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At least 1,693 records · Page 94
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

On Temperature Limits for the Outer Surface of a Container in a Disposal Facility - 20313

The temperature in a disposal facility for high-level radioactive waste (HLRW) is an issue for safety. When the recommencement of the search for and selection of a site for a disposal facility for HLRW in Germany was stipulated by the Site Selection Act (StandAG 2017) in 2017, a precautionary temperature limit of 100 deg. C on the outer surface of the containers with high-level radioactive waste in the disposal facility section was set. This precautionary temperature limit shall be applied in preliminary safety analyses provided that the 'maximum physically possible temperatures' in the respective host rocks have not yet been determined due to pending research. Therefore, this issue is addressed and discussed in this paper, contributing to 'pending research' by a review of the literature. This paper briefly discusses a few examples of thermohydraulic, mechanical, chemical and biological processes in a disposal facility, because temperature limits are derived based on safety impacts regarding THMCB-processes. The temperature-dependent processes have been extracted from databases for features, events and processes (FEP-databases). Furthermore, it is discussed if the feasibility to retrieve and recover HLRW is hampered at high temperatures. A design temperature concerning single components of a disposal facility for the preservation of their features can be derived when a safety concept is established. However, the interactions of all relevant processes in a disposal concept must be considered to determine a specific temperature limit for the outer surface of the containers. Therefore, applicable temperature limits may vary for particular safety and disposal concepts in the following host rocks: rock salt, clay stone and crystalline rock. Technical solutions for retrieval and design options for recovery seem to be viable up to temperatures of 200 deg. C with different, sometimes severe, downsides according to expert judgement. Temperature limits regarding the outer surface of the containers can be derived specifically for each safety concept and design of the disposal facility in a host rock. General temperature limits without reference to specific safety concepts or the particular design of the disposal facility may narrow down the possibilities for optimisation of the disposal facility and could adversely affect the site selection process in finding the best suitable site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Challenges when performing economic optimization of waste treatment: A review

Highlights: • Review of main optimization tools in the field of waste management. • Different optimization methods are applied. • Different fractions are analyzed. • There is focus on different parameters in different geographical regions. • More research is needed which encompasses both recycling and energy solutions. - Abstract: Strategic and operational decisions in waste management, in particular with respect to investments in new treatment facilities, are needed due to a number of factors, including continuously increasing amounts of waste, political demands for efficient utilization of waste resources, and the decommissioning of existing waste treatment facilities. Optimization models can assist in ensuring that these investment strategies are economically feasible. Various economic optimization models for waste treatment have been developed which focus on different parameters. Models focusing on transport are one example, but models focusing on energy production have also been developed, as well as models which take into account a plant’s economies of scale, environmental impact, material recovery and social costs. Finally, models combining different criteria for the selection of waste treatment methods in multi-criteria analysis have been developed. A thorough updated review of the existing models is presented, and the main challenges and crucial parameters that need to be taken into account when assessing the economic performance of waste treatment alternatives are identified. The review article will assist both policy-makers and model-developers involved in assessing the economic performance of waste treatment alternatives.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Glass Formulation Testing to Increase Sulfate Incorporation

About 50 million gallons of high-level mixed waste is currently in storage in underground tanks at The United States Department of Energy's (DOE's) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE's Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

The Off-Site Source Recovery Program Capabilities in Recovering Transuranic Sealed Sources - 17599

Over the past 16 years, the mission of the Off-Site Source Recovery Program (OSRP) has been recovering disused radioactive sealed sources in the interest of national security, public health and safety. This program, which is sponsored by the National Nuclear Security Administration's Office of Global Material Security's Office of Radiological Security, is managed at Los Alamos National Laboratory and Idaho National Laboratory and has recovered sources worldwide. To date, OSRP has recovered more than 38,000 sources from over 1,300 locations worldwide. Since its inception, OSRP has grown from recovering only sealed sources containing one transuranic (TRU) isotope, Pu-239, in the early days of the program, to recovering sealed sources containing several other TRU and non-TRU isotopes. Primarily, OSRP's mission includes Pu-238, Pu-239, Am-241, Cf-252, Cm-244, Sr-90, Cs-137, Co-60, Ir-192 and Ra-226 sources that are US-origin, defense related and not commercially disposable. Over the lifetime of this program the types of recoveries performed have expanded from a single site to multi-site round-up recoveries. An additional approach, referred to as self-ship recoveries, is also used in cases where only a few sources may be present at a licensee's site. In these situations, the licensee ships their source(s) to an OSRP staging facility for ultimate consolidation and disposal. OSRP recovers sources from many types of licensees, with the majority of sources being removed from industry, universities and hospitals. These include sealed sources used in industrial radiological devices like moisture-density gauges and thickness gauges, medical devices such as blood irradiators and radiotherapy devices and sealed sources from nuclear power plants, among others. The recovery and management of sealed sources is a complex process covering multidisciplinary areas, including source construction, characteristics of the radioactive material, source packaging and transportation per DOT regulations, safe and secure storage, and final disposition per waste acceptance criteria of repository sites. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Dounreay (UK) Shaft and Silo Decommissioning - Remote Handling in an Extreme and Uncertain Environment - 18420

DSRL is a wholly-owned subsidiary of the Cavendish Dounreay Partnership Ltd, a consortium of Cavendish Nuclear, CH2M and AECOM. It is funded by the Nuclear Decommissioning Authority (NDA) to deliver the site closure programme. Today, Dounreay is a site of construction, demolition and waste management, all of it designed to return the site to an interim end state. Dounreay is recognised as the most complex nuclear site closure project in Europe. The experimental nature of many of the sites redundant facilities means the clean-up and demolition requires innovation as well as great care. It is driving innovation in the approach to remote handling and robotics in the highly complex, hazardous and uncertain environment, all within tight financial constraints by UK Government. Veolia Nuclear Solutions (VNS) is instrumental in delivery of the overall solutions for the shaft and silo decommissioning project working with other members of industry and directly for DSRL to deliver the overall concept design for waste treatment as well as specialised equipment to recover highly hazardous and uncertain materials in this unique project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Modeling and Analysis of the Transport and Disposal of Beryllium Moderator Blocks and Greater than Class C (GTCC) Waste

Radioactive waste in the United states is categorized based on its source, radioactivity, and security risk. The categorization method employed by the Nuclear Regulatory Committee (NRC) for low-level radioactive waste is not currently applied to any waste generated by the Department of Energy (DOE) or disposed of in DOE facilities. As a result, there is a large volume of DOE-generated waste with characteristics similar to the NRC’s “Greater than Class C” (GTCC) waste category which presently do not have a path for long-term disposal. Much of this waste cannot undergo the standard commercial disposition process due to it being categorized as Transuranic (TRU) waste under DOE guidelines, due to elevated concentrations of key fission products. Possible solutions to this dilemma are explored in the Environmental Impact Statement for the “Disposal of Greater-Than-Class-C Low Level Radioactive Wave” (EIS-0375). This paper analyzes several EIS possible paths for disposal for this “orphaned” waste. One example that highlights this categorization issue is the spent beryllium cladding that has been extracted from the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) throughout its operational lifetime. These beryllium blocks surround the reactor’s main chambers and components and serve as neutron moderators. This paper will use this particular waste form to analyze the economic and technological viability of transporting and storing this material under the storage and transportation criteria of the Waste isolation Pilot Plant (WIPP), which was deemed the most feasible disposition path according to the EIS. The example of this waste form used in this paper that highlights this categorization issue is the spent beryllium cladding extracted from Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR). These beryllium blocks surround the reactor’s components and serve as neutron moderators, and thus have accumulated high concentration of high-activity transuranic isotopes. This paper analyzed the viability of transporting and storing this material under the storage and transportation criteria of the five primary disposition paths covered in EIS-0375. An array of calculations was carried out to assess the viability of the various disposition paths for the beryllium shipments as well as other waste shipments that fall within the GTCC category. Modeling with MCNP 6.2 was conducted to determine whether the beryllium blocks could be safely stored and transported within a 72-B cask; the standard shipping container for remote-handled, transuranic waste, while also meeting regulatory limits at various disposition paths. A cost analysis of the transportation and long-term disposal paths mentioned in the EIS was also carried out using available data and information from similar waste shipments. Lastly, a geochemical analysis of the various geological repository discussed in the EIS was also carried out using the Geochemists Workbench Release 14. Long-term disposal of the beryllium blocks at the Waste Isolation Pilot Plant (WIPP) proved to be the most cost-effective long-term disposition option of the ones considered in the EIS. A dose rate calculation at both contact and remote-handling distances indicate that the analyzed beryllium shipments should not exceed the exposure limits at any of the considered locations. Greater-than-class-C waste has been left in a regulatory state of limbo for years, resulting in backlogs of inventory across several research sites in the United States. Due to its high activity and presence of transuranic isotopes, it is imperative to ensure that it remains inaccessible and sequestered both in its short-term interim as well as a long-term geologic time scale. The information and assessments done in the paper could potentially serve as a reference for any future shipments of this waste form at the WIPP facility as well as other disposition paths that may be considered in the future.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Actinide Removal Process October 2014 Sample Analysis

Filtration within the Actinide Removal Process (ARP) currently limits the throughput in interim salt processing at the Savannah River Site. In this process, batches of salt solution with Monosodium Titanate (MST) sorbent are concentrated by crossflow filtration. The filtrate is subsequently processed to remove cesium in the Modular Caustic Side Solvent Extraction Unit (MCU) followed by disposal in saltstone grout. The concentrated MST slurry is washed and sent to the Defense Waste Processing Facility (DWPF) for vitrification. During past ARP processing, there has been a degradation of filter performance manifested as the inability to maintain high filtrate flux throughout a multi-batch cycles during Salt Batches 6, 5 and 4. The objectives of the current effort for Salt Batch 7 samples, which are similar to the most recent ARP sample analysis from 2013, were to characterize the feed streams, to determine if solids (in addition to MST) are precipitating and causing the degraded performance of the filters, and to assess the particle size and rheological data to address potential filtration impacts. The samples from the LWPT and LWHT were obtained from several stages of processing of Salt Batch 7B, Cycle 1, Batch 67.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Normal Pulse Voltammetry for Real Time Analysis of Electro-refiner Salt with High Concentrations of Analytes

Spent nuclear fuel pyro-processing could be a preferred alternative for aqueous processing. However, there will be a large hold up of actinides in the molten salt media that must be safeguarded in commercial facilities. A real or near real time monitoring method for electro-refiner salt needs to be developed to meet this requirement. In addition, real time monitoring method can also be applied to optimize process control. Voltammetry would be an ideal method to monitor the ER salt composition because of the low cost and robustness. The equipment required to carry out voltammetry analysis include only a potentiostat and three electrodes, which can be placed in-situ.

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

Control of Radioactive Methyl Iodide Gases by Mesoporous Hybrid Adsorbents at High Humid Condition - 19329

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. There has been some concern regarding the use of TEDA-impregnated activated carbon as an adsorbent for radioiodine removal, because of the low ignition point of TEDA and the possible release of TEDA from the charcoal in continuously flowing air streams. Therefore, the TEDA-impregnated activated carbon cannot use to trap radioiodine generated from high humid condition, due to the rapid drop of the removal efficiency. It is obvious that high humid can be expected during design basis event such as loss-of-coolant accident (LOCA). Therefore, investigation on the retention of organic radioiodine (CH3I131) by various activated carbons at high humid condition has extensively been performed so far. But there are no good results on the adsorption of methyl iodide at high humid condition up to now. As a result, the promising approach to the problem of iodine removal under high humid conditions is to use organic-inorganic hybrid adsorbents in which are enough retention characteristics from hydrophobic property. In order to develop the best removal efficiency for the control of radioiodine gas generated from a high humid process, a hydrophobic adsorbents was developed and the performance test of adsorbent using radioactive methyl iodide tracer was also performed. Demonstration test of adsorbent using radioactive methyl iodide tracer was performed at high humid conditions up to 97% relative humidity. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Radioactive Demonstrations Of Fluidized Bed Steam Reforming (FBSR) With Hanford Low Activity Wastes

Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) are being evaluated. One immobilization technology being considered is Fluidized Bed Steam Reforming (FBSR) which offers a low temperature (700-750?C) continuous method by which wastes high in organics, nitrates, sulfates/sulfides, or other aqueous components may be processed into a crystalline ceramic (mineral) waste form. The granular waste form produced by co-processing the waste with kaolin clay has been shown to be as durable as LAW glass. The FBSR granular product will be monolithed into a final waste form. The granular component is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals such as sodalite. Production of the FBSR mineral product has been demonstrated both at the industrial, engineering, pilot, and laboratory scales on simulants. Radioactive testing at SRNL commenced in late 2010 to demonstrate the technology on radioactive LAW streams which is the focus of this study.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Near-Complete Retention of Technetium During Vitrification of Hanford Low Activity Waste With Recycle - 17315

Technetium is one of the more volatile radionuclides for which retention in low activity waste (LAW) glass under the high temperature conditions during vitrification can vary depending on feed composition, feed chemistry, and melter operating parameters. High retention of technetium in the glass is desirable in order to minimize the fraction of technetium that is directed to secondary waste streams, which include liquid off-gas treatment system effluents. Consequently, test data are crucial to assess and underpin the assumptions underlying such mass balance projections, the compositions of secondary waste streams, and, in particular, the effects of recycle on increasing the incorporation of technetium in the glass products and minimizing the fraction in non-glass waste forms. Previous testing performed on a scaled melter system that incorporates prototypical off-gas treatment components and real-time recycle of the liquid effluents back to the melter feed demonstrated that recycle resulted in significant increases of the incorporation of technetium into the LAW glass product and behavior that was in good agreement with process models for the system. However, the maximum technetium retention achieved in those tests was less than predicted. Tests conducted in the present work demonstrated, for the first time, near-complete retention of technetium in the LAW glass product with recycle. Holdup of technetium in the vacuum evaporator, which concentrates the liquid effluents prior to their recycle to the melter feed, was identified as the primary reason for the previously observed shortfall. The fate of other volatile constituents was also assessed. In particular, in addition to technetium, rhenium, which is frequently used as a non-radioactive surrogate for technetium, was included in these tests. This allows for a direct comparison of the behavior of rhenium and technetium during vitrification and in the down-stream processes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Numerical comparison of bubbling in a waste glass melter

Radioactive tank waste is scheduled for vitrification at the Hanford Tank Waste Treatment and Immobilization Plant, also known as the Vit Plant, being constructed at the Hanford Site. Testing of the pilot-scale, joule-heated DuraMelter 1200 at the Vitreous State Laboratory has shown that bubbling increases the melt rate, and as a result, melter throughput. Computational fluid dynamics (CFD) models of this pilot-scale waste glass melter were developed to improve our understanding of the processes that occur within the melter to aid in process optimization and troubleshooting of the Vit Plant melters. Unfortunately, model validation is complicated by the difficulty of obtaining suitable experimental data for operational melters attributed to the large-scale, radioactivity and high temperatures at which these melters are operated. This study focuses on confirming the fidelity of the CFD models to accurately reproduce the bubbling behavior. Because of the paucity of experimental data at the resolution required for CFD validation, a code-to-code comparison was used to evaluate two common approaches for simulating flows of immiscible fluids on numerical grids and resolving multiphase interfaces. Here, the volume of fluid and level set methods are used to track the dynamically evolving interfaces between the molten glass and the air bubbles. To aid in the verification of the results of these codes, a comparison of the bubble behavior, growth, and frequency of bubble generation are presented and a grid convergence study is performed for the two approaches. In addition to comparing the numerical results to available experimental data and observations from pilot-scale testing, the results are evaluated in relation to laboratory experiments with bubbles injected into viscous fluids, empirical correlations obtained from the published literature, and the basic laws of fluid dynamics. Furthermore, the results of this work can assist in validating waste glass melter models and provide a better understanding of the flow patterns within the WTP melters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Technical Insights for Saltstone PA Maintenance

The Cementitious Barriers Partnership (CBP) is a collaborative program sponsored by the US DOE Office of Waste Processing. The objective of the CBP is to develop a set of computational tools to improve understanding and prediction of the long-term structural, hydraulic, and chemical performance of cementitious barriers and waste forms used in nuclear applications. CBP tools are expected to better characterize and reduce the uncertainties of current methodologies for assessing cementitious barrier performance and increase the consistency and transparency of the assessment process, as the five-year program progresses. In September 2009, entering its second year of funded effort, the CBP sought opportunities to provide near-term tangible support to DOE Performance Assessments (PAs). The Savannah River Saltstone Disposal Facility (SDF) was selected for the initial PA support effort because (1) cementitious waste forms and barriers play a prominent role in the performance of the facility, (2) certain important long-term behaviors of cementitious materials composing the facility are uncertain, (3) review of the SDF PA by external stakeholders is ongoing, and (4) the DOE contractor responsible for the SDF PA is open to receiving technical assistance from the CBP. A review of the current (SRR Closure & Waste Disposal Authority 2009) and prior Saltstone PAs (e.g., Cook et al. 2005) suggested five potential opportunities for improving predictions. The candidate topics considered were (1) concrete degradation from external sulfate attack, (2) impact of atmospheric exposure to concrete and grout before closure, such as accelerated slag and Tc-99 oxidation, (3) mechanistic prediction of geochemical conditions, (4) concrete degradation from rebar corrosion due to carbonation, and (5) early age cracking from drying and/or thermal shrinkage. The candidate topics were down-selected considering the feasibility of addressing each issue within approximately six months, and compatibility with existing CBP expertise and already-planned activities. Based on these criteria, the five original topics were down-selected to two: external sulfate attack and mechanistic geochemical prediction. For each of the selected topics, the CBP communicated with the PA analysts and subject matter experts at Savannah River to acquire input data specific to the Saltstone facility and related laboratory experiments. Simulations and analyses were performed for both topics using STADIUM (SIMCO 2008), LeachXS/ORCHESTRA (ECN 2007, Meeussen 2003), and other software tools. These supplemental CBP analyses produced valuable technical insights that can be used to strengthen the Saltstone PA using the ongoing PA maintenance process. This report in part summarizes key information gleaned from more comprehensive documents prepared by Sarkar et al. (2010), Samson (2010), and Sarkar (2010).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Impact of dilution-induced precipitates on the filtration of Hanford liquid tank wastes

To facilitate its vitrification, a portion of the radioactive wastes currently stored at the Hanford Site will be staged and treated using tank-side operations to render their chemistry compliant with the requirements of Hanford’s low-activity waste vitrification facility. Initial sampling of staged feeds indicates that process-water dilution, used to reduce waste feed sodium content to levels acceptable for vitrification (5–6 M Na), may cause precipitation of fine, difficult to settle solids that could affect downstream tank-side filtration and ion-exchange. However, dilution-induced precipitation has not been demonstrated under controlled, rigorous laboratory conditions. This paper presents a set of qualitative and quantitative assessments of dilution-induced precipitation using a nonradioactive, Hanford Tank AP-105 simulated waste. Further, these studies found that dilution of AP-farm waste simulants induced precipitation of up to 150 ppm solids, regardless of whether dilution was done with process water (which contains, among other analytes, naturally occurring Ca and Mg) or deionized water. Naturally occurring process-water analytes appeared to accelerate the rate (and possibly extent) of precipitation. Filtration of diluted waste simulants also found that the precipitated solids challenged prototypic tank-side filter operations; however, the impact to filtration performance was readily managed through waste staging settle/decant operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Solid waste bin detection and classification using Dynamic Time Warping and MLP classifier

Highlights: • Solid waste bin level detection using Dynamic Time Warping (DTW). • Gabor wavelet filter is used to extract the solid waste image features. • Multi-Layer Perceptron classifier network is used for bin image classification. • The classification performance evaluated by ROC curve analysis. - Abstract: The increasing requirement for Solid Waste Management (SWM) has become a significant challenge for municipal authorities. A number of integrated systems and methods have introduced to overcome this challenge. Many researchers have aimed to develop an ideal SWM system, including approaches involving software-based routing, Geographic Information Systems (GIS), Radio-frequency Identification (RFID), or sensor intelligent bins. Image processing solutions for the Solid Waste (SW) collection have also been developed; however, during capturing the bin image, it is challenging to position the camera for getting a bin area centralized image. As yet, there is no ideal system which can correctly estimate the amount of SW. This paper briefly discusses an efficient image processing solution to overcome these problems. Dynamic Time Warping (DTW) was used for detecting and cropping the bin area and Gabor wavelet (GW) was introduced for feature extraction of the waste bin image. Image features were used to train the classifier. A Multi-Layer Perceptron (MLP) classifier was used to classify the waste bin level and estimate the amount of waste inside the bin. The area under the Receiver Operating Characteristic (ROC) curves was used to statistically evaluate classifier performance. The results of this developed system are comparable to previous image processing based system. The system demonstration using DTW with GW for feature extraction and an MLP classifier led to promising results with respect to the accuracy of waste level estimation (98.50%). The application can be used to optimize the routing of waste collection based on the estimated bin level.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

French Atomic Energy Commission Decommissioning Programme and Feedback Experience - 12230

Since the French Atomic and Alternatives Energy Commission (CEA) was founded in 1945 to carry out research programmes on use of nuclear, and its application France has set up and run various types of installations: research or prototypes reactors, process study or examination laboratories, pilot installations, accelerators, nuclear power plants and processing facilities. Some of these are currently being dismantled or must be dismantled soon so that the DEN, the Nuclear Energy Division, can construct new equipment and thus have available a range of R and D facilities in line with the issues of the nuclear industry of the future. Since the 1960's and 1970's in all its centres, the CEA has acquired experience and know-how through dismantling various nuclear facilities. The dismantling techniques are nowadays operational, even if sometimes certain specific developments are necessary to reduce the cost of operations. Thanks to availability of techniques and guarantees of dismantling programme financing now from two dedicated funds, close to euro 15,000 M for the next thirty years, for current or projected dismantling operations, the CEA's Nuclear Energy Division has been able to develop, when necessary, its immediate dismantling strategy. Currently, nearly thirty facilities are being dismantled by the CEA's Nuclear Energy Division operational units with industrial partners. Thus the next decade will see completion of the dismantling and radioactive clean-up of the Grenoble site and of the facilities on the Fontenay-aux-Roses site. By 2016, the dismantling of the UP1 plant at Marcoule, the largest dismantling work in France, will be well advanced, with all the process equipment dismantled. After an overview of the French regulatory framework, the paper will describe the DD and R (Decontamination Decommissioning and Remediation) strategy, programme and feedback experience inside the CEA's Nuclear Energy Division. A special feature of dismantling operations at the CEA comes from the diversity of facilities to be dismantled, which are predominantly research facilities and therefore have no series advantage. There is tremendous operating feedback, however. For more than twenty years in all its centres, the CEA has acquired experience and know-how through dismantling research reactors or critical models and laboratories or plants. The dismantling techniques are nowadays operational, even if sometimes certain specific developments are necessary to reduce the cost of operations. Thanks to availability of techniques and guarantees of dismantling programme financing from two dedicated funds, close to euro 15,000 Millions for the next thirty years, for current or projected dismantling operations, the Nuclear Energy Division has been able to develop, when necessary, its immediate dismantling strategy. Currently, nearly thirty facilities are being dismantled by the CEA's Nuclear Energy Division operational units with industrial partners. Thus the next decade will see completion of the dismantling and radioactive clean-up of the Grenoble site and of the facilities on the Fontenay-aux-Roses site. By 2020, the dismantling of the UP1 plant at Marcoule, one of the largest dismantling works in the world, will be well advanced, with all the process equipment dismantled. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Water O–H Stretching Raman Signature for Strong Acid Monitoring via Multivariate Analysis

Spectroscopic techniques have been applied extensively for quantification and analysis of solution compositions. In addition to static measurements, these techniques have been implemented in flow systems providing real-time solution information. A distinct need exists for information regarding acid concentration as it affects extraction efficiency and selectivity of many separation processes. Despite of the seeming simplicity of the problem, no practical solution has been offered yet particularly for the large-scale schemes involving toxic streams such as highly radioactive nuclear wastes. Classic potentiometric technique is not amiable for on-line measurements in nuclear fuel reprocessing due to requirements of frequent calibration/maintenance and poor long-term stability in the aggressive chemical and radiation environments. In this work, the potential of using Raman spectroscopic measurements for on-line monitoring of strong acid concentration in the solutions relevant to the dissolved used fuel was investigated. The Raman water signature was monitored and recorded for nitric and hydrochloric acid solution systems of systematically varied chemical composition, ionic strength, and temperature. The generated Raman spectroscopic database was used to develop predictive chemometric models for the quantification of the acid concentration (H+), neodymium concentration (Nd3+), nitrate concentration (NO3-), density, and ionic strength. This approach was validated using a flow solvent extraction system.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

The potential of high-frequency forced pulsed water jet for nuclear decommissioning - 15367

Forced Pulse Water Jetting (FPWJ) has been demonstrated in other works as a credible and consistent method for removing coatings or otherwise preparing surfaces across numerous industries. Its applicability to the nuclear industry as a powerful tool for decommissioning and waste treatment has only recently being recognized and underpinning study is required before any deployment can be considered. Use of FPWJ within the nuclear sector for decommissioning and waste treatment, requires a different degree of surface finish to achieve a decontamination effect or an assured performance. This necessitates a greater erosion of the base metal than in other industries. The addition of a Forced Pulse Modulation Unit (FPMU) to a high pressure water jetting system offers performance shifts akin to the differences between High Pressure (HP) and Ultra High Pressure (UHP) water jets. Trials have been undertaken in a controlled environment to establish the number of cycles to achieve a given mass loss and or depth of penetration that would otherwise provide decontamination of mostly flat or simple geometry surfaces using a robotic arm for consistency and safety. It has also been shown that metalwork can be readily penetrated without risk of explosion, for example to allow draining of a gas cylinder. Most nuclear facilities have a plethora of pipe systems spanning various diameters. A considerable portion of the UK Sellafield site has 2 inch bore pipe. The FPMU technology is being adapted to consider applications in enclosed systems, such as decontamination of pipes. Exploratory works to date show there to be many physical attributes that influence performance that include, standoff of the jet, established propagated forced pulse jet length, interference from resultant water spray and angle of attack. Comparatively small changes to these parameters can have a profound effect on performance. The process is capable of removing 10's microns to several mms of metal progressively and repeatedly. This enables a credible potential treatment of plant and equipment at Intermediate Level Waste and High Level Waste thresholds. In considering this as a treatment method, a balance between performance, cost and environmental impact of effluents must be made and will be studied further during future active trials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗