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DOE OSTI2019

CoDCon Project: FY 2019 Status Report

The co-decontamination (CoDCon) project was established in FY 2016 with the objectives of (a) evaluating the uncertainty in the uranium (U)/plutonium (Pu) ratio in a mixed U/Pu product from a tributyl phosphate (TBP)–based solvent extraction flowsheet and (b) developing and demonstrating on-line optical spectroscopy for real-time monitoring of key components (e.g., Pu, U, and HNO 3 concentrations) in the process solutions. We are interested in establishing the accuracy and precision to which a specific uranium-to-plutonium (U/Pu) ratio can be achieved, which for the purposes of this project is set at a U/Pu mass ratio of 7/3. The uncertainty associated with achieving this specific target U/Pu ratio is being established through multiple flowsheet tests using laboratory-scale solvent extraction equipment. In addition, optical spectroscopic techniques have been incorporated into the CoDCon solvent extraction testing system, allowing real time monitoring of all input and output process streams. Two CoDCon flowsheet tests were successfully performed in FY 2018. In FY 2019, two additional flowsheet tests were performed. For the first of these (CoDCon Run 3), the dissolved fuel simulant was similar to that used in the first two tests, with the inclusion of 1 mM neptunium (Np). The second test conducted in FY 2019 (CoDCon Run 4) used a more representative dissolved fuel simulant, including addition of non-radioactive fission product elements. All tests used a bank of sixteen 2 cm centrifugal contactors. The tests involved first loading the solvent (30 vol% TBP dissolved in n-dodecane) with U and Pu, then the Pu was stripped from the loaded solvent with a U(IV) solution (~50 mM) and the flowsheet conditions were adjusted such that some U partitioned into the Pu-containing product stream. The amount of U accompanying the Pu was monitored in real time using optical spectroscopic techniques coupled with chemometric modeling. Based on the real-time spectroscopic measurement of the U/Pu ratio, adjustments were made to the flowrate of the fresh TBP solvent phase used to scrub U from the aqueous Pu-containing product. This proved to be a very effective way to control the U/Pu mass ratio in the product.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Developing a new criticality methodology to support decommissioning of legacy plutonium facilities

To decommission a legacy plutonium finishing facility, a number of gloveboxes need to be disposed of. These gloveboxes have been in place since the 1960's and were operational until the 1980's. A period of initial clean out and decommissioning took place in the 1990's and the first gloveboxes were removed from the line in the early 2000's. Compliance with downstream waste handling facility requirements meant that the gloveboxes needed to be removed using intensive 'hands-on' methods in order to size reduce them to fit into 200 litre (55 gallon) drums. This is the standard approach for handling waste items known as Plutonium Contaminated Material (PCM) on the Sellafield site and was credited in the criticality safety case for the decommissioning of the gloveboxes as it allowed each waste drum to be individually assayed such that the fissile content of the waste could be known. Not only was this approach time-consuming and laborious but as it involved operators working with cutting tools in a highly contaminated area it was challenged as being overly restrictive to operations and potentially hazardous to the workers. In order to decommission the remaining nine gloveboxes in the facility, a new decommissioning approach and associated criticality methodology has been developed. Instead of intensive size reduction and waste drum assay, the gloveboxes will be removed with little or no size reduction and put into storage crates without assay. The criticality assessor will review the operational history of each glovebox, along with the decommissioning records and relevant plant data, to make an evidence-based judgement that the crated glovebox represents no greater a residual criticality risk than would have been realised from the equivalent number of PCM drums that the standard decommissioning approach would have generated. This method identifies, in conjunction with the plant personnel, a number of review points to ensure that before the glovebox is removed any bulk residue material has been recovered and no more than a safe amount of moderating material (fixative) has been added to fix any remaining fissile contamination. Finally, once these review points have been confirmed to be complete - through a stakeholder meeting with criticality safety, plant operators and project managers - the glovebox will be placed in a crate, and this crate will be declared a Low Criticality Risk crate. No formal assay will be required prior to exporting the crate to the downstream waste handling facility and no fissile mass value will be assigned to the contents of the crate. No further intrusive work will be required on the crate and it can be safely stored without any spacing requirements, pending final disposal. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Hestia SW-IFL Onroad Fossil Fuel Carbon Dioxide (FFCO2) product: Road segment-level annual FFCO2 emissions across Arizona (2017-2022), version 1.1

The SW-IFL onroad fossil fuel carbon dioxide (FFCO2) emissions data product represents CO2 emissions from the combustion of fossil fuels by motor vehicles (e.g., passenger cars, trucks, buses, motorcycles) traveling on designated roadways. The emissions are represented geographically on each road segment within the state of Arizona spanning the 2017 to 2022 time period. This data product was developed as part of the Southwest Urban Corridor Integrated Field Laboratory (SW-IFL) project, which aims to provide new knowledge and tools that address urban environmental issues by integrating high-resolution observations, modeling, and civic engagement. The emissions data are provided in CSV (input data, ONR_FFCO2_AZ_county.csv) and GeoPackage form (output polyline objects - about 786,000 road segments, XXXX_AZ_v1.1.gpkg) designated by road class (interstates, arterials, collectors, local). The metadata file (Metadata_SW-IFL_Onroad_annualFFCO2_v1.1.docx) provides details about attributes and data formats. The GeoPackage emissions data are provided separately for local roads and nonlocal roads (interstates, arterials, collectors). The method file (Methods_SW-IFL_Onroad_annualFFCO2_v1.1.docx) describes the data processing flow and data sources. Update on 2024-04-17: Updates were made to both the input emission data file (.csv) and output segment-level emission file (.gpkg). There was an update in county-level emission input data (ONR_FFCO2_AZ_county.csv) and the entire road segments were reprocessed to reflect this update.Update on 2024-04-29: Update was made to one output segment-level emission file (Nonlocal_AZ_v1.0.gpkg). There was an error in the AADT values and the data were reprocessed to reflect this update.Update on 2024-10-22: Temporal coverage was extended to include 2022. VMT values were recalculated using new AADT data and the entire road segments were reprocessed to reflect these updates.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2014

The effect of birthrate granularity on the release-to-birth ratio for the AGR-1 in-core experiment

The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-h measurements. Birth rate calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rate using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-h activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-h release activity. The results of this study are presented in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Proliferation Persuasion. Coercive Bargaining with Nuclear Technology

Why do states wait for prolonged periods of time with the technical capacity to produce nuclear weapons? Only a handful of countries have ever acquired the sensitive nuclear fuel cycle technology needed to produce fissile material for nuclear weapons. Yet the enduring trend over the last five decades is for these states to delay or forgo exercising the nuclear weapons option provided by uranium enrichment or plutonium reprocessing capabilities. I show that states pause at this threshold stage because they use nuclear technology to bargain for concessions from both allies and adversaries. But when does nuclear latency offer bargaining benefits? My central argument is that challengers must surmount a dilemma to make coercive diplomacy work: the more they threaten to proliferate, the harder it becomes to reassure others that compliance will be rewarded with nuclear restraint. I identify a range of mechanisms able to solve this credibility problem, from arms control over breakout capacity to third party mediation and confidence building measures. Since each step towards the bomb raises the costs of implementing these policies, a state hits a sweet spot when it first acquires enrichment and/or reprocessing (ENR) technology. Subsequent increases in proliferation capability generate diminishing returns at the bargaining table for two reasons: the state must go to greater lengths to make a credible nonproliferation promise, and nuclear programs exhibit considerable path dependency as they mature over time. Contrary to the conventional wisdom about power in world politics, less nuclear latency thereby yields more coercive threat advantages. I marshal new primary source evidence from archives and interviews to identify episodes in the historical record when states made clear decisions to use ENR technology as a bargaining chip, and employ this theory of proliferation persuasion to explain how Japan, North Korea, and Iran succeeded and failed to barter concessions from the United States. By clarifying when countries are able to leverage steps towards the bomb for international political gain, my work advances our understanding of proliferation and coercive diplomacy.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2015

Characterization and Modeling of Materials for Kr-Xe Separations

We sought to identify practical adsorbents for the separation of Kr from Xe through pressure swing adsorption. We spent appreciable efforts on two categories of materials: metal-organic frameworks (MOFs) and zeolites. MOFs represent a new and exciting sorbent with numerous new framework topologies and surface chemistries. Zeolites are widely used and available commercial adsorbents. We have employed a combination of gas sorption analysis to analyze gas – surface interactions, computational modelling to both aid in interpreting experimental results and to predict practical adsorbents, and in-situ crystallographic studies to confirm specific experimental results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1961

Laboratory Development of a Pot Calcination Process for Converting Liquid Wastes to Solids

Simulated Darex, Purex, Thorex, and TBP-25 wastes were batch-evaporated and calcined to solids. The weight of residue was decreased to ~ 8% of the original waste for Darex, to ~ 4% for Purex, to ~ 2% for Thorex, and ~ 7% for TBP-25 by calcination to 1200°C. Semicontinuous evaporation and calcination to solids at 900°C of Darex and Purex wastes gave volume reduction factors of ~ 8 and 7 to 10, respectively. The nitrate contents of the residues from batch calcination to 800°C were from ~ 0.07 to 0.5 wt % while the nitrate contents of the residues from semicontinuous calcination to 900°C were from ~ 0.1 to 0.7 wt %. Sodium, calcium, and magnesium additives to Purex waste decreased the percent of sulfate volatilized into the condensate from ~ 30% to <0.5%. Replacement of the atmosphere above the waste with nitric oxide decreased ruthenium volatility from 50 to 60% of that originally present in Purex waste to 0.5 to 3.5%. The thermal conductivities for the calcined wastes, measured in situ, were all >0.1 Btu/hr ft °F at ~ 400°F and increased almost linearly with increasing temperature to >0.3 Btu/hr ft °F at ~ 1600°F in all cases studied.

052001* -- Nuclear Fuels-- Waste Processing↗

DOE OSTI2020

Malleable Thermosets (Vitrimers) from CO 2 and Plants

Fiber-reinforced composite materials are increasingly used to replace heavy metal components in transportation applications for lightweighting purposes. Polymer matrix materials used in automotive parts, such as epoxies, nylons, and polypropylenes, are almost all produced from fossil fuels. Higher performance thermoset composites are generally not repairable and lower-cost thermoplastic composites do not have enough performance to replace steel. Vitrimers are a new class of polymer materials with long shelf life, low waste in production, ease of processing, and repairability that have the potential to break through the high-cost thermoset/low-performance thermoplastic barrier to vehicle light-weighting. Vitrimers also exhibit easier chemical recyclability than conventional thermosets with reversible chemical crosslinking that enables reprocessing of scrap from production and end of life components. Vitrimers are a new class of engineered plastics that are weldable, repairable and recyclable like thermoplastics but have high mechanical properties like thermosets. Bio-based vitrimers have only recently been reported in the academic and patent literature. These materials represent an unusual opportunity to meet both VTO light-weighting targets for low-density, high performance materials and BETO sustainability targets for value-added high-volume applications of bio-fuel byproducts. Also, PNNL researchers have previously demonstrated a process to convert CO 2 into potential vitrimer precursors. Washington State University (WSU) success in demonstration of bio-based vitrimers and recent steps toward commercialization of petroleum-based vitrimers support the opportunity for PNNL success in development of sustainable vitrimer materials for recyclable, high-performance fiber-reinforced composites for the transportation sectors.

36 MATERIALS SCIENCE↗

DOE OSTI2018

The Strategic Benefit of Self Shielded Boxes for Storage and Disposal of Metallic Uranic Material-18141

Sellafield Ltd, on behalf of the Nuclear Decommissioning Authority (NDA), is responsible for the management of a number of facilities whose inventories include a range of metallic uranic material. Additional fuel and fuel bearing materials exist in the NDA estate, some of which are similar to the Sellafield-managed inventory and others present different management challenges. Whilst UK Government policy directs that some of these materials are assets and are therefore managed with eventual future use in mind, there remains a large quantity of material, principally either out-of-scope for reprocessing or which derives from legacy facilities, which is likely to be disposed of to a future Geological Disposal Facility (GDF). The majority of this material is stored under wet conditions to facilitate dose shielding and heat dissipation and will therefore require removal from wet storage, consolidation and conditioning, prior to transfer to GDF. In order to accelerate high hazard and risk reduction by sooner retrieval of legacy material from wetted storage, one of the legacy storage facilities has led technical work to develop the design of Self Shielded Boxes (SSBs) and to substantiate their use for interim storage of the materials in question. Use of the SSBs would represent a change of strategy on the Sellafield Site: historically, timescales for retrieval of material have been reliant upon a waste treatment and storage/disposal route being available. SSBs provide the opportunity to retrieve uranic materials from aging legacy storage facilities for immediate buffer storage, by decoupling the timing of retrieval from the availability of an onward route. An understanding was required of the scope for defining optimal routes for each metallic uranic fuel and fuel bearing material group for safe transport and storage commensurate with current final disposal assumptions and risks, taking into account international best practice. A study is underway to understand options for transfer of material into SSBs, buffer/interim storage locations, treatment/conditioning processes and transport to the GDF. This work forms the basis of a prioritised road-map for the development of routes for study material groups aligned with ongoing and existing planned technical work, which is a pre-requisite to strategy development. The ultimate output of the study will be identification of full life cycle management strategies for metallic uranic material from an enterprise perspective. The combination of this strategic work and ongoing technical work will underpin a series of decisions to divert sub-groups of uranic material away from currently assumed baseline routes in support of earliest practicable hazard and risk reduction. Hence, each decision will reduce the scope of a placeholder future uranic fuel treatment capability, which will diminish the associated British Pound 1 bn capital cost. To date, a technical case for interim storage of uranic material in SSBs has been substantiated for around two thirds of the study materials. The requirement for a new strategy at Sellafield was driven by technical development of SSBs in response to the needs of one programme, but this prompted the study into the potential utilisation of SSBs in order to derive benefit for the wider enterprise and across the NDA estate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Nuclear Fuel and Pu Redox Studies from The Glenn T. Seaborg Institute at Idaho National Laboratory

The Glenn T. Seaborg Institute at Idaho National Laboratory (INL-GTSI) focuses on advancing fundamental research in the actinide sciences by providing unique opportunities to early career scientists and engineers to gain experience studying the actinide elements and their associated systems. The INL-GTSI is built from three focus areas that are based on the expertise and supporting infrastructure at INL and include solid state chemistry and physics, solution phase chemistry and physics, and forensic and isotope science. INL is the lead Laboratory for nuclear energy research and development in the U. S. and the research on nuclear fuels performed under the INL-GTSI gives good examples of solid state studies. Uranium-Molybdenum (U-Mo) alloys are leading fuel candidates for conversion of high performance research and test reactors to low-enriched fuels. During irradiation, generated fission gas accumulates into bubbles and self-organizes into a gas bubble superlattice (GBS) that effectively stores fission gases and inhibits fuel swelling. A study on the early self- organizing behavior of the GBS shows that not only grain boundaries but the interfaces between the U-Mo matrix and uranium carbide (UC) impurities are important to GBS formation.[1] In solution, understanding the complex redox behavior of plutonium in aqueous environments is critical for establishing optimized nuclear waste reprocessing solvent systems and storage tank environments. INL-GTSI researchers have produced an experimentally validated multi-scale model of the gamma radiation induced behavior of plutonium ions in concentrated aqueous HNO3 solutions.[2] Here, gamma radiation effected only minimal steady state changes in the redox distribution of the plutonium oxidation states. The redox cycling between Pu(IV) and Pu(III) is demonstrated to be mediated by the •OH/NO3• radical oxidation of Pu(III) and the H2O2/HNO3 driven reduction of Pu(IV). The INL-GTSI offers young researchers the unique chance to work directly with actinide bearing materials in a U. S. National Laboratory environment. Further topical areas of interest to the INL-GTSI include, but are not limited to, fundamental actinide properties, structure/property (electronic, magnetic, thermal) relations, actinide quantum criticality, f- electron interactions, electron correlations, computational studies, new phases, defect effects, interface interactions, isotope production and separation, forensic analytical chemistry, structure and dynamic properties of actinides in non-aqueous media, separations chemistry and kinetics for advanced nuclear fuel cycles, radiation effects, and innovative and advanced ligand design for complexation of the actinides.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Application of Hybrid K-Edge Densitometry in Reprocessing Facilities

During reprocessing operations, accurately determining the elemental concentrations of uranium and plutonium is critical for nuclear security. Reprocessing facilities processes hundreds of tons of nuclear material annually, requiring accurate measurements to ensure effective nuclear material control and accountability. Hybrid K-edge densitometry (HKED) system combines K-edge absorption densitometry with x-ray fluorescence to measure actinide elemental concentrations with a low uncertainty. This system uses high-purity germanium gamma-ray detectors and advanced signal processing equipment to detect x-ray accurately. This document provides guidance on how to achieve effective performance from an HKED system for measuring uranium and plutonium concentrations in reprocessing facilities. It includes best practices for the setup and operation of an HKED system, while highlighting factors influencing uncertainties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Characterization of solids deposited on the modular caustic-side solvent extraction unit (MCU) coalescer media removed in May and October 2014

During routine maintenance, the coalescers utilized in the Modular Caustic-Side Solvent Extraction Unit (MCU) processing of Salt Batch 6 and a portion of Salt Batch 7 were sampled and submitted to the Savannah River National Laboratory (SRNL) for characterization, for the purpose of identifying solid phase constituents that may be accumulating in these coalescers. Specifically, two samples were received and characterized: A decontaminated salt solution (DSS) coalescer sample and a strip effluent (SE) coalescer sample. Aliquots of the samples were analyzed by XRD, Fourier Transform Infrared (FTIR) Spectroscopy, SEM, and EDS. Other aliquots of the samples were leached in acid solution, and the leachates were analyzed by ICP-AES. In addition, modeling was performed to provide a basis for comparison of the analytical results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Separation of Zirconium from Zircaloy-2 for Spent Nuclear Fuel Cladding Applications using a Chloride Volatility Process

The United States produces a large amount of cladding material from used nuclear fuel, expected to approach 1,000 MT/year for the next 50 years. The recovery of zirconium from the cladding material would represent a ∼50% savings on the geologic waste disposal requirement, and savings of >$40 M/year due to the cost of hafnium-free zirconium. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 93m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. In a chlorination method, zirconium cladding reacts with Cl{sub 2} at 350-380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low boiling point (sublimes at 330 deg. C) which allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. This chlorination approach has been proposed for the reprocessing of zirconium. However, previous work has observed element impurities (Fe, Sn, Cr, Cs, Sb, and Nb) in the recovered ZrCl{sub 4} meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl{sub 4} to the metallic form. The chlorination process for zirconium reprocessing has been highly encouraged as it has proven more efficient than previous mechanical processes, redox processes, and other chemical methods such as iodination. The overall goal of this project is to develop the road map for a chlorination process that will yield high purity zirconium. The current study focuses on the recovery of ZrCl{sub 4} from non-irradiated alloys used as nuclear fuel cladding. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage

This final report is summarized four years of work of awarded project number DE-NE0008442\IRP-15-9318, title as: “Innovative Approach to SCC Inspection and Evaluation of Canister in Dry” supported by Nuclear Energy University Program\Integrated Research Program, NEUP\IRP, where last year is an extension of no cost project. This research work is focused on the possible degradation of dry storage canister holding used nuclear fuel (UNF), due to exposure of marine atmospheric environment. The canister usually designed from a welded austenitic stainless-steel of types 304 or 316, encased in a concrete shell known as an overpack. This canister designed and licensed for 20 years with an extension period need to be determined, in addition 20 years or more. These casks are located throughout the country in Independent Spent Fuel Storage Installations (IFSFIs) often on the site close to nuclear power plants, that are often located in coastal or lake-side regions where a marine atmospheric salt air environment exist. As a result of several recent decisions, such as not pursuing Yucca Mountain as a long-term geologic repository storage facility for UNF and the cancellation of the construction of a reprocessing facility in the 1980s due to proliferation risks, interim dry storage facilities for long-term storage are vital option for the immediate future of nuclear energy technology. One of the primary concerns with respect to the long-term performance of the storage casks is the potential for corrosion initiation due to deliquescence of salts deposited on the canister surface as aerosols; in regions of high residual weld stresses, this may lead to localized stress corrosion cracking (SCC). Dust and aerosols in the air being drawn though ventilation openings in the overpacks of passively ventilated dry canister storage systems may be deposited on the stainless-steel canister outer surfaces. Under these conditions, localized corrosion attack can occur. Chlorides ions deliquescence on the canister surface, a process by which the salts essentially pull moisture out of the air, creating a concentrated salt brine on the canister surface. Chloride-induced stress corrosion cracking (CISCC) of welded zones is of special concern, as it a well-documented mode of attack for austenitic stainless steels marine environments, and many independent spent fuel storage installations (ISFSIs) are located in coastal areas. Recent canister inspections have shown that chloride salts are present on the surface of in-service canisters at near-marine settings. However, canister surface inspections of satisfactory resolution to detect SCC have never been carried out in details, because access to the canister surfaces through vents in the overpacks is extremely limited, and high radiation fields make removal of the canisters from the overpacks undesirable. Although no significant corrosion damage has not been detected up to date, there is concern that this may lead to pitting corrosion, which could in turn develop into stress corrosion cracking, causing canister structural failure. The initiation and propagation of pitting corrosion has been widely studied in full immersion conditions, but limited studies have been done on marine atmospheric conditions, which is a focus of this work, these includes laboratory and field experimentally and propagation rate modeling of pitting and transition to crack, and crack propagation rate. It is known, however, that the rate of atmospheric corrosion can be affected by variations in relative humidity (RH), salt density, salt type, droplet area, temperature and stainless-steel microstructure and residual stress particularly in welded zones of canisters. The integrated research program has identified potential deterioration mechanisms for the stainless-steel canisters containing the UNF in dry storage that require detailed research and investigation and will have an impact on the performance of long-term interim storage under the normal and extreme environmental conditions experienced during the duration of this storage. The prediction and monitoring of canister corrosion processes while in storage can provide important data for the assessment of interim storage performance and the safety to the public. This report comprises of mainly the published results in different journals and covering a wide range spectrum from susceptible stainless-steel types 304\316 microstructural analysis through the implant test at atmospheric environment, residual stress analysis of mockup canister and simulation, inspection methodology includes nonlinear ultrasound and vibro-thermography and stochastic modeling and simulation of pits propagation rates and transition to cracks. Description of field specimens installed at ISFSI of Main Yankee is given in the following chapters and will be analyzed in the coming future and serves as an important data for a stochastic analysis as it was drawn from the publications and reports described in this report. The purpose of this report is to bring together the project efforts that includes improving in our understanding of structural aging performance at current dry storage casks fleet at ISFSIs in the context of remedial maintenance and inspection applications technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Rapid Automated Dissolution and Analysis Techniques for Radionuclides in Recycle Process Streams

The analysis of process samples for radionuclide content is an important part of current procedures for material balance and accountancy in the different process streams of a recycling plant. The destructive sample analysis techniques currently available necessitate a significant amount of time. It is therefore desirable to develop new sample analysis procedures that allow for a quick turnaround time and increased sample throughput with a minimum of deviation between samples. In particular, new capabilities for rapid sample dissolution and radiochemical separation are required. Most of the radioanalytical techniques currently employed for sample analysis are based on manual laboratory procedures. Such procedures are time- and labor-intensive, and not well suited for situations in which a rapid sample analysis is required and/or large number of samples need to be analyzed. To address this issue we are currently investigating radiochemical separation methods based on extraction chromatography that have been specifically optimized for the analysis of process stream samples. The influence of potential interferences present in the process samples as well as mass loading, flow rate and resin performance is being studied. In addition, the potential to automate these procedures utilizing a robotic platform is evaluated. Initial studies have been carried out using the commercially available DGA resin. This resin shows an affinity for Am, Pu, U, and Th and is also exhibiting signs of a possible synergistic effects in the presence of iron.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1976

Cryogenic recovery

Because of the low boiling temperature of cryogenic propellants to be used on the Space Shuttle, loss of cryogens from boiloff could become very costly. This paper describes how this shuttle problem is being solved at Kennedy Space Center. Cryogenic losses are categorized relative to the particular cryogenic involved, the Space Shuttle servicing operation causing boiloff and the magnitude of the loss. The techniques under consideration are discussed in detail. These techniques include reclaiming the boiloff by reliquefaction, upgrading the reclaimed boiloff by purification, and interim boiloff storage in metal hydride prior to reprocessing. One of the reliquefaction processes discussed in detail utilizes the cooling effect of venting some of the liquid hydrogen boiloff to provide a simple hydrogen reliquefaction unit. Possible future applications of these cryogenics recovery techniques to industry and transportation systems using liquid hydrogen for energy storage and fuel are also discussed.

Howard, F. S.↗

DOE OSTI1993

Nuclear Safety [Vol. 34, No. 1, January-March 1993]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 An Integrated Regional Approach To Risk Management of Industrial Systems, S. Chakraborty and R. Stratton; 9 Annual Technical Meeting of the NRC Cooperative Severe Accident Research Program, E. G. Silver; ACCIDENT ANALYSIS: 13 Deterministic Severe Accident Criteria as Severe Accident Design Criteria and Policy for the New Production Reactor-Heavy Water Reactor, P. T. Rhoads; 20 Proposed Deterministic Severe Accident Criteria for the Heavy Water Reactor-New Production Reactor Containment, K. D. Bergeron, S. E. Slezak, and C. E. Leach; 33 Quasi-Static Core Liquid Level Depression and Long-Term Core Uncovery During a PWR LOCA, Y. Kukita, R. R. Schultz, H. Nakamura, and J. Katayama; 49 LIRA: An Advanced Containment System to Minimize the Accidental Radioactivity Releases, A. Turricchia; 63 Errata to “A Review of Hydrogen Production During Melt/Water Interaction in LWRs,” Vol. 33, No. 4; CONTROL AND INSTRUMENTATION: 64 The Nuplex 80+™ Advanced Control Complex from ABB Combustion Engineering, F. Ridolfo, D. Harmon, and K. Scarola; DESIGN FEATURES: 76 A Scheme for Passive Isolation of the Containment of a Reactor, A. K. Ghosh, V. V. Raj, and K. Kakodkar; 84 Utility Requirements for Safety in the Passive Advanced Light-Water Reactor, T. U. Marston, W. H. Layman, and G. Bockhold, Jr.; WASTE AND SPENT FUEL MANAGEMENT: 91 Activities Related to Waste and Spent Fuel Management, Compiled by M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 103 Testing Deficiencies in Auxiliary Systems Feedwater Systems, J. D. Kueck; 110 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 113 Selected Safety-Related Events, Compiled by G. A. Murphy; 115 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; RECENT DEVELOPMENTS: 133 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 145 Reports, Standards, and Safety Guides, D. S. Queener; 150 Proposed Rule Changes as of Sept. 30, 1992; ANNOUNCEMENTS: 75 Harvard Short Course on Risk: Science, Assessment, and Management; 102 International Conference on Reactor Physics and Reactor Computations; 102 The Fourth International Conference on Nuclear Waste Reprocessing and Waste Management [RECOD ’94]; 132 International Symposium on the Radiological Exposure of the Population of the European Community from Radioactivity in the Mediterranean Sea [MARINA-MED]; 162 International High-Level Radioactive Waste Conference ‘Technology Integration—Above and Below”; 154 The Authors; 158 Indexes to Nuclear Safety, Volume 33.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1993

Nuclear Safety [Vol. 34, No. 1, January-March 1993]

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗