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Abatement of radioiodine in aqueous reprocessing off-gas

The reprocessing used nuclear fuel (UNF) releases volatile fission and activation products, including 129 I, into the off-gas of a processing plant. Mitigation of the release of vapor phase radionuclides is necessary for meeting regulatory requirements in the United States and other countries. In an aqueous reprocessing plant, volatile radioiodine could be present in several forms, depending on the chemistry of the process used. Inorganic iodine will be the predominate species in any shearing or voloxidation pretreatment off-gas and dissolver off-gas (DOG). Organic iodides such as CH 3 I, C 4 H 9 I, and C 12 H 25 I have been proposed to be generated during solvent extraction; thus, these species must be captured from the vessel off-gas (VOG). The abatement of inorganic and organic iodide species to meet United States regulatory requirements has been demonstrated in laboratory experiments using Ag-based solid sorbents. The data presented in this paper includes the effect of gas composition (e.g., the presence of water vapor and NO x ), iodine speciation (I 2 , CH 3 I, C 4 H 9 I, C 12 H 25 I), and sorbent bed parameters (e.g., temperature, sorbent age) on complete iodine capture on Ag-mordenite in an aqueous reprocessing plant.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Major Considerations on the Nuclear Material Accounting of PB-HTR Commercial Plants

The pebble bed high temperature reactor (PB-HTR) is a promising reactor type with inherent safety features in the next generation nuclear systems Based on the experience of studying and operating test PB-HTR (the HTR-10) for decades, China is constructing the first commercial nuclear power plant of modular PB-HTR in the world, i.e. the HTR-PM, which is expected to commission by the end of 2017. Since the PB-HTRs are quite different from the conventional PWRs in many aspects, including the large amount of fuel pebbles, the continuous refueling, the low heavy metal density in fuels, the safeguards of PB-HTR is important to investigate. Unfortunately, the research on this topic is insufficient by far, especially for the upcoming commercial PB-HTR plants. The PB-HTR has excellent features on proliferation resistance The graphite fuel pebbles with diameter of 6 cm are used as fuel elements, embedded with more than ten thousand TRISO coated fuel particles. This kind of fuel design provides more than one order of magnitude lower heavy metal concentration in fuel than other fuels. For the typical uranium loading of 7 grams in the HTR-PM, more than 10,000 fresh fuel pebbles (about 1.2 tons) are required to provide a significant quantity of indirect-use uranium with enrichment less than 20%. Moreover, the reprocessing of TRISO particle fuels is difficult because of the rigid SiC layer in the coating. Also, the possibility of utilizing materials from PB-HTR's spent fuels to product dirty bombs is quite low just due to the same reason as reprocessing. On the other hand, the large amount of fuel pebbles and the continuous on-line fueling introduce new challenges in nuclear safeguards and material accounting of PB-HTRs. This summary discusses the major considerations on the nuclear material accounting of PB-HTRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Choosing Your Nuclear Fuel Cycle: A Life Cycle Assessment Perspective - 16425

Nuclear waste management has always been a much debated topic and a key driver for decisions within the nuclear industry. A standardised global approach has not yet been developed and at present many countries are re-considering their position. In the UK the reprocessing of Spent Nuclear Fuel is projected to stop, but no clear indication about the future approach has been given. This contribution sets out the early stages of an approach to compare the possible alternatives based on Life Cycle Assessment (LCA), which is widely used in other sectors to assess the overall environmental burdens of a product or a service across the whole life cycle - i.e. from cradle to grave. The LCA may serve to support decision-making processes within the nuclear industry and, provided that it is used in an open and transparent way, to improve public attitudes towards nuclear energy. To date few LCA studies within the nuclear industry have been carried out, the main issue being the lack of a standardised methodology to evaluate the impacts of radionuclides. The present study proposes a new framework for this purpose, and demonstrates its application in an LCA study of the current UK approach to the management of spent nuclear fuel based on specific data from plants at Sellafield. The final results of the study will show the overall environmental footprint of the reprocessing of spent nuclear fuel, in terms of both radiological and non-radiological impacts. A 'hot spot' analysis will also be performed to highlight the critical processing steps. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Controllability of plutonium concentration for FBR fuel at a solvent extraction process in the PUREX process

Typical Purex solvent extraction systems for the reprocessing of spent nuclear fuel have a feed material containing dilute, 1% in weight, plutonium, along with uranium and fission products. Current reprocessing proposals call for no separation of the pure plutonium. The work described in this paper studied, by computer simulation, the fundamental feasibility of preparing a 20% concentrated plutonium product solution from the 1% feed by adjusting only the feed rates and acid concentrations of the incoming streams and without the addition of redox reagents for the plutonium. A set of process design flowsheets has been developed to realize a concentrated plutonium solution of a 20% stream from the dilute plutonium feed without using redox reagents. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Results of Re-evaluation of FEPs Related to Implementing the ABD Glass Program

The Savannah River Site plans to reprocess defense spent nuclear fuel currently stored in their L-Basin via the Accelerated Basin Deinventory (ABD) Program. The previous plan for the L-Basin spent nuclear fuel was to dispose of it directly in the federal repository without reprocessing. Implementing the ABD Program will result in final disposal of approximately 900 fewer canisters of defense spent nuclear fuel and the production of approximately 521 more canisters of vitrified high-level waste glass with some specific differences from the planned high-level waste glass. Because the 235U in the L-Basin spent nuclear fuel is not intended to be recovered, the fissile mass loading of the vitrified high-level glass waste form to be produced must be increased above the current value of 897 g/m 3 to a maximum of 2,500 g/m 3 . Therefore, implementing the ABD Program would produce a variant of high-level waste glass—the ABD glass—that needs to be evaluated for future repository licensing, which includes both preclosure safety and postclosure performance. This report describes the approach to and summarizes the results of an evaluation of the potential effects of implementing the ABD Program at the Savannah River Site on the technical basis for future repository licensing for a generic repository that is similar to Yucca Mountain and for one that is fully generic. This evaluation includes the effects on preclosure safety analyses and postclosure performance assessment for both repository settings. The license application for the proposed Yucca Mountain repository (DOE 2008), which is serving as a framework for this evaluation, concluded that the proposed Yucca Mountain repository would meet all applicable regulatory requirements. The evaluation documented in this report found that implementing the ABD Program is not expected to change that conclusion for a generic repository similar to Yucca Mountain or for a generic repository with respect to the preclosure safety analyses. With respect to the postclosure performance of a generic repository, no concerns were identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Back-end of the fuel cycle - Indian scenario

Nuclear power has a key role in meeting the energy demands of India. This can be sustained by ensuring robust technology for the back end of the fuel cycle. Considering the modest indigenous resources of U and a huge Th reserve, India has adopted a three stage Nuclear Power Programme (NPP) based on 'closed fuel cycle' approach. This option on 'Recovery and Recycle' serves twin objectives of ensuring adequate supply of nuclear fuel and also reducing the long term radio-toxicity of the wastes. Reprocessing of the spent fuel by Purex process is currently employed. High Level Liquid Waste (HLW) generated during reprocessing is vitrified and undergoes interim storage. Back-end technologies are constantly modified to address waste volume minimization and radio-toxicity reduction. Long-term management of HLW in Indian context would involve partitioning of long lived minor actinides and recovery of valuable fission products specifically cesium. Recovery of minor actinides from HLW and its recycle is highly desirable for the sustained growth of India's NPPs. In this context, programme for developing and deploying partitioning technologies on industrial scale is pursued. The partitioned elements could be either transmuted in Fast Reactors (FRs)/Accelerated Driven Systems (ADS) as an integral part of sustainable Indian NPP. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fuel Cycle Study to Developing a Viable Long Term Nuclear Power Scenario for the UK

Within the UK a closed fuel cycle has been in operation for decades but with the introduction of a new build programme to create 16 GWe capacity using LWR technology and with the closure of the UK's reprocessing industry, an open fuel cycle will be implemented beyond 2020. However, it is expected that 16 GWe nuclear capacity will not be enough for the UK to curb greenhouse gas emissions through increased demand for low carbon electricity, and tackle energy security. As such further nuclear expansion scenarios are being considered between 40 and 75 GWe by 2050. This study evaluates a viable 40 GWe UK nuclear scenario, evaluating the extent to which key drivers can impact a future UK nuclear energy strategy, including; the ability to sustain a relatively high nuclear capacity, fresh uranium availability, accumulation of spent fuel at decommissioning nuclear reactor sites and capacity requirements of future GDF host sites. The thermal reactor open fuel cycle is the de-facto standard around the world for many reasons including the reliance on extremely mature technology, significant operational experience and most importantly a relatively low cost. However, the premise that a thermal reactor open fuel cycle is the correct one to adopt long term is dependent on the notion that large quantities of UO{sub 2} spent fuel can eventually be disposed of in a geological disposal facility and the future availability of fresh uranium. The total amount of available uranium is unknown and if extraction from seawater proves to be economic and practical then any future concerns regarding supply disappear. Spent fuel disposability on the other hand has never been performed on a commercial scale and there is some uncertainty when and whether this can be achieved. This paper will follow from previous work already presented in 2013 [REF] which examined UK fuel cycles involving the transition to a closed fast reactor fleet by the end of this century. The revised scenarios now assume fast reactors are required later in the scenario, in-line with other countries. In addition, the new study has used physics methods originally developed by SKB to assess the physical size of the repository required to dispose of all intermediate and high level waste whilst ensuring temperature limits imposed by the engineered barriers and surrounding geology are adhered to. The disposition of less Am-241 in the repository will no doubt lead to a reduction in GDF footprint. However, this may not be important to GDF design and economics. The purpose of this paper is to also critically review the importance of and put into perspective a smaller GDF footprint and lower fissile inventory. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Investigation of Benefits from U/TRU Recycle - Quantification and Comparison to U/Pu Recycle

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there was no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.

Fast reactor recycle↗

Radiolytic degradation of n -dodecane: Insights from time resolved pulsed electron radiolysis using aromatic scavengers

This study uses time-resolved picosecond pulsed electron radiolysis and transient absorbance spectroscopy to investigate early radiolytic processes in n-dodecane, a prototypical organic solvent for reprocessing used nuclear fuel (UNF). The primary radical cation, solvated electron, and excited states in neat n-dodecane are formed and consumed on timescales that are difficult to observe directly. Aromatic solutes, biphenyl and benzophenone, were utilized as scavengers to convert these short-lived species into longer-lived, optically detectable transient products to serve as indirect probes of the fraction of the initial n-dodecane radiolytic products that can be intercepted before undergoing ultrafast decay or competing reactions. The transient absorbance spectra of these scavengers and their radiolytic products were analyzed. Results indicate that while biphenyl and benzophenone effectively capture several radiolytic species, they do not fully scavenge all n-dodecane excited states, suggesting potential inefficiencies or additional side reactions. This study provides insight into the formation and accessibility of radiation-induced excited states and charge carriers in n-dodecane, which play a key role in the degradation of solvent extraction systems used in UNF reprocessing.

58 GEOSCIENCES↗

Initial Efforts Organizing WPNCS SG-8: Preservation of Expert Knowledge and Judgement Applied to Criticality Benchmarks

The Working Party on Nuclear Criticality Safety (WPNCS) under the guidance of the Organization for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) has over 20 years of experience addressing concerns related to static and transient configurations encountered within the nuclear fuel cycle: fuel fabrication, transportation, reprocessing, storage, and geological disposal. One of the cornerstone activities of the WPNCS is the International Criticality Safety Benchmark Evaluation Project (ICSBEP), which was established to identify a comprehensive set of criticality benchmark data, evaluate the data, including quantification of overall uncertainties; compile the data into a standardized format, perform sample calculations utilizing modern nuclear data sets and codes utilized in nuclear criticality safety, and formally document the work into a single source of verified benchmark data. Annually, members of the ICSBEP Technical Review Group (TRG) contribute evaluated benchmark data that undergoes comprehensive technical review prior to publication in the ICSBEP Handbook. In the years since the ICSBEP was established, there has been much work to prepare benchmark data to support validation activities in nuclear criticality safety. The 2020 edition of the ICSBEP Handbook contains acceptable benchmark specifications for 5,053 critical, subcritical, or near-critical configurations in 582 benchmark evaluations. Modern benchmark development benefits from decades of experienced international participants, a well-established handbook format, supplementary guides to deal with uncertainty quantification, and a comprehensive review process based upon independent reviews from international experts. The ICSBEP Handbook also contains 838 configurations deemed unacceptable to support criticality safety efforts. They are recorded, with the reasoning for their rejection, to preserve the experimental data, prevent reevaluation of data that are incomplete or contain known errors, and/or to potentially allow future reevaluation of the experiment pending the identification of sufficient data to resolve identified inconsistencies and errors. Users of the ICSBEP Handbook today might notice that the rigor and quality of modern criticality safety benchmarks is much greater than those prepared within the initial decade of the project. Benchmarks with 1s uncertainties in k eff greater than 1% were traditionally rejected unless they were identified as unique experiment types that encompassed materials, fuels, or designs not available from other benchmark experiments. However, benchmarks developed using modern experimental techniques and practices typically have uncertainties on the order of a few tenths of a percent. There have been ongoing efforts to improve the overall quality of previously published benchmark evaluations. Seventy-eight evaluations, containing approximately 600 configurations, have been revised just within the past decade. An additional eleven benchmarks are under revision for updated release in the 2020 edition of the ICSBEP Handbook. If some of the historic benchmarks were resubmitted in their current form to the TRG today, they would be rejected due to lack of data, missing components in the uncertainty analysis, or incomplete benchmark model development. The use of historic criticality safety benchmarks that underestimate the total uncertainty, lack properly quantified biases, or provide inadequate benchmark specifications do not sufficiently support modern criticality safety and nuclear data efforts. Although the ICSBEP Handbook is recognized by regulating bodies to support criticality safety, users are required to justify their reasons to ignore historic benchmark data and include additional safety margins within their designs. Discussions were held at the WPNCS 23rd Annual Meeting in September 2019 regarding the aforementioned issues. The resultant decision was to establish Subgroup 8 (SG-8): Preservation of Expert Knowledge and Judgement Applied to Criticality Benchmarks. The current activities of SG-8 are discussed herein.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan

The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluation plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Component effects on crystallization of RE-containing aluminoborosilicate glass

Lanthanide-aluminoborosilicate (LABS) glass is one option for immobilizing rare earth (RE) oxide fission products generated during reprocessing of pyroprocessed fuel. This glass system can accommodate a high loading of RE oxides and has excellent chemical durability. The present study describes efforts to model equilibrium crystallinity as a function of glass composition and temperature as well as liquidus temperature (TL) as a function of glass composition. The experimental method for determining TL was ASTM C1720-11. Typically, three crystalline phases were formed in each glass: Ce-borosilicate (Ce 3 BSi 2 O 10 ), mullite (Al 10 Si 2 O 19 ), and corundum (Al 2 O 3 ). Cerianite (CeO 2 ) was a common minor crystalline phase and Nd-silicate (Nd 2 Si 2 O 7 ) occurred in some of the glasses. In the composition region studied, TL decreased as SiO 2 and B 2 O 3 fractions increased and strongly increased with increasing fractions of RE oxides; Al 2 O 3 had a moderate effect on the TL but, as expected, it strongly affected the precipitation of Alcontaining crystals.

36 MATERIALS SCIENCE↗

Spectroscopic Characterization of Aqua [ fac -Tc(CO) 3 ] + Complexes at High Ionic Strength

Understanding fundamental Tc chemistry is important to both the remediation of nuclear waste and the reprocessing of nuclear fuel; however, current knowledge of the electronic structure and spectral signatures of low-valent Tc compounds significantly lags behind the remainder of the d-block elements. In particular, identification and treatment of Tc speciation in legacy nuclear waste is challenging due to the lack of reference data especially for Tc compounds in the less common oxidation states (I–VI). In an effort to establish a spectroscopic library corresponding to the relevant conditions of extremely high ionic strength typical for the legacy nuclear waste, compounds with the general formula of [ fac -Tc(CO) 3 (OH 2 ) 3– n (OH) n ] 1– n (where n = 0–3) were examined by a range of spectroscopic techniques including 99 Tc/ 13 C NMR, IR, XPS, and XAS. In the series of monomeric aqua species, stepwise hydrolysis results in the increase of the Tc metal center electron density and corresponding progressive decrease of the Tc–C bond distances, Tc electron binding energies, and carbonyl stretching frequencies in the order [ fac -Tc(CO) 3 (OH 2 ) 3 ] > [ fac -Tc(CO) 3 (OH 2 ) 2 (OH)] > [ fac -Tc(CO) 3 (OH 2 )(OH) 2 ] – . These results correlate with established trends of the 99 Tc upfield chemical shift and carbonyl 13 C downfield chemical shift. The lone exception is [ fac -Tc(CO) 3 (OH)] 4 which exhibits a comparatively low electron density at the metal center attributed to the μ 3 -bridging nature of the –OH ligands causing less σ-donation and no π-donation. As a result, this work also reports the first observations of these compounds by XPS and [ fac -Tc(CO) 3 Cl 3 ] 2– by XAS. The unique and distinguishable spectral features of the aqua [ fac -Tc(CO) 3 ] + complexes lay the foundation for their identification in the complex aqueous matrixes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

X-ray Induced Cycling of Rare-Earth Elements between Bulk and Interfacial Liquid

Reversible cycling of rare-earth elements between an aqueous electrolyte solution and its free surface is achieved by X-ray exposure. This exposure alters the competitive equilibrium between lanthanide ions bound to a chelating ligand, diethylenetriamine pentaacetic acid (DTPA), in the bulk solution and to insoluble monolayers of extractant di-hexadecyl phosphoric acid (DHDP) at its surface. Evidence for the exposure-induced temporal variations in the lanthanide surface density is provided by X-ray fluorescence near total reflection measurements. Comparison of results when X-rays are confined to the aqueous surface region to results when X-rays transmit into the bulk solution suggests the importance of aqueous radiolysis in the adsorption cycle. Amine binding sites in DTPA are identified as a likely target of radiolysis products. The molecules DTPA and DHDP are like those used in the separation of lanthanides from ores and in the reprocessing of nuclear fuel. Furthermore, these results suggest that an external source of X-rays can be used to drive rare-earth element separations. More generally, use of X-rays to controllably dose a liquid interface with lanthanides could trigger a range of interfacial processes, including enhanced metal ion extraction, catalysis, and materials synthesis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Cyclopentadienyl coordination induces unexpected ionic Am-N bonding in an americium bipyridyl complex

Variations in bonding between trivalent lanthanides and actinides is critical for reprocessing spent nuclear fuel. The ability to tune bonding and the coordination environment in these trivalent systems is a key factor in identifying a solution for separating lanthanides and actinides. Coordination of 4,4'-bipyridine (4,4'-bpy) and trimethylsilylcyclopentadienide (Cp') to americium introduces unexpectedly ionic Am-N bonding character and unique spectroscopic properties. Here we report the structural characterization of (Cp' 3 Am) 2 (μ -4,4'-bpy) and its lanthanide analogue, (Cp' 3 Nd) 2 (μ - 4,4'-bpy), by single-crystal X-ray diffraction. Spectroscopic techniques in both solid and solution phase are performed in conjunction with theoretical calculations to probe the effects the unique coordination environment has on the electronic structure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development and testing of a novel micro-Raman probe and application of calibration transfer method for the quantitative analysis of microfluidic nitric acid streams

To simplify and improve the safety of reprocessing used nuclear fuel, an initial assessment was made of Raman microscopy applied to microfluidic volumes with a view toward the on-line spectroscopic measurement of highly radioactive solutions. This study compares a microscopic Raman probe (excitation focal point diameter 70 µm) to a larger, well studied probe (excitation focal point diameter 125 µm) used in prior investigations. This was done by chemometrically modeling and predicting concentrations of HNO3 solutions (0 M to 8 M) as they flowed through microfluidic cells based upon spectra from each probe. Spectra recorded for each probe using the same static HNO3 solution set contained in 2 dram glass vials were used as training sets to produce models for the respective probes. Modeling required baseline, normalization and smoothing preprocessing to compensate for a reduced path length between the static glass vial training set (4 cm) and the reduced path length flow cell (1 cm), wide ranging solution concentrations, and the associated non-linear spectral changes, and abrupt and uneven concentration changes of flowing solutions. The micro-Raman probe is able to produce spectra that may be analyzed chemometrically to accurately predict the concentration of flowing HNO3 solutions down to microliter volumes. Based upon RMSECV and RMSEP modeling statistics concentration predictions of the micro-Raman probe are comparable to those obtained for a macro-Raman on identical samples.

microscopic Raman probe, MicroRaman, nitric acid, ↗

Method Development for High Temperature In-Situ Neutron Diffraction Measurements of Glass Crystallization on Cooling from Melt

A glass-ceramic borosilicate waste form is being considered for immobilization of waste streams of alkali, alkaline-earth, lanthanide, and transition metals generated by transuranic extraction for reprocessing used nuclear fuel. Waste forms are created by partial crystallization on cooling, primarily of oxyapatite and powellite phases. In-situ neutron diffraction experiments were performed to obtain detailed information about crystallization upon cooling from 1200°C. The combination of high temperatures and reactivity of borosilicate glass with typical containers used in neutron experiments, such as vanadium and niobium, prevented their use here. Thus, methods using sealed thick-walled silica ampoules were developed for the in-situ studies. Unexpectedly, high neutron absorption, low crystal fraction, and high silica container background made quantification difficult for these high temperature measurements. As a follow-up, proof-of-concept measurements were performed on different potential high-temperature container materials, emphasizing crystalline materials so that residual glass in the waste form sample could be more easily analyzed. Room temperature measurements were conducted with a pre-crystallized sample in ‘ideal’ containers stable at low temperatures (i.e., vanadium and thin-wall silica capillaries) and compared to the same measurements in containers stable at high temperatures (i.e, platinum, single crystal sapphire, and thick-walled silica ampoules). Results indicated that Pt is probably the best choice if suitably sealed to prevent contamination from the sample after neutron activation.

36 MATERIALS SCIENCE↗