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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Design and Materials of Reference Electrodes for Radioactive Waste Tank Service – A Literature Review

The Hanford site stores approximately 55 million gallons of radioactive and chemically hazardous wastes from the production of weapons materials. The wastes are stored in 177 underground, carbon steel storage tanks, 149 of these are single shell tanks (SSTs) and 28 of these are double shell tanks (DSTs). The DSTs provide critical retrieval and interim storage before the waste is vitrified in the Waste Treatment and Isolation Plant (WTP). The DSTs have been in service for 38 to 56 years and current plans indicate that WTP operations will be completed in 2075. Thus, the tanks will need to remain in service far beyond the initial 40-year life expectancy. For life extension of the tanks, effective corrosion control practices must remain in force. This effort includes direct measurements of the extent of corrosion (e.g., ultrasonic measurements and corrosion coupons) and electrochemical processes (e.g., linear polarization measurements).

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DOE OSTI2016

Radioactive Waste Water Treatment for Fukushima Daiichi Nuclear Power Plant - 16081

The Great East Japan Earthquake of 11 March 2011 caused serious damage to the Fukushima Daiichi Nuclear Power Station (NPS) of the Tokyo Electric Power Company, Inc.. Toshiba and Shaw/AVANTech installed a Simplified Active Water Retrieve and Recovery System (SARRY{sup TM}) to remove radioactive cesium (Cs) from the contaminated water. Toshiba, in partnership with EnergySolutions, also developed and installed the Multi Radionuclide Removal System (MRRSTM), which can remove radioactive nuclides being treated, except tritium, to a level less than the concentration limit in water. As contaminated water treatment has progressed, the composition of the contaminated water has changed. Both salinity and radioactive Cs concentration in the contaminated water have been decreasing. Thus, a new adsorbent that is suitable for the present water chemistry conditions is required. In order to provide an appropriate contaminated-water treatment system, Toshiba has been examining the removal of not only radioactive Cs but also radioactive strontium (Sr) by SARRY{sup TM}. SARRY{sup TM} has been operated for removing Cs by using a Cs adsorbent. Instead of a Cs adsorbent, a new media that has high selectivity for both Cs and Sr should be installed. This paper describes the results obtained when using the newly developed adsorbent for removing Cs and Sr at once. As the Cs and Sr adsorbent, we have selected crystalline silico-titanate (CST). The distribution coefficient (K{sub d}) of CST was determined by the batch method. The lifetime of the adsorbent was evaluated by a column examination. The selected CST had a K{sub d} value of over 10{sup 4} ml/g for both Cs and Sr. The highest-selectivity CST had a lifetime (defined as the period for the outlet concentration to decrease to one tenth of the inlet level) of more than 125000 BV (bed volume) for Cs and more than 18000 BV for Sr, when the test was conducted at 10 BV/h. Calculation code was also developed for estimating the relation between the BV and the decontamination factor (DF) using an adsorption rate model for diffusion and adsorption. The calculation results of the Cs and Sr adsorption behavior agreed well with the experimental results. To evaluate the replacement cycle of adsorption tower in actual equipment, the radiation exposure during the replacement operation and the required Cs and Sr removal performance were used as limiting conditions for the estimation. The results of the evaluation showed that the selected CST has sufficient performance for practical usage. Based on this finding, Toshiba proposed loading the CST in SARRY{sup TM}, and SARRY{sup TM} loaded with this CST started removing Cs and Sr in December 2014 resulting in decreased contaminated water. It has been continuously working since then. (authors)

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NASA NTRS1977

An analysis of the technical status of high level radioactive waste and spent fuel management systems

The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

English, T.↗
DOE OSTI2015

Talking about radioactive waste management: feedback on France's experience and on the 2013 public debate on the Cigeo project - 15202

Following a brief presentation of the legal and historical framework in France, the article will take a retrospective look at the debate, from its preparation and the running of the debate process to the French national commission for public debate's publication of the report and the conclusions of this debate. Finally, it will describe the experience of information and dialogue from Andra's point of view and will explain Andra's follow-up to the debate, giving a first glimpse of the implementation of this follow-up. (authors)

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DOE OSTI2016

Salt Rock - the 60-Year-Old Prodigy Host Rock for Consent-Based Disposal of Long-Lived Radioactive Waste - 16010

The USA's HLW-disposal program has been on hold since 2010 pending enabling legislation for its only candidate HLW-repository since 1987 at the Yucca Mountain site in Nevada and/or other, 'consent-based', HLW-disposition solutions. Related projections in 2008 and 2012 suggested they might open 9-12 years and 35 years after being adequately enabled, respectively. In 2014, domestic HLW arisings exceeded the capacity of the candidate repository by >4,000 metric tons (MT) and they were projected to grow by 2,000-2,300 MT/year. One or more new HLW repositories are thus highly-likely in the future. In the meantime, the DOE should make every effort to promptly take title to civilian-generated HLW by other means. Historical evidence in the USA and abroad during the past 30+ years shows that all HLW-repository siting and development schedules and costs are governed by the inherent, intricate, relationships and related domino effects of the: 1. Maturity/robustness of the disposal concept; 2. Related levels of relevant domestic repository-sciences/engineering expertise, data, institutional knowledge, and applicable regulations; 3. Abundance and 'ease' of physical access to potentially-suitable host-rock sites; 4. 'Trust' in the implementing and regulating organizations; 5. Timely resolution of inevitable contentions and lawsuits; 6. Facility-host-acceptance level; 7. Timely and adequate enabling legislation, and therefore, ultimately, by 8. Prevailing Political Will. Based upon their respective status at the end of 2015, salt rock still offers the most, but not the only, promising path forward for new, mined, HLW repositories in the USA, and ∼5-km-deep boreholes offer the most-promising disposal-solution for small HLW containers/packages/pellets. (authors)

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DOE OSTI2013

Evaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste - 13187

Burn up calculations facilitate a determination of the composition and nuclear inventory of spent nuclear fuel, if operational history is known. In case this information is not available, the total nuclear inventory can be determined by means of destructive or, even on industrial scale, nondestructive measurement methods. For non-destructive measurements however only a few easy-to-measure, so-called key nuclides, are determined due to their characteristic gamma lines or neutron emission. From these measured activities the fuel burn up and cooling time are derived to facilitate the numerical inventory determination of spent fuel elements. Most regulatory bodies require an independent assessment of nuclear waste properties and their documentation. Prominent part of this assessment is a consistency check of inventory declaration. The waste packages often contain wastes from different types of spent fuels of different history and information about the secondary reactor parameters may not be available. In this case the so-called characteristic fuel burn up and cooling time are determined. These values are obtained from a correlations involving key-nuclides with a certain bandwidth, thus with upper and lower limits. The bandwidth is strongly dependent on secondary reactor parameter such as initial enrichment, temperature and density of the fuel and moderator, hence the reactor type, fuel element geometry and plant operation history. The purpose of our investigation is to look into the scaling and correlation limitations, to define and verify the range of validity and to scrutinize the dependencies and propagation of uncertainties that affect the waste inventory declarations and their independent verification. This is accomplished by numerical assessment and simulation of waste production using well accepted codes SCALE 6.0 and 6.1 to simulate the cooling time and burn up of a spent fuel element. The simulations are benchmarked against spent fuel from the real reactor Obrigheim in Germany for which sufficiently precise experimental reference data are available. (authors)

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DOE OSTI2019

A Generic Study of High-Level Radioactive Waste Isolation in Crystalline Rock - 19633

This paper is based on a numerical study aimed at predicting long-term (1 Ma period) placement room performance of a generic site for a Deep Geologic Repository (DGR) of used nuclear fuel in a crystalline rock setting. As part of the study, coupled Thermo-Hydro-Mechanical (THM) processes were analyzed in a sequence of bounding numerical simulations of rare, strong earthquakes, advancing and retreating glacial ice loading, extreme repository-induced effects including used-fuel canister heat generation and repository gas pressure generation, and considering long-term rock strength degradation. The details of the THM modeling approach, and the selection and evaluation of the poro-mechanical parameters used in the analyses are described. The results of these analyses are presented and compared with results of analyses where evolution of pore pressure due to temperature change and pore pressure dissipation were not considered. The sensitivity of the response to a variation of poro-elastic parameters used in the analyses is discussed. (authors)

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DOE OSTI2017

Proposal of a Descriptive Modeling of Portland Cement Behavior under Radioactive Waste Repository Environment - 17267

Storage and disposal of DSRS is a problem in many countries over the world. Brazilian inventory is about 300,000 sources (in use or stored), many of them with high activity and long lived radionuclides. Our proposal to deposition of DSRS: Dispose of the DSRS in a deep borehole repository. The IPEN concept was adapted from the BOSS Concept, developed by the NECSA/IAEA. Cementitious materials are intended to be used as a backfill material between the steel casing and the geological formation around the borehole and Casing. The required service life of the installation is counted in thousands of years. Issues to be addressed: Assessment of cement paste long-term durability and its behavior under radiation, high pressure and temperature and penetration of groundwater containing chemically aggressive species. The present study aims at defining the relevant external factors and their influence on the durability of hydrated cement. This study will be helpful on modeling the long-term reactions in the cement matrix and can improve the safety assessment. The synergism between all factors that can affect the cement is hard to predict and have to take into account many processes. For example, the calcium ion content in a cementitious matrix can be altered by many factors and depends on the cement mix designs and environmental factors that will prevail at the repository.

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DOE OSTI2013

A Framework Approach to Evaluate Cross-Cultural Adaptation of Public Engagement Strategies for Radioactive Waste Management - 13430

The complex interplay of politics, economics and culture undermines attempts to define universal best practices for public engagement in the management of nuclear materials. In the international context, communicators must rely on careful adaptation and creative execution to make standard communication techniques succeed in their local communities. Nuclear professionals need an approach to assess and adapt culturally specific public engagement strategies to meet the demands of their particular political, economic and social structures. Using participant interviews and public sources, the Potomac Communications Group reviewed country-specific examples of nuclear-related communication efforts to provide insight into a proposed approach. The review considered a spectrum of cultural dimensions related to diversity, authority, conformity, proximity and time. Comparisons help to identify cross-cultural influences of various public engagement tactics and to inform a framework for communicators. While not prescriptive in its application, the framework offers a way for communicators to assess the salience of outreach tactics in specific situations. The approach can guide communicators to evaluate and tailor engagement strategies to achieve localized public outreach goals. (authors)

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DOE OSTI2022

Radioactive waste immobilization of Hanford sludge in magnesium potassium phosphate ceramic forms

Here, this paper evaluates immobilization of Hanford K-Basin tank sludge in magnesium potassium phosphate ceramic forms. The waste forms were produced using two simulated non-radioactive sludge streams, each with distinct characteristics and composition. Ceramicrete with wollastonite as filler was used as the matrix for this purpose. The resulting waste forms were tested for their mechanical properties, radiation stability, and leaching resistance. In another series of tests, the samples were vitrified in borosilicate glass and glass waste forms were produced. The Product Consistency Test, the American Nuclear Society's ANS 16.1 test, and the Toxicity Characteristic Leaching Procedure, which are used to develop waste acceptance criteria in the United States for permanent storage of treated waste, were used for evaluation of the leaching resistance of all waste forms.

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DOE OSTI2019

The Durability of Containers for the Disposal of Radioactive Waste - Swiss and Belgian R and D Studies - 19392

In this paper, we present a summary of the results from research and development (R and D) studies associated with the durability of steel-based disposal canisters for high-level waste (HLW) and spent nuclear fuel (SNF), focusing on experimental studies carried out in support of the Swiss and Belgian disposal programmes. We present the results of laboratory studies carried out for periods of the order of ten to twenty years, and compare them with more recent observations from in situ experiments carried out at the Mont Terri and Grimsel underground laboratories. For the Belgian programme, which is currently envisaging relatively high gamma dose rates outside the steel overpack, we also show the results of studies focusing on gamma radiation at dose rates compatible with those expected in the disposal system. These studies show how, in bentonite systems, corrosion rates decrease to values lower than 1 μm year{sup -1} after an initial period. In the case of cement-based systems, corrosion rates lower than 0.1 μm year{sup -1} are typically observed, with modest effects of gamma irradiation. (authors)

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DOE OSTI2020

Characterization Strategy of a Limited Number of Radioactive Waste Effluent Tanks in View of Decommissioning a Large Number of Tanks - 20286

In order to organize the D and D of the former effluent treatment facility at the CEA Cadarache Center (France), the key data to gather through the facility characterization are the following:: volumes and masses, liquid/solid distribution and specific activity. By studying the treatment processes and the history of the facility, families of tanks have been defined and a reference tank chosen for investigation in each family. The intrusive investigations (video, residue thickness, dose rate..), the gamma spectrometry measurements carried out on site on the residue samples and radiochemical results collected in laboratories, led us to conclude that there was a relative radiological homogeneity (for specific activity and Am-241/Cs-137 ratio) within each tank. The next step consisted in extrapolating the specific activity evaluated for the reference tanks, to the other tanks of each family by using data such as geometry, dose rate or the function of the tanks within the process. Thus, a cartography of the 130 tanks of the facility is available. Finally, the conclusions drawn from the radiological homogeneity of the tanks and their affiliation to homogeneous, well-identified tank clusters lead to a simple and efficient operational approach for the D and D of the 130 tanks of the facility. Indeed, because of the residue characteristics (homogeneous activity and strong differentiation of Cs/Am ratio), a sample collected when opening the tank, and measured on site by gamma spectrometry will allow us to validate the operational design and the emptying techniques defined during the D and D study phase thanks to the characterization of a limited number of reference tanks. (authors)

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DOE OSTI2019

Development of Robotic Crawlers for Inspection of Radioactive Waste Tanks and Transfer Lines - 19573

The leaks in the primary liner of Double Shell Tank (DST) 241-AY-102 have motivated a number of research efforts to develop inspection tools that can obtain information regarding the integrity of the Hanford DSTs. Recently, there have been additional concerns regarding the DST secondary liners as well various pipelines within the waste transfer system. Over the past few years, Florida International University (FIU) has developed a pneumatic pipe crawler that can navigate through 3- and 4-inch diameter pipes and provide visual feedback while traveling through pipelines in excess of 100 feet and multiple 90 degree elbows. Recent updates to the crawler include a new sensor module that can provide information regarding environmental conditions within the pipeline and a mapping of the interior surface of the line. The sensor module is a carousel type device that can rotate 360 degrees and is placed near the front of the 3- and 4- inch crawler. The module can house a number of sensors including temperature, pressure, radiation, Lidar, optical camera and an inertial measurement unit (IMU). The intent of the module is to improve the functionality of the crawler by providing environmental conditions within the pipe and internal surface topography maps that can assist in understanding potential wear, pitting or thinning of the pipe. In addition to the 3- and 4-inch crawler, FIU is developing a 6-inch crawler that utilizes the same principles as the smaller unit. The objective of this tool is to navigate through the 6-inch drain lines leading to the drain channels within the foundation underneath the secondary liner of the DSTs. This crawler will house a miniature rover that will be deployed at the end of the drain pipe, where it connects to the foundation. The miniature rover will operate upside down, via magnets, underneath the secondary liner within the drain slots and provide visual information regarding the condition of the secondary liner. Thinning of the secondary liner has been noted from the annulus of the some of the DSTs, but it is unclear if the thinning is isolated to locations near the annulus. (authors)

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DOE OSTI2018

Steam Reforming Process for Treating Radioactive Waste - 18172

In 2005 the THOR steam reforming fluidized-bed reactor (FBR) technology was selected by the DOE for treatment of sodium-bearing waste (SBW) at the INL. During commissioning a number of equipment reliability and process control challenges were encountered preventing the start of radiological operations. To address these challenges the team adopted a phased systematic mechanistic approach, underpinned by scientific principles and supported by some of the world's leading experts in FBR technology. The Phase 1 effort broadly addressed the entire plant and associated modifications. Phase 2 focused on the most critical issues associated with the denitration and mineralization reformer (DMR) vessel: specifically, mitigation of wall scale (mineral deposits) that formed on the inside of the DMR; mitigation of friable agglomerates (sand-castles) that caused operational instabilities; repair of damaged lower fluidizing gas distributor (ring header), and resolution of auger-grinder (A/G) reliability issues. Basic chemistry and fluidization, as well as operational issues were evaluated using engineering analysis, modeling, bench-scale, pilot scale, and full-scale runs to validate the resolutions. The bench scale testing revealed the need to ensure at least a 5 vol% concentration of CO{sub 2} in the fluidizing gas to minimize undesirable reaction intermediates that could form in less fluidized regions of the bed. Engineering analysis and modeling results pointed to the need to redesign the fluidizing gas distributor to overcome fluidization problems. Chemistry evaluations guided efforts to minimize the deposition of wall-scale. The process modifications were confirmed on the pilot level and are being validated in the full-scale unit (Phase 3). (authors)

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DOE OSTI2020

Deep Isolation-Development of the Safety Case for Disposal of Radioactive Wastes in Horizontal Boreholes - 20028

Deep Isolation has developed a safe, secure, and permanent geological disposal method for high-level waste, including spent nuclear fuel as well as sealed sources and other highly radioactive materials. The method leverages well established directional drilling technology to create horizontal repositories deep underground. The combination of great depth (1-3 km) and the ability to precisely position repositories in a horizontal orientation provides access to geologic strata that are inaccessible for typical mined and deep vertical borehole repositories. A number of potential direct and indirect safety benefits accrue including:1) greater depth below regional freshwater aquifers; 2) increased flexibility and more geologic options for siting; 3) access to deep formations with persistent and sustained reducing conditions. In addition there are a number of safety elements related specifically to the horizontal repository geometry including: 1) passive direction of thermally driven fluid and radionuclide movement away from the vertical access hole and toward the 'dead end' portion of the repository; 2) mitigation of seismic hazards by orienting repositories parallel to local and regional fault structures. In this paper we explore and discuss some of the key technologic, geologic and hydrologic elements that support the deep horizontal borehole safety case. The stalemate seen across the globe on the disposal of spent nuclear fuel and high-level waste can be broken. Deep Isolation offers a novel option for safe, secure, and permanent deep geological disposal of nuclear waste that can be developed as centralized repositories or adapted to smaller regional or site-specific repositories located near waste sources. (authors)

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