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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Used Nuclear Fuel Management in Asia

Used fuel discharged from nuclear reactors are presently stored in wet pools or in dry cask at-reactor and/or away-from- reactor sites. Only a few countries reprocess their civilian used fuel, notably France, Russia, and UK. No country has yet operated a geologic repository to dispose of civilian used fuel or high-level radioactive wastes (HLW). As a result, the management and disposal of used fuel and HLW are viewed as the Achilles' heel of nuclear power. Figure 1 shows the used fuel inventory in the United States (US) and East Asian nuclear power programs. The inventory is estimated based on that in 2010 and the projected growth of nuclear power in these programs to 2020 and 2030. Figure 1 indicates that in 2010, the US had about 63,900 tons of used fuel from light-water reactors (LWR), more than the statutory limit of what was supposed to go into the now-abandoned Yucca Mountain Repository, and accounting for more than 33% of global used LWR fuel total. Currently, about 2,000 tons of used fuel are generated in the existing US LWRs annually. Russia had 25,100 tons of used fuel in 2010. Based on a modest growth of nuclear power, its used fuel inventory is expected to increase to about 31,800 and 38,000 tons by 2020 and 2030, respectively. The nuclear energy programs in East Asia (China, Japan, ROK, and Taiwan) account for 19% of global used LWR fuel inventory in 2010. China's used fuel inventory in 2010 was relatively small at 1,900 tons, but is expected to grow rapidly to {approx}12,400 and {approx}31,000 tons in 2020 and 2030, respectively, based on its expanding nuclear power program (from the current installed capacity of 21.8 GW to {approx}60 and {approx}200 GW in 2020 and 2030, respectively). Besides PWRs, China also operates 2 Candu reactors which each produces roughly 8 times more used fuel in tons than a PWR of the same power. The used Candu fuel inventory would increase from 950 tons in 2010 to 2,850 and 4,750 tons in 2020 and 2030, respectively. Used fuel in Japan, ROK, and Taiwan are US-obligated. Japan's future trends of used fuel inventory are uncertain due to the unknown numbers and timing of the currently-shutdown nuclear power plants that would be allowed for restart, and the uncertain future of Japan's reprocessing at Rokkasho. Under two scenario cases (a no-nuke/no-reprocessing and a 80%-nuke with reprocessing), the Japanese used fuel inventory could be capped from 2020 to 2030 to {approx}21,500 tons for the first case, and to {approx}22,500 tons for the second. The ROK's used PWR fuel inventory would increase from 4,900 tons in 2010 to about 8,800 and 14,000 tons in 2020 and 2030, respectively. Its used Candu fuel inventory would also increase from 5,900 tons in 2010 to about 9,700 and 13,500 tons in 2020 and 2030, respectively. The Taiwan used fuel inventory should be kept at current level as nuclear power is being phase out. Its growth should be modest from 3,000 tons in 2010 to about 4,300 and 5,000 tons in 2020 and 2030, respectively even if the 4. nuclear power station would start and the operation of existing LWRs is extended. Used fuel inventory in other East Asian countries (DPRK, Vietnam, etc.) should be small, if any before 2030.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Direct Measurement of U235 and Pu239 in Spent Fuel Rods with Gamma-Ray Mirrors

The amounts of fissile Pu and U in spent nuclear fuel are of primary concern to the safeguards community. In particular, there are issues when safeguards transitions from an item accountancy basis (such as fuel bundles) to a fissile material mass basis as occurs when spent fuel enters a reprocessing plant. Discrepancies occur because item accountancy requires estimating the content of fissile material using indirect techniques such as the fuel burn-up and item-level measurements of radiation emissions from fission by-products. Direct measurement of the fissile content by monitoring line emissions from fissile species themselves is impossible because the lines are much weaker than those emitted by shorter-lived isotopes in the fuel. The goal of this project is to develop a technique to directly measure these weaker lines despite the presence of overwhelming radiation from other isotopes. This is achieved by using gamma-ray mirrors as a narrow band-pass filter. The mirrors reflect only energies of interest toward a HPGe detector that is shielded from direct view of the spent fuel and its fierce emissions. This can significantly improve the reliability with which the mass of fissile material is tracked.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Parameter survey on heating conditions for glass-bonded sodalite ceramic waste fabrication from type-A zeolite containing simulating FPs

In the pyrometallurgical reprocessing for metal fuel cycle, fission products (FPs) in the spent salt is converted into a glass-bonded sodalite by means of Pressure-less Consolidation (PC). Although a standard condition of the PC method was reported, the reason for choosing this condition is not clear particularly for the temperature condition. In the present study, a parameter survey on the heating condition was performed by fabricating the glass-bonded sodalite containing simulating FPs under various conditions. The maximum temperature, the heating duration, the glass ratio in the initial material, and the weight load for pressing the material were chosen as the variable parameters. The mass balance of each element and the quality of the product were evaluated. It was exhibited that both the volatilized content and the free salt in the product were reduced by lowering the maximum temperature. The apparent density of the product was enhanced by both increasing the heating duration and increasing the weight load. Accordingly, lowering the maximum temperature while increasing the weight load was chosen to modify the fabricating condition. As a result of the modified condition, both the volatilized and the free salt ratios were reduced without significant change in the apparent density of the product. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Compositions and methods for monitoring actinides

Compositions and methods for monitoring the quantity of actinides present in a test sample are disclosed. Compositions and methods for monitoring the motion of special nuclear materials through space are also described. Compositions and methods for monitoring the quantity of a fissile special nuclear material present in a test sample are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 30-year cool down are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 180 day cool down are also disclosed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Domestic Safeguards Material Control and Accountancy Considerations for Molten Salt Reactors

Molten salt reactors (MSRs) are a class of nuclear reactor designs with features and operational characteristics that vary significantly more than the class of light water reactors (LWRs). MSR design concepts can be broadly categorized as solid-fueled reactors with molten salt as the coolant, liquid-fueled reactors with fuel dissolved in molten salt coolant and liquid-fueled reactors with fuel dissolved in molten salt that is contained in distinct fuel tubes with a nonfissile coolant. Proposed MSR concepts include features that are specific to each design. MSR concepts vary widely across these design features, including the physical, chemical, and isotopic composition of fresh and irradiated fuel. Operational neutron energy spectrums and breeding ratios also vary significantly across design concepts. Some concepts are burner reactors designed to transmute the spent nuclear fuel from LWRs or pressurized heavy water reactors (PHWRs), while others are breeder reactors designed to breed fissile 233 U from naturally occurring fertile 232 Th. Some MSR concepts are designed to be a part of a once-through fuel cycle, while others involve chemical separation. Those that include plans to recycle the fuel differ by whether the chemical processing would be done onsite, as a process connected to the fuel salt itself, or offsite at a reprocessing facility similar to how LWR or PHWR spent fuel is reprocessed in some countries. Each overall design concept contains various aspects of each of these features to produce a unique facility

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Development of high temperature transport technology for LiCl-KCl eutectic salt in pyroprocessing

The development of high-temperature transport technologies for molten salt is a prerequisite and a key issue in the industrialization of pyro-reprocessing for advanced fuel cycle scenarios. The solution of a molten salt centrifugal pump was discarded because of the high corrosion power of a high temperature molten salt, so the suction pump solution was selected. An apparatus for salt transport experiments by suction was designed and tested using LiC-KCl eutectic salt. The experimental results of lab-scale molten salt transport by suction showed a 99.5% transport rate (ratio of transported salt to total salt) under a vacuum range of 100 mtorr - 10 torr at 500 Celsius degrees. The suction system has been integrated to the PRIDE (pyroprocessing integrated inactive demonstration) facility that is a demonstrator using non-irradiated materials (natural uranium and surrogate materials). The performance of the suction pump for the transport of molten salts has been confirmed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Early-stage oxidation and subsequent damage of the used nuclear fuel extractant TODGA; electron pulse radiolysis and theoretical insights

Radiation induced damage of extractant molecules is a well-known phenomenon responsible for reducing efficiency and increasing the waste and cost of reprocessing used nuclear fuel (UNF). As such, understanding early-stage (pico- to nanoseconds) radiation-induced reaction mechanisms is essential for informing the design of next generation extractants with enhanced radiation robustness. Here, in this work, we utilized picosecond and nanosecond electron pulse radiolysis experiments to probe the early-stage radioactive environment experienced by the organic phase extractant N,N,N',N'-tetraoctyldiglycolamide (TODGA), proposed for separating highly radioactive trivalent minor actinides (specifically americium and curium) from the trivalent lanthanides. Using comparisons to the similar ionization potential (IP) solute p-xylene, this work determined the mechanism of reaction with the ionized diluent (i.e., n-dodecane radical cation, DD˙ + ) is hole transfer to produce TODGA˙ + . At high TODGA concentrations (>100 mM), the majority of this transfer occurs faster than 10 ps via the capture of DD˙ + holes prior to their solvation with a C 37 = 300 mM. The surviving solvated holes were captured with k = (2.38 ± 0.15) × 10 10 M -1 s -1 . Attempts at subsequent hole transfer to lower IP solutes found that only 10% of holes were transferred, indicating bond rupture of TODGA˙ + occurs within 2.6 ns at 200 mM TODGA. Possible reaction pathways for the rapid decomposition of TODGA˙ + were explored using a combination of experiments and density functional theory (DFT) calculations.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principle be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Extraction of Neptunium, Plutonium, Americium, Zirconium, and Technetium by Di-(2-Ethylhexyl)- Iso -Butyramide (DEH i BA) at High Metal Loadings

Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. Further, by increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. In conclusion, the extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management.

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

Machine learning surrogates for surface complexation model of uranium sorption to oxides

Abstract The safety assessments of the geological storage of spent nuclear fuel require understanding the underground radionuclide mobility in case of a leakage from multi-barrier canisters. Uranium, the most common radionuclide in non-reprocessed spent nuclear fuels, is immobile in reduced form (U(IV) and highly mobile in an oxidized state (U(VI)). The latter form is considered one of the most dangerous environmental threats in the safety assessments of spent nuclear fuel repositories. The sorption of uranium to mineral surfaces surrounding the repository limits their mobility. We quantify uranium sorption using surface complexation models (SCMs). Unfortunately, numerical SCM solvers often encounter convergence problems due to the complex nature of convoluted equations and correlations between model parameters. This study explored two machine learning surrogates for the 2-pK Triple Layer Model of uranium retention by oxide surfaces if released as U(IV) in the oxidizing conditions: random forest regressor and deep neural networks. Our surrogate models, particularly DNN, accurately reproduce SCM model predictions at a fraction of the computational cost without any convergence issues. The safety assessment of spent fuel repositories, specifically the migration of leaked radioactive waste, will benefit from having ultrafast AI/ML surrogates for the computationally expensive sorption models that can be easily incorporated into larger-scale contaminant migration models. One such model is presented here.

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

Experimental validation of the Cristal V2.0 French criticality calculation package - 14587

A new version of the CRISTAL criticality calculation package, named CRISTAL V2.0 [1], has been developed by IRSN and CEA, in collaboration with AREVA and EDF. Four calculation routes are now available, using nuclear data issued from the JEFF-3.1.1 evaluation: two multigroup routes based on 281 groups cross-sections (APOLLO2 - MORET 5 or APOLLO2-Sn calculations), a pointwise Monte Carlo route (TRIPOLI-4{sup R}) and a criticality standard calculation route. The CRISTAL V2.0 package, which will be delivered soon to OECD/NEA data bank, benefits from a broad validation database covering almost all areas of criticality-safety applications, from fuel fabrication to reprocessing, including transportation. 3,127 critical experiments, corresponding to 342 series, either issued from the OECD/ICSBEP Handbook or performed in French facilities (confidential data) were selected for the CRISTAL V2.0 validation database. Currently, more than 2,300 experiments are already part of the validation database. This validation work highlights an overall consistency to experimental data and inter-comparisons between the different calculation routes mainly show good agreements on a large set of common experiments. (authors)

42 ENGINEERING↗
DOE OSTI2016

Prototype Demonstration of Gamma- Blind Tensioned Metastable Fluid Neutron/Multiplicity/Alpha Detector – Real Time Methods for Advanced Fuel Cycle Applications

The content of this report summarizes a multi-year effort to develop prototype detection equipment using the Tensioned Metastable Fluid Detector (TMFD) technology developed by Taleyarkhan [1]. The context of this development effort was to create new methods for evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU) isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The Tensioned Metastable Fluid Detector (TMFD) is a transformational technology that is uniquely capable of both alpha and neutron spectroscopy while being “blind” to the intense gamma field that typically accompanies used fuel – simultaneously with the ability to provide multiplicity information as well [1-3]. The TMFD technology was proven (lab-scale) as part of a 2008 NERI-C program [1-7]. The bulk of this report describes the advancements and demonstrations made in TMFD technology. One final point to present before turning to the TMFD demonstrations is the context for discussing real-time monitoring of SNM. It is useful to review the spectrum of isotopes generated within nuclear fuel during reactor operations. Used nuclear fuel (UNF) from a light water reactor (LWR) contains fission products as well as TRU elements formed through neutron absorption/decay chains. The majority of the fission products are gamma and beta emitters and they represent the more significant hazards from a radiation protection standpoint. However, alpha and neutron emitting uranium and TRU elements represent the more significant safeguards and security concerns. Table 1.1 presents a representative PWR inventory of the uranium and actinide isotopes present in a used fuel assembly. The uranium and actinide isotopes (chiefly the Pu, Am and Cm elements) are all emitters of alpha particles and some of them release significant quantities of neutrons through spontaneous fissions

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

A Literature Survey to Identify Potentially Volatile Iodine-Bearing Species Present in Off-Gas Streams

Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3H, 14C, 85Kr, and 129I. Of these, 129I has the longest half-life and potentially high biological impact. Accordingly, control of the release of 129I is most critical with respect to the regulations for the release of radioactive material in stack emissions. It is estimated that current EPA regulations (EPA 2010) would require any reprocessing plant in the United States to limit 129I release to less than 0.05 Ci/MTIHM for a typical fuel burnup of 55 gigawatt days per metric tonne (GWd/t) (Jubin 2011). The study of inorganic iodide in off-gas systems has been almost exclusively limited to I2 and the focus of organic iodide studies has been CH3I. In this document, we provide the results of an examination of publically available literature that is relevant to the presence and sources of both inorganic and organic iodine-bearing species in reprocessing plants. We especially focus on those that have the potential to be poorly sequestered with traditional capture methodologies. Based on the results of the literature survey and some limited thermodynamic modeling, the inorganic iodine species hypoiodous acid (HOI) and iodine monochloride (ICl) were identified as potentially low-sorbing iodine species that could present in off-gas systems. Organic species of interest included both short chain alkyl iodides such as methyl iodide (CH3I) and longer alkyl iodides up to iodododecane (C10H21I). It was found that fuel dissolution may provide conditions conducive to HOI formation and has been shown to result in volatile long-chain alkyl iodides, though these may not volatilize until later in the reprocessing sequence. Solvent extraction processes were found to be significant sources of various organic iodine-bearing species; formation of these was facilitated by the presence of radiolytic decomposition products resulting from radiolysis of tri-n-butyl phosphate and dodecane. Primarily inorganic iodine compounds were expected from waste management processes, including chlorinated species such as ICl. Critical knowledge gaps that must still be addressed include confirmation of the existence and quantification of low-sorbing species in the off-gas of reprocessing facilities. The contributions from penetrating forms of iodine to the plant DF are largely unknown and highly dependent on the magnitude of their presence. These species are likely to be more difficult to remove and it is likely that their sequestration could be improved through the use of different sorbents, through design modifications of the off-gas capture system, or through chemical conversion prior to iodine abatement that would produce more easily captured forms.

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

Sorption Modeling and Verification for Off-Gas Treatment

The project has made progress toward developing a comprehensive modeling capability for the capture of target species in off gas evolved during the reprocessing of nuclear fuel. The effort has integrated experimentation, model development, and computer code development for adsorption and absorption processes. For adsorption, a modeling library has been initiated to include (a) equilibrium models for uptake of off-gas components by adsorbents, (b) mass transfer models to describe mass transfer to a particle, diffusion through the pores of the particle and adsorption on the active sites of the particle, and (c) interconnection of these models to fixed bed adsorption modeling which includes advection through the bed. For single-component equilibria, a Generalized Statistical Thermodynamic Adsorption (GSTA) code was developed to represent experimental data from a broad range of isotherm types; this is equivalent to a Langmuir isotherm in the two-parameter case, and was demonstrated for Kr on INL-engineered sorbent HZ PAN, water sorption on molecular sieve A sorbent material (MS3A), and Kr and Xe capture on metal-organic framework (MOF) materials. The GSTA isotherm was extended to multicomponent systems through application of a modified spreading pressure surface activity model and generalized predictive adsorbed solution theory; the result is the capability to estimate multicomponent adsorption equilibria from single-component isotherms. This advance, which enhances the capability to simulate systems related to off-gas treatment, has been demonstrated for a range of real-gas systems in the literature and is ready for testing with data currently being collected for multicomponent systems of interest, including iodine and water on MS3A. A diffusion kinetic model for sorbent pellets involving pore and surface diffusion as well as external mass transfer has been established, and a methodology was developed for determining unknown diffusivity parameters from transient uptake data. Two parallel approaches have been explored for integrating the kernels described above into a mass-transport model for adsorption in fixed beds. In one, the GSTA isotherm kernel has been incorporated into the MOOSE framework; in the other approach, a focused finite-difference framework and PDE kernels have been developed. Issues, including oscillatory behavior in MOOSE solutions to advection-diffusion problems, and opportunities have been identified for each approach, and a path forward has been identified toward developing a stronger modeling platform. Experimental systems were established for collection of microscopic kinetics and equilibria data for single and multicomponent uptake of gaseous species on solid sorbents. The systems, which can operate at ambient temperature to 250°C and dew points from -69 to 17°C, are useful for collecting data needed for modeling performance of sorbents of interest. Experiments were conducted to determine applicable models and parameters for isotherms and mass transfer for water and/or iodine adsorption on MS3A. Validation experiments were also conducted for water adsorption on fixed beds of MS3A. For absorption, work involved modeling with supportive experimentation. A dynamic model was developed to simulate CO 2 absorption with chemical reaction using high alkaline content water solutions. A computer code was developed to implement the model based upon transient mass and energy balances. Experiments were conducted in a laboratory-scale column to determine model parameters. The influence of geometric parameters and operating variables on CO 2 absorption was studied over a wide range of conditions. This project has resulted in 7 publications, with 3 manuscripts in preparation. Also, 15 presentations were given at national meetings of ANS and AIChE and at Material Recovery and Waste Forms Campaign Working Group meetings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preparation of NO 2 -Aged Silver-Functionalized Silica-Aerogel and Silver Mordenite Samples

Reprocessing used nuclear fuel can result in the volatilization of radioactive gaseous species, including 129 I, into the various process off-gas streams. In order to comply with US regulatory requirements, plant off-gas streams must be treated to remove the iodine prior to discharging the off-gas into the environment. The performance of available gas removal methods depends not only on the concentration of the volatile radioisotope of interest, but also on other constituents that could be present in the reprocessing off-gas streams. Some of the constituents, such as NOx produced during fuel dissolution, are known to have deleterious effects on the capture performance of silver-based sorbents used for iodine removal. Commercially available reduced silver mordenite (AgZ) has an iodine saturation concentration of 7.0-9.0 wt%, and its iodine sorption capacity is reduced by 20-50% as a result of NO2 aging. Silverfunctionalized silica aerogel (AgAerogel), an alternative for iodine capture, has an initial iodine saturation of 29.0 wt% and its iodine capacity is only reduced by 15% from NO2 aging. Understanding the differences in aging behavior between AgZ and AgAerogel is critical to determining the behavior of these sorbents under realistic off-gas conditions. To assist in future technical studies on this topic, samples of both AgZ and AgAerogel were aged with NO2. In the experiment, 10.2190 g of AgZ and 10.1771 g of AgAerogel were exposed to a static 0.75% NO 2 /dry air blend for a period of 28 days. The samples were then removed and stored under argon until needed for future experiments.=

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Iodine Retention of Long-chain Organic Iodides on Silver-based Sorbents under DOG and VOG Conditions

The aqueous reprocessing of used nuclear fuel releases four key volatile radionuclides ( 3 H, 14 C, 85 Kr, and 129 I) from the used fuel into the off-gas streams of a reprocessing facility. Some semi-volatile radionuclides such as 106 Ru may also be released into the gas streams. Compliance with US regulations may require removal of these radionuclides from the off-gas streams before their discharge to the environment (Jubin et al. 2017). Iodine is released during multiple unit operations within a plant and can be found in gas streams such as the shear off-gas, dissolver off-gas, vessel off-gas (VOG), and waste solidification off-gas. To achieve DFs >1,000, treatment of the plant off-gas streams must include, at minimum, removal of both the dissolver off-gas and VOG. A 2015 analysis showing that the speciation of iodine in the VOG is primarily composed of organic alkyl iodides and that abatement technology for organic iodides is not well developed has produced increased interest in removing iodine from VOG. The lack of knowledge surrounding iodine removal from the VOG prompted the Office of Nuclear Energy within the US Department of Energy to initiate experimental efforts targeted at understanding organic iodide removal from prototypic VOG streams. In 2018 a joint test plan was developed and issued by Oak Ridge National Laboratory and Idaho National Laboratory. This test plan is intended to guide a multiyear experimental effort that will provide substantial amounts of information about the removal of organic iodides from VOG streams. This document reports on the progress of Idaho National Laboratory and Oak Ridge National Laboratory in the execution of the joint test plan as of the end of fiscal year 2019. Silver-based sorbents are the primary removal technology being characterized, and this report includes data on iodine adsorption by reduced silver-exchanged mordenite (also called silver-exchanged zeolite [AgZ]) (as a baseline iodine sorbent) and silver-functionalized silica aerogel (AgAero or Ag Aerogel) (as an advanced iodine sorbent). Conclusions are preliminary in nature and reflect the ongoing execution of the test plan through close collaboration between the two laboratories.

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

Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235 U, 239 Pu, and 241 Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241 Pu, a significant fissile constituent in spent fuel, was measured and compared to 239 Pu. The 241 Pu/ 239 Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235 U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an existing framework and codes were adapted as needed for analyzing experiments and assessing application-­specific assay concepts. A de-­convolution analysis of the delayed gamma-­ray response spectra modeled for spent fuel assemblies was performed using the same method that was applied to the experimental spectra.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

A literature review of pyroprocessing safeguards

Pyroprocessing is a promising technology for reprocessing used nuclear fuel (UNF) from light water reactors (LWR) and sodium fast reactors (SFR). With the advancement of fast reactors with projects from companies such as TerraPower, the prospects of pyroprocessing are more promising than ever before, as fast reactors can consume all actinides, not just uranium and plutonium. Before pyroprocessing can be implemented commercially, pre-existing safeguards used for aqueous reprocessing must be adapted for pyroprocessing. Safeguards, as determined by both the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), are discussed in this review as guidelines for approaches used in both domestic and international reprocessing plants. The implementation of safeguards in aqueous reprocessing was then reviewed based on the experience from existing aqueous facilities. The experimental pyroprocessing facilities were identified as pyroprocessing plants in the design stage with one operating exception of the Fuel Conditioning Facility (FCF). The safeguard methods implemented or designed for each and the accompanying challenges of utilizing existing safeguards in pyroprocessing are considered and a summary of applicable approaches is included.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗