treatment of LRW generated while uranium and plutonium chemical and metallurgical manufacturing in Federal State Unitary Enterprise (FSUE) Mayak and cease its discharge into open water reservoirs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)

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DOE OSTI2017

Immobilization in Cement of Rice Husk used as Bio-sorbent of Radionuclides Present in Radioactive Liquid Waste - 17225

The use of new techniques for the treatment of radioactive liquid wastes has been studied in the last few years by many countries. In this context, the bio-sorption has been an attractive process. In Brazil the Radioactive Waste Management Laboratory of Nuclear and Energy Research Institute (IPEN-CNEN/SP), studied some bio-masses to be used as bio-sorbents to remove radionuclides of the radioactive liquid waste stored in this institute. The rice husk is a low cost and wide available byproduct of the agriculture and its bio-sorption capacity has been proved as feasible. Thus, it is necessary to evaluate if its immobilization in a cementitious matrix will result in a waste form that meets the requirements of Brazilian regulations for final disposal of low and intermediate level waste. This work aimed at evaluating the properties of the immobilization of rice husk in a cementitious matrix. Some properties of the waste form with incorporated rice husk were evaluated, as free water, workability, setting time and mechanical strength. The results indicated that cement matrices with 5 and 10% of biomass and water/cement ratio of 0.30 and 0.35 are viable for the production of waste forms that meets the Brazilian regulations. (authors)

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DOE OSTI2017

Proposal of a Categorization of the Disused Sealed Radioactive Sources with a View to Safe - 17285

In the Nuclear and Energy Research Institute (IPEN-CNEN/Brazil) many works to improve and develop characterization of radioactive waste techniques have been deployed by the Radioactive Waste Management Department (IPEN-CNEN/SP). In this line, this work aims to suggest a new categorization of sealed sources, which considers their activities, as suggested by the International Atomic Energy Agency (IAEA), but also the time required for the activity to decrease to a safe value. We considered the sealed sources categorization system suggested by the IAEA and we applied such categorization system to various sources currently stored by institutes of Brazilian Nuclear and Energy Commission (CNEN). Then, we also applied the categorization system defined herein, it follows that the categorization system presented by the IAEA has many applications and their use should not be abandoned. However, since the new proposed categorization systems allow to differentiate sources that in the IAEA system are in the same category, they can help define the long-term management of radioactive waste. (authors)

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DOE OSTI2017

Project Basis for Automation of a Quality Assurance System in Radioactive Waste Management - 17317

A low- and intermediate-level radioactive waste management facility is required to comply with Regulation 1.16 of the Brazilian National Nuclear Energy Commission 'Quality Assurance for the Safety of Nuclear Power Plants and Other Facilities'. However, the text of this regulation is very generic and does not address the more specific controls necessary for the management of quality. Therefore, the objective of this paper is to identify such detailed controls in all applicable activities of the facility and to provide an implementation plan in the form of flowcharts, for further development of an automated system. This work takes as a basis the recommendations of the International Atomic Energy Agency and the rules and procedures implemented by the U.S. Department of Energy's Office of Civilian Waste Management related to Quality Assurance. In such way, we intend to provide a more reliable implementation system of quality assurance for the management of radioactive waste in Brazil. Keywords: radioactive waste; quality assurance; automated system. (authors)

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DOE OSTI2017

SRS Defense Waste Processing Facility - the Nation's Only Operating Vitrification Facility - Completes 20 Years of Radioactive Operations - 17407

Savannah River Remediation (SRR) is the current Liquid Waste contractor at the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina, and operates the Defense Waste Processing Facility (DWPF), along with the other SRS liquid waste facilities. The DWPF, the nation's only operating vitrification facility, completed 20 years of radioactive operations in 2016. During this period, DWPF has safely treated the high-level liquid waste at this DOE facility since radioactive operations began with the first transfer of sludge waste feed to DWPF on March 12, 1996. A little over a month later, on April 29, 1996, the first radioactive canister was poured. DWPF poured its 4,000. canister of vitrified waste on December 31, 2015. As of September 30, 2016, DWPF has removed approximately 59.9 million curies from the liquid waste system at SRS and placed them in 4,106 canisters. SRS has operationally closed 8 high-level waste tanks. The DWPF melter has poured nearly 7.14 million kg of molten glass, and the facility has used only two melters in its lifetime. Melter 1 poured 1,339 canisters (1996-2002), while melter 2 to date (September 30, 2016) has poured 2,767 canisters (2003-present). Melter 2 continues to safely and efficiently operate. In order to achieve these accomplishments, the facility has chronicled a history of continued operational improvements such as the installation of bubblers in the melter to increase production, incorporation of glass compositions to achieve higher waste loadings, and improvements to melter design to extend melter life. As SRR moves forward to the next 20 years, continued improvements as well as maintaining the aging infrastructure through a proactive system health program are key priorities. This paper chronicles 20 years of challenges, improvements and accomplishments, and discusses future operations and infrastructure upgrades including implementation of a key flowsheet change to significantly reduce hydrogen generation in the Chemical Process Cell. (authors)

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DOE OSTI2015

Effect of low-level radioactive waste leachate on antioxidant depletion in HDPE geo-membranes - 15135

Low-level radioactive waste (LLW) and mixed waste (MW) disposal facilities in the US are required to have a service life in excess of 1000 yr. Understanding the rate of degradation of high-density polyethylene (HDPE) geo-membranes (GMs) used in barriers of LLW and MW facilities is necessary to assess their service life. HDPE GMs undergo degradation in three stages: antioxidant depletion (Stage I), induction time (stage II), and polymer degradation (Stage III). For LLW and MW facilities, these degradation mechanisms must be assessed with the GM in contact with leachates representative of LLW or MW. This paper focuses on antioxidant depletion in LLW leachate. HDPE GM coupons (2-mm thick) were immersed in radioactive synthetic leachate (RSL) with chemistry representative of leachate in LLW disposal facilities operated by the US Department of Energy's environmental restoration programs. Depletion was assessed at four temperatures (25, 50, 70, and 90 deg.C). Comparative tests were conducted with non-radioactive synthetic leachate (NSL), which has the same chemistry as RSL except radionuclides are excluded. Control tests were conducted with deionized (DI) water. Specimens were removed periodically and tested to determine their mechanical and chemical properties. Antioxidant depletion was measured by both standard and high-pressure oxidative induction time (OIT) tests. The rate of antioxidant depletion in RSL is statistically no different than depletion in NSL. Extrapolating the laboratory data at elevated temperatures to a typical LLW liner at 15 deg.C via Arrhenius modeling showed that the time for antioxidant depletion for a GM in a composite liner is approximately 650 yr when exposed to LLW leachate. Actual service life of the GM will be longer. (authors)

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DOE OSTI2015

Considerations for improving the biosphere assessment methodology and the associated conceptual and mathematical models for application to LLW/ILW radioactive wastes in Spain - 15194

We give consideration to future needs in assessment studies relating to the disposal of very low level radioactive waste (VLLW), low level radioactive waste (LLW) and intermediate level radioactive waste (ILW) in Spain. As these wastes are primarily disposed at El Cabril, particular attention is given to the characteristics of that site. The focus is on post-closure performance assessment studies and specifically on the biosphere aspect of those studies. In particular, it gives consideration to the assessment methodology that should be used, taking into account the recommendations made in the international BIOMASS and BIOCLIM programs, as well as more recent developments. It then goes on to consider the conceptual models of the biosphere that should be developed and comments on how these should be transformed into mathematical models. The issues arising in conceptual and mathematical model development are illustrated by reference to El Cabril, but several of these issues would also be applicable to other engineered facilities excavated from the surface in a variety of geographical and geological contexts. (authors)

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DOE OSTI2015

Repository planning for spent nuclear fuel and high-level radioactive waste: insights from U.S.. and international programs - 15638

For decades, the United States and other nations with nuclear programs have been planning for permanent disposition of spent nuclear fuel and high-level waste, to address the increasingly difficult problem of storing these materials for the long term. These plans are complicated by issues ranging from technological factors to regulatory approaches and public concerns about health and environmental protection. In its 2012 report, the Blue Ribbon Commission (BRC) on America's Nuclear Future identified a number of recommendations for creating a safe, long-term solution to the nation's problem of managing and disposing of its spent fuel and high-level radioactive waste. These include a recommendation that the U.S. Environmental Protection Agency (EPA) revise its standards for a high-level waste repository. The EPA is currently reviewing and synthesizing literature relevant to such standards, including: (1) existing national and international standards and guidelines for repositories; and (2) benefits and barriers to such standards, including limitations related to current technologies, as well as siting, licensing, and construction processes. This paper highlights findings of this literature review and synthesis. More than 90 countries, from Albania to Vietnam and Yemen, were identified as potential sources of relevant information based on existing nuclear programs or consideration of such programs to address future energy needs. Pursuant to the BRC recommendations, the U.S. Department of Energy (DOE) has compiled extensive information resources relevant to spent fuel and radioactive waste repositories in the Centralized Used Fuel Resource and Information Exchange (CURIE) database developed by Oak Ridge National Laboratory. Further valuable sources of information include annual proceedings of recent Waste Management Symposia, which provide insights into recent trends and progress across various national and international programs. Recent literature from a number of countries, including the United States and Canada, and the United Kingdom, Finland, Sweden, France, Germany, and other countries in Western Europe, as well as other regions including Australia, indicates that consent-based siting has emerged as a dominant theme. In addition to effective siting processes, key program needs identified in the literature include regulatory standards and performance context, as well as reliable technologies and engineered systems, such as for deep drilling and assured safe containment of these radioactive materials extending into the very long term. (authors)

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DOE OSTI2016

The Advanced Mixed Waste Treatment Project: Optimizing use of This National Asset to Treat Challenging Radioactive Waste From Other Department of Energy Sites - 16064

The U.S. Department of Energy's (DOE) Advanced Mixed Waste Treatment Project (AMWTP) is a world-class radioactive waste treatment facility with a highly experienced and innovative workforce located at the Idaho National Laboratory (INL). The AMWTP's state-of-the-art capabilities include a 62-ton super-compactor, remote cutting and sorting, macro-encapsulation systems, assay units, drum repackaging, waste characterization, and payload assembly and loading. Since operations began in 2003, the facility has been the largest shipper for contact-handled (CH) transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP). It has successfully processed over 56,000 cubic meters (m{sup 3}) of CH TRU waste and mixed low-level waste (MLLW) from the Idaho Site and approximately 700 m{sup 3} of CH TRU waste from 15 other DOE sites. While the AMWTP is nearing successful completion of its Idaho mission, other DOE environmental cleanup sites will require safe and proven capabilities to treat, package, characterize, and certify legacy radioactive waste inventories well into the next decade. In addition, DOE's national security, nuclear energy, and science programs need sustainable treatment capabilities to support ongoing and new missions. The AMWTP could continue to serve a vital role beyond its current Idaho mission to help address these larger DOE needs. To this end and in recognition that AMWTP's current priority is to complete the processing and removal of Idaho waste, DOE's Office of Environmental Management has initiated strategic planning to assess the feasibility of continued AMWTP operations to treat challenging radioactive waste from other generator sites consistent with existing regulatory agreements. This paper describes AMWTP's unique capabilities and identifies potential benefits, opportunities, and possible solutions to packaging and transportation challenges to fully use AMWTP as a national asset. A panel discussion of subject matter experts with audience participation would help to inform DOE's strategic planning on how best to use this national asset into the future. (authors)

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DOE OSTI2016

Role of Human Intrusion in Decision-Making for Radioactive Waste Disposal - Results of the IAEA HIDRA Project - 16287

Disposal of radioactive waste aims to protect people and the environment from the potential hazards of the materials by removing the wastes from the human environment. This approach is considered desirable to dispersal of the waste in the environment. However, the waste becomes concentrated in one location, which poses potential hazards should the waste become disturbed at a future date. For radioactive waste, it is generally expected that in preparing safety cases for such disposal facilities, the possibility of the wastes being inadvertently disturbed at some point in the future and the potential consequences of such an intrusion need to be considered. Potential inadvertent human intrusion (IHI) has a role in the decision-making and the safety case throughout the lifecycle of a disposal facility. In 2013, the International Atomic Energy Agency (IAEA) formally initiated a project entitled 'Human Intrusion in the Context of Disposal of Radioactive Waste' (HIDRA) to provide insights regarding the role of IHI for these decisions and to provide a common framework for consideration of IHI as part of development of a safety case for a disposal facility. HIDRA was implemented with a series of annual plenary meetings involving participants from more than 20 countries organised into three working groups. The working groups addressed: societal factors, stylised scenarios, and protective measures. The project resulted in a general approach with specific suggestions for consideration of IHI during the lifecycle of a disposal facility. From a general perspective, it was emphasized that IHI is a contributing factor to the safety case, but does not outweigh considerations related to safety of operations and long-term performance for the normal evolution of the facility. (authors)

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DOE OSTI2018

Performance Test of Removal of Gaseous Radioactive Methyl Iodide by TEDA-Impregnated Activated Carbon - 18049

During the operation of nuclear facility as well as decommissioning stage, various radioactive gaseous wastes are generated. It is generally recognized that radioiodine is potentially one of the most hazardous gas on the human body and environment. Iodine-131 release in the gaseous effluents at the nuclear facilities had been a critical environmental concern. Reducing iodine release is a critical issue in radioactive waste system evaluation, especially in their adherence to the 'as low as reasonably achievable' (ALARA) criteria because radioiodine among these gaseous wastes has radiological harm. Therefore, numerous studies for the removal of radioiodine from the off-gas stream of nuclear facilities have been performed with formation processes and chemical structure. This paper aims to evaluate the performance of removal efficiency of radioiodine gas by TEDA-impregnated activated carbon, which is used for radioiodine retention in nuclear facilities. The removal efficiencies of methyl iodide were evaluated by using a radioactive methyl iodide tracer at various temperatures up to 250 deg. C. (authors)

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

Savannah River Site High Level Radioactive Waste Tank Sample Thermolytic Hydrogen Generation - 20346

To support the revision of flammability calculations and controls for the Savannah River Site (SRS) High-Level Radioactive Waste (HLW) tanks, Savannah River National Laboratory (SRNL) conducted laboratory measurements with the goal of quantifying hydrogen produced in HLW by non-radiolytic chemical reactions (i.e., thermolysis). Testing was performed with HLW tank supernate samples that represented a cross section of types of waste at SRS, including tanks that store typical evaporator concentrate, dissolved salt-cake, dilute recycle stream from the Defense Waste Processing Facility (DWPF), evaporator concentrate of the recycle stream from DWPF, fresh waste from the canyon separations facility, and waste that has been through salt processing cesium removal. Non-radioactive simulant testing included the classes of organic compounds historically introduced into the HLW tanks that were recently shown to be the most active toward the thermolytic generation of hydrogen. The reaction rate equations developed for thermolytic hydrogen generation from each class of organic compound showed direct proportionality to organic compound concentration and hydroxide concentration. Simulant testing identified that highly concentrated waste with high hydroxide concentration had the highest rates of thermolytic hydrogen generation. Using the radioactive tank sample thermolysis measurements at high temperatures and the mechanistic salt dependence of the simulant models, a global model was developed for tank waste thermolytic hydrogen generation as a function of total organic carbon content of SRS HLW. Through the primary functionality of the global model (organic carbon and free hydroxide concentration), the relative reactivity of the organic carbon in the waste was quantified. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

The DECOVALEX international collaboration on modeling of coupled subsurface processes and its contribution to confidence building in radioactive waste disposal

Abstract The long-lived radiotoxicity of the high-level radioactive waste generated by nuclear power plants requires safe isolation from the biosphere for many hundreds of thousands of years. An international consensus has emerged that such isolation can best be provided by disposal in mined geologic repositories, a strategy that today is pursued by most countries dealing with radioactive waste. However, the need to predict the performance of such repositories over very long time periods generates large uncertainties that have to be accounted for in safety assessments. The findings from such safety assessments need to be conveyed to all stakeholders in a clear way, such that public confidence in geologic disposal solutions can be achieved. It is suggested here that close international collaboration on the technical aspects of geologic waste disposal has helped, and will continue to help, building trust and increasing confidence. This paper discusses a particular international collaboration initiative referred to as DECOVALEX, which brings together multiple teams and disciplines to collectively tackle complex experimental and modeling challenges related to geologic disposal. By describing how DECOVALEX works and by providing joint research examples, a case is made that such international collaboration contributes to knowledge transfer and confidence building in radioactive waste disposal science.

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

Feasibility of Using Steel Drums for the Confinement of Burned Low-Level Radioactive Waste Ashes

The selection of suitable materials to construct a facility for the final disposal of radioactive wastes (repository) is a challenge in materials science, since these facilities must be designed to provide confinement to the radionuclides for long periods of time. In the case of the low-level radioactive waste (LLRW) repository, the confinement of the species must be guaranteed for more than 300 years. In this context, the evaluation of the proposed materials must be performed under accelerated laboratory tests. One type of LLRW consists of ashes that result from the incineration of solid combustible radioactive waste. In Argentina, one of the proposed projects is to place these ashes inside steel drums and immobilized them by a cementation process. The issue to be resolved is whether the steel drums will undergo internal corrosion depending on the presence of certain deleterious species. The aim of the present work is to evaluate the corrosion susceptibility of steel drums containing simulated cemented incineration ashes. To this purpose, a special type of specimen was designed and the corrosion process was followed up for almost 4 years. It was found that, although the deleterious effect of chloride is undeniable, corrosion rates are not as high as to compromise the integrity of the steel drums for a design period of 300 years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Scenario development for safety assessment in deep geologic disposal of high-level radioactive waste and spent nuclear fuel: A review

Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk-based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man-made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or “base” evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario-specific SA models and the recombination of their results into an assessment of the risk of harm. Furthermore, while the FEP-based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom–up vs. top–down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches).

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