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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 703 records · Page 39
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Laser size reduction of contaminated Magnox Pond skips - 15166

The Magnox Pond Program Team at Hinkley Point A Power Station have successfully demonstrated the use of an automated Computer Numeric Controlled (CNC) dry milling process to efficiently decontaminate radioactive pond skips that were once used to store and transport Magnox spent fuel assemblies between reactor sites and Sellafield for reprocessing. This decontamination process has ensured ILW skips meet metal melt recycling criteria, and even Out of Scope levels, while at the same time producing dry swarf that is easily disposed of as ILW. However, the dry milling method requires the skips to be size reduced to a very simple flat geometry to simplify the milling operation. Several size reduction techniques (mechanical and thermal) have been evaluated. Based on the advantages of remote handling, high cutting speed, flexibility with complex geometries, reduced secondary waste production and significantly reduced worker exposure, robotically operated fibre delivered laser cutting was chosen. This paper describes the development, installation and operation of an automated laser cutting system commissioned at Hinkley Point A to size reduce contaminated skips in to five flat sections. It was demonstrated that a complete skip can be remotely size reduced in less than 60 minutes. This successful demonstration, coupled with CNC milling, has the potential to save an estimated British Pounds 30 million across the Magnox sites alone, with much higher saving potentials for skips stored at Sellafield, enabling this new size reduction process to meet cost reduction objectives of the UK's Nuclear Decommissioning Authority. (authors)

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

Evaluation of fuel cycle scenarios on MOX fuel recycling in PWRs and SFRs

Prospects on advanced fuel cycle scenario are considered for achieving a progressive integration of Sodium Fast Reactor (SFR) technology within the current French Pressurized Water Reactor (PWR) nuclear fleet, in a view to benefit from fissile material multi-recycling capability. A step by step process is envisioned, and emphasis is put on its potential implementation through the nuclear mass inventory calculations with the COSAC code. The overall time scale is not optimized. The first step, already implemented in several countries, the plutonium coming from the reprocessing of used Light Water Reactor (LWR) fuels is recycled into a small number of LWRs. The second step is the progressive introduction of the first SFRs, in parallel with the continuation of step 1. This second step lets to prepare the optimized multi recycling of MOX fuel which is considered in step 3. Step 3 is characterized by the introduction of a greater number of SFR and MOX management between EPR reactors and SFRs. In the final step 4, all the fleet is formed with SFRs. This study assesses the viability of each step of the overall scenario. The switch from one step to the other one could result from different constrains related to issues such as resources, waste, experience feedback, public acceptance, country policy, etc.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Safeguarding Reactors and Spent Nuclear Fuel [Slides]

Highly enriched fuel contains large quantities of ²³⁵U; Burned fuel may contain large quantities of ²³⁹Pu; Reprocessing is the major threat (discussed at length later), but we need to ensure the amount of plutonium going to the reprocessing plant is what we believe it to be; All nuclear material needs to be accounted for (material accountancy); There may be a loss of continuity of knowledge in which case fuel needs to be measured to determine how much plutonium and uranium it contains.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

UMo Solutions Processing in La Hague Cold Crucible Induction Melter: the Feedback From the First Years of Operation - 16376

250 cubic meters of high-level liquid waste from reprocessed U-Mo-Sn-Al spent fuel, used in Gas Cooled Reactors (GCR), were produced during the mid-1960's at La Hague facility. These 'UMo' solutions are less radioactive than the current fission product concentrates coming from ongoing reprocessing activities, but are very rich in molybdenum and phosphorus whose contents make the molten glass quite corrosive and require a high-temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum oxide). Hence the use of the Cold Crucible Induction Melter (CCIM) technology to process such solutions has been deemed a good opportunity for AREVA to meet its performance expectations. In addition to being very corrosive, the UMo waste is quite challenging to process as the molybdenum has a strong tendency to stick in the calciner and causes clogging issues in off-gas treatment equipment. Therefore, the process and technological qualifications were deployed in order to address these specific issues. UMo solutions processing in the La Hague CCIM has started in January 2013 and is currently ongoing. During this period (from 2013 to 2015), many data have been collected to confirm the process parameters that were defined during the qualification of this innovating process. Even if some difficulties occurred, operations teams experience along with engineering and R and D support allowed managing them. This paper presents the start-up methodology and the feedback from the first years of UMo solutions processing with the CCIM technology at La Hague site. Lessons learned are presented with the difficulties encountered and the solutions implemented, emphasizing the benefits of a close integration between R and D, engineering and operations teams. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Sellafield FGMSP Additional Sludge Retrievals A Significant Step in Decommissioning Part of the U.K.'s Nuclear Legacy - 16180

The First Generation Magnox Storage Pond (FGMSP) is an open-air pond on the Sellafield site (U.K.) constructed to receive and store spent nuclear fuel from the UK's fleet of Magnox reactors. Operations ceased in 1986, leaving a significant mobile (sludge from the corrosion of fuel cladding) and solids (fuel, fuel fragments and other debris) inventory. In December 2013, Westinghouse Electric Company and EnergySolutions proposed to create an Additional Sludge Retrieval (ASR) capability, with the aim of starting sustainable sludge transfer and hazard reduction as soon as possible. The ASR project team proposed a sludge retrievals concept based on the use of a self-propelled, remote controlled Floating Platform. The deployment system was derived from a Floating Platform concept developed by Westinghouse to support field operations in European nuclear power plants. The system was designed, manufactured and successfully tested in real scale conditions, and was installed and fully commissioned in March 2015. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Regulation of the First British Power Reactor to Enter into an Extended Period of 'Care and Maintenance' - 18028

All eleven of the British first-generation Magnox nuclear power plants have now ceased operating and are in various stages of decommissioning. The Magnox reactors were graphite moderated, using carbon dioxide as the primary coolant. The reactor cores are contained in steel reactor pressure vessels, except for the last two stations which have pre-stressed concrete pressure vessels. All but two of the Magnox reactor stations have been completely de-fueled. The spent fuel removed has been despatched for reprocessing at Sellafield in the north-west of England. On behalf of the UK government the Nuclear Decommissioning Authority (NDA) is responsible for cleaning up the majority of the civil nuclear legacy. The current NDA decommissioning strategy is to prepare the Magnox reactor sites for an extended period of Care and Maintenance (C and M), during which quiescent safe storage will take place (also known as safe enclosure), whereby the cores and associated structures will accrue the benefits of radioactive decay over a period of decades. This will reduce the radiation doses to workers during final dismantling and site clearance. The two Magnox rector units at the Bradwell licensed nuclear site in the south-east of England are the first scheduled to enter this quiescent safe-storage period. This is planned to commence in the second-half of 2018, and is scheduled to last for around 70 years. The Office for Nuclear Regulation (ONR) is the independent regulatory body, responsible for enforcing safety and security requirements at nuclear installations in Great Britain (GB). This paper outlines the national strategic context and how ONR's regulatory regime is applied to nuclear power plant decommissioning. It addresses some of the specific challenges posed at Bradwell during the preparatory works for entry to C and M and how ONR has addressed these. It describes how this will culminate in a regulatory decision on the acceptability of the safety and security of the Bradwell site to enter this new stage of its life cycle. Over the last few years ONR has assessed and inspected various safety aspects of decommissioning, radioactive waste management and other preparations for Bradwell's entry into C and M. Some of the operator's proposals have provided challenges to ONR, particularly where there have been changes to established approaches taken at other sites, as the licensee seeks to minimise its costs. These have included how far to remove redundant structures, systems and components prior to C and M; how to define interim and end-states for fuel cooling ponds and other contaminated areas; and how to treat and store Intermediate Level Waste (ILW), for which there is no geological disposal facility yet available. ONR has begun its final assessment of safety and security justifications supplied by the site licensee, Magnox Ltd (ML). ONR is using its legal powers conferred to it under the site licence conditions to authorise this first-of-a-kind transition into a 70 year period of safe storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Spent fuel nondestructive assay integrated characterization from active neutron, passive neutron, and passive gamma

Spent nuclear fuel comprises a wide range of irradiated isotopic material compositions, and characterization through nondestructive measurements is beneficial in verifying declared parameters before the fuel is placed in storage, final disposal, and/or reprocessed. This paper discusses results from three nondestructive assay instruments, including passive gamma, passive neutron, and active neutron methods, that measured fifty spent fuel assemblies at the Clab interim storage facility in Sweden. Here, integrated analysis of the measurements from the three different instruments allowed parametric assessments of cooling time, burnup, neutron multiplication, fissile mass, initial enrichment, and decay heat of each individual fuel assembly. Passive gamma measurements were found to be the most beneficial in predicting cooling time, passive neutron for determining burnup, active neutron in estimating initial enrichment, and both passive and total neutron for multiplication correlations. Fissile mass was best estimated using any combination of any two of the instruments such that corrections for isotopic changes in the fuel could be accounted for with the first set of measurements and multiplication of the assembly in the second. In conclusion, the nondestructive assay technologies demonstrated through this effort enhance the characterization of spent nuclear fuel assemblies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Applicability of HBU Fuel Cask R and D Project to Al Clad Fuel Storage - 18558

Aluminum clad spent nuclear fuel (SNF) not destined to be reprocessed at Savannah River's H Canyon will eventually be moved from existing wet and dry storage facilities into standardized dry storage systems (e.g., Universal Canisters), ideally suited for disposal. With the current lack of disposal capacity in the U.S., placement of this SNF into dry storage requires plans for demonstrating to the regulator the safety of both the extended storage and the subsequent transportation to a future disposal facility of this SNF. For commercial SNF, concerns related to extended dry storage and subsequent transportation have been, and continue to be, addressed through several demonstration projects including, but not limited to, one for the lower burnup SNF at Idaho National Laboratory (INL) and one for the higher burnup SNF at Dominion's North Anna Power Station. Lessons learned from the more recent Orano led higher burnup demonstration project (HBU Demo Project) provides relevant information for advancing the concept of a 'smart cask' system for dry interim storage, which will allow for collection of data to support the safety demonstration for extended dry storage and subsequent transportation of the aluminum clad SNF. This 'smart cask' would eliminate the need for multiple demonstration projects that may have to be performed for different types of aluminum clad SNF designs and/or for potentially age distinguishing parameters, such as burnup or initial enrichment. This 'smart cask' would be developed and deployed by leveraging the HBU Demo Project's experience associated with modifying and deploying a TN-32B bolted carbon steel vertical storage cask system to monitor key parameters related to the criticality, structural, confinement, and thermal safety of the SNF and the canister system containing this SNF. This paper will focus on how the extensive design, licensing, and fabrication experience accumulated by the HBU Demo Project's cask system can be leveraged to support the design, licensing, and fabrication of a 'smart cask' system intended to support/confirm the interim dry storage, potential extended storage, and subsequent transportation of aluminum clad SNF, such as the Advanced Test Reactor (ATR) SNF. (authors)

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

An interface tracking model for droplet electrocoalescence.

This report describes an Early Career Laboratory Directed Research and Development (LDRD) project to develop an interface tracking model for droplet electrocoalescence. Many fluid-based technologies rely on electrical fields to control the motion of droplets, e.g. microfluidic devices for high-speed droplet sorting, solution separation for chemical detectors, and purification of biodiesel fuel. Precise control over droplets is crucial to these applications. However, electric fields can induce complex and unpredictable fluid dynamics. Recent experiments (Ristenpart et al. 2009) have demonstrated that oppositely charged droplets bounce rather than coalesce in the presence of strong electric fields. A transient aqueous bridge forms between approaching drops prior to pinch-off. This observation applies to many types of fluids, but neither theory nor experiments have been able to offer a satisfactory explanation. Analytic hydrodynamic approximations for interfaces become invalid near coalescence, and therefore detailed numerical simulations are necessary. This is a computationally challenging problem that involves tracking a moving interface and solving complex multi-physics and multi-scale dynamics, which are beyond the capabilities of most state-of-the-art simulations. An interface-tracking model for electro-coalescence can provide a new perspective to a variety of applications in which interfacial physics are coupled with electrodynamics, including electro-osmosis, fabrication of microelectronics, fuel atomization, oil dehydration, nuclear waste reprocessing and solution separation for chemical detectors. We present a conformal decomposition finite element (CDFEM) interface-tracking method for the electrohydrodynamics of two-phase flow to demonstrate electro-coalescence. CDFEM is a sharp interface method that decomposes elements along fluid-fluid boundaries and uses a level set function to represent the interface.

Erickson, Lindsay Crowl↗
DOE OSTI2016

Building Effective Collaborations to Bring Innovation into Waste Management and Decommissioning - 16477

A strong record of research collaboration has been developing between the UK nuclear industry and key universities with nuclear specialism over the past 15 years or so. The DISTINCTIVE program, focusing on decommissioning, immobilization and storage solutions for nuclear waste inventories is an example. It is a consortium of eleven universities funded by the Engineering and Physical Sciences Research Council, National Nuclear Laboratory, Nuclear Decommissioning Authority and Sellafield Limited. The motivation for industry collaboration with universities is explained and the current progress of DISTINCTIVE in the context of these motivations is assessed. (authors)

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

Safe solution dissolution of high flux isotope reactor cores - 14490

The H-Canyon facility at the Savannah River Site is the U.S.A.'s only remaining full-scale chemical separations plant. H-Canyon maintains the capacity to safely and efficiently disposition a large inventory of excess nuclear material from across the Department of Energy (DOE) complex and used nuclear fuel from foreign and domestic research reactors. One such fuel that H-Canyon may dissolve is cores from the High Flux Isotope Reactor (HFIR). By employing realistic analysis, the need for administrative controls was eliminated. A new criticality modeling approach has eliminated the need for administrative controls for HFIR dissolution operations in H-Canyon. The primary upsets of concern are an over-mass event and an overconcentration event. Several initiating events were considered for each type of upset. Over-concentration events were deemed to be not credible by taking credit for the chemistry of acid consumption, i.e. there are not enough moles of acid to dissolve enough moles of uranium in order to reach a critical configuration. Over-mass events were deemed to be not credible by taking credit for the annular design of the fuel elements. By using realistic analysis, both types of upsets have been shown to be subcritical, and therefore administrative controls are not needed to prevent an inadvertent criticality. In this paper, the new methodology for performing criticality analyses for dissolving HFIR fuel elements will be presented, and the elimination of administrative controls will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fuel cycle cost uncertainty from nuclear fuel cycle comparison

This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for the discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Measuring Nd(III) Solution Concentration in the Presence of Interfering Er(III) and Cu(II) Ions: A Partial Least Squares Analysis of Ultraviolet–Visible Spectra

Optical spectroscopy is a powerful characterization tool with applications ranging from fundamental studies to real-time process monitoring. However, it can be difficult to apply to complex samples that contain interfering analytes which are common in processing streams. Multivariate (chemometric) analysis has been examined for providing selectivity and accuracy to the analysis of optical spectra and expanding its potential applications. Here we will discuss chemometric modeling with an in-depth comparison to more simplistic analysis approaches and outline how chemometric modeling works while exploring the limits on modeling accuracy. Understanding the limitations of the chemometric model can provide better analytical assessment regarding the accuracy and precision of the analytical result. This will be explored in the context of UV–Vis absorbance of neodymium (Nd 3+ ) in the presence of interferents, erbium (Er 3+ ) and copper (Cu 2+ ) under conditions simulating the liquid–liquid extraction approach used to recycle plutonium (Pu) and uranium (U) in used nuclear fuel worldwide. Finally, the selected chemometric model, partial least squares regression, accurately quantifies Nd 3+ with a low percentage error in the presence of interfering analytes and even under conditions that the training set does not describe.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Utilization of Used Nuclear Fuel in a Potential Future US Fuel Cycle Scenario - 13499

To date, the US reactor fleet has generated approximately 68,000 MTHM of used nuclear fuel (UNF) and even with no new nuclear build in the US, this stockpile will continue to grow at approximately 2,000 MTHM per year for several more decades. In the absence of reprocessing and recycle, this UNF is a liability and needs to be dealt with accordingly. However, with the development of future fuel cycle and reactor technologies in the decades ahead, there is potential for UNF to be used effectively and efficiently within a future US nuclear reactor fleet. Based on the detailed expected operating lifetimes, the future UNF discharges from the existing reactor fleet have been calculated on a yearly basis. Assuming a given electricity demand growth in the US and a corresponding growth demand for nuclear energy via new nuclear build, the future discharges of UNF have also been calculated on a yearly basis. Using realistic assumptions about reprocessing technologies and timescales and which future fuels are likely to be reprocessed, the amount of plutonium that could be separated and stored for future reactor technologies has been determined. With fast reactors (FRs) unlikely to be commercially available until 2050, any new nuclear build prior to then is assumed to be a light water reactor (LWR). If the decision is made for the US to proceed with reprocessing by 2030, the analysis shows that the UNF from future fuels discharged from 2025 onwards from the new and existing fleet of LWRs is sufficient to fuel a realistic future demand from FRs. The UNF arising from the existing LWR fleet prior to 2025 can be disposed of directly with no adverse effect on the potential to deploy a FR fleet from 2050 onwards. Furthermore, only a proportion of the UNF is required to be reprocessed from the existing fleet after 2025. All of the analyses and conclusions are based on realistic deployment timescales for reprocessing and reactor deployment. The impact of the delay in recycling the UNF from the FRs due to time in the core, cooling time, reprocessing, and re-fabrication time is built into the analysis, along with impacts in delays and other key assumptions and sensitivities have been investigated. The results of this assessment highlight how the UNF from future reactors (LWRs and FRs) and the resulting fissile materials (U and Pu) from reprocessing can be effectively utilized, and show that the timings of future nuclear programs are key considerations (both for reactors and fuel cycle facilities). The analysis also highlights how the timings are relevant to managing the UNF and how such an analysis can therefore assist in informing the potential future R and D strategy and needs of the US fuel cycle programs and reactor technology. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Ab Initio Evaluation of Plutonium Dioxide $S(α,β)$ and Thermal Neutron Cross Sections

Plutonium Dioxide (PuO 2 ) is an advanced fuel-cycle nuclear fuel material for thermal and fast reactors, with a melting point at 3261.8 K and thermal conductivity above 3 Wm -1 K -1 for temperatures below 1500 K. Plutonium from spent UO 2 is reprocessed to form chemically stable mixed oxide (MOX) fuel. In contrast to metallic fuels, the low PuO 2 conductivity results in high thermal gradients between fuel center and surface, resulting in structural annealing and efficient fission product migration over burnup. Crystal binding affects the doppler broadening of epithermal resonances for nuclear fuels and scattering behavior of thermal neutrons. Currently, there is no ENDF/B Thermal Scattering Law, i.e., $S(α,β)$, evaluation for PuO 2 , only free atom cross sections for Plutonium and Oxygen, respectively, which neglect lattice contributions to total cross section. Herein, ab initio lattice dynamics (AILD) techniques are employed to calculate the phonon density of states (DOS) using spin-orbit-coupling density functional theory (DFT) to predict the PuO2 paramagnetic ground-state structure. These results highlight the evaluation of $S(α,β)$ for Plutonium and Oxygen, respectively in PuO 2 and consequential generation of thermal neutron scattering cross sections for high fidelity criticality safety analysis and reactor calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗