Search NASA⌕ Search

SEARCH · Search NASA

Results for “Cladding design”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Investigating Accident Tolerant Fuel Concepts using the BISON Fuel Performance Code

There has been an increased interest in research of accident tolerant fuels (ATF) and claddings for use in commercial reactors ever since the events that occurred at the Fukushima Daiichi nuclear power plant in 2011. The economic impact of these events on the nuclear energy sector is significant. As a result, the United States Department of Energy's Office of Nuclear Energy accelerated research in nuclear fuel concepts with enhanced accident tolerance as part of its Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). A report produced by the AFC entitled Light Water Reactor Accident Tolerant Fuel Performance Metrics outlines a set of metrics that can be used to guide selection of promising ATF concepts. Moreover, the report specifies that a down-selection is to occur in the 2016-2017 time frame, at which time the program will move from a proof-of-concept to a proof-of-principle stage with continued research and development into a small set of concepts. Given the time necessary for comprehensive irradiation experiments and post-irradiation examination, it is impossible to perform a complete set of experiments for material characterization. Therefore, the AFC is utilizing computational analysis tools in an effort to understand the proposed materials under a variety of normal operating and accident conditions. For some time, the DOE's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has been developing computational analysis tools for nuclear fuel simulation. These codes include BISON and Marmot which investigate fuel behavior at the continuum and microstructural scales, respectively. In 2014, NEAMS introduced what it calls High Impact Problems (HIPs) into its program plan. These HIPs are intended to make significant advancements in a particular area of nuclear power research in a short period of time (3 years or less). NEAMS has chosen an ATF project, which emphasizes utilizing BISON and Marmot to model proposed materials, as its first HIP. This summary focuses on the methodology being utilized for investigation of ATF concepts using the BISON fuel performance code. First, we give an overview of the model development for two candidate ATF materials, FeCrAl for cladding and uranium silicide (U{sub 3}Si{sub 2}) for fuel. Then, we present the approach for deciding which simulations are necessary given the importance of such simulations in providing a detailed report into the fuel performance behavior of these materials by the conclusion of the HIP. In absence of experimental data and given the aggressive schedule outlined by the AFC for ATF research and development, computational analysis tools are required. These tools are being used to help guide experiment design and to perform bounding and sensitivity studies to gain valuable information for which experimental data is limited. In support of this effort, models have been incorporated into BISON for simulation of U{sub 3}Si{sub 2} and FeCrAl. Continued material model development and engineering analysis will help address questions related to the performance of these ATF materials under normal operating and accident conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Conceptual design study of small long-life PWR based on thorium cycle fuel

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Modeling Approach for the Aluminum-clad Dry Storage Pilot using HFIR Fuel

To confirm that the dry storage of aluminum-clad research reactor spent nuclear fuel (ASNF) will remain within the safety envelope after applied drying schemes and that the resulting evolution of the gas space composition, temperature, and pressure conditions are understood, a dry storage pilot project is being established. The pilot will incorporate an instrumented lid for discrete interval or for on-demand gas composition and temperature monitoring of two DOE Standard Canisters (DSCs) loaded with three High Flux Isotope Reactor (HFIR) inner cores per DSC. Each DSC would be subjected to a separate alternative candidate drying scheme. Canisters will undergo 1 to 5 years of monitoring, including internal temperature and gas sampling to track pressure and composition changes. This report outlines the approach for modeling the ASNF-in-canister behavior in terms of evolving gas space conditions for the ASNF dry storage pilot using HFIR fuel. The ASNF has an adherent surface oxyhydroxide layer comprised of boehmite/bayerite that generates hydrogen when subjected to irradiation. Three-dimensional multi-physics computational fluid dynamics simulations will be executed to compute the thermal field within the DSC and provide inputs to a chemical model employed to compute pressure buildup as hydrogen is generated in the system. Implemented in Cantera, the chemical model solves gas phase and aluminum oxyhydroxide surface-mediated radiolysis reactions. Gas phase reactions are sourced from Wittman and Hanson (2015), whereas surface-mediated reactions are incorporated by fitting experimental data using an optimization algorithm (Abboud, 2023). Water radiolysis reactions from Wren and Ball (2001) are adopted with modifications as described in Abboud (2023c). Understanding the effect of the hydrogen buildup over time is important for long-term storage safety considerations. Modeling results will include the canister pressure, temperature, and composition evolution from the initial helium backfill with the addition of radiolytically-evolved chemical species (e.g., hydrogen and oxygen). The specific HFIR cores for the pilot program have not yet been selected, and the overall design is still in development. The CFD-chemical model used for this work will be based on prior models with necessary updates to allow for improved accuracy and efficiency. The experimental data obtained from the HFIR demonstration will be used to improve and validate the computational models to predict the ASNF-in-canister behavior.

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

One-way coupled simulation of FIV in a 7-pin wire-wrapped fuel pin bundle

A normal fuel geometry found in sodium cooled reactor designs is the hexagonal lattice of wire-wrapped fuel pins. The wire-wrappers, which replace the more common spacer grids, act as a spacer between the fuel pins, while simultaneously increasing the fluid mixing between subchannels. The induced transverse flow, while good for mixing, increases the force load experienced by the fuel pins, increasing the possibility of problematic flow induced vibrations. The primary source of vibrations is turbulent buffeting, so very small in amplitude, but capable of rupturing the fuel cladding by fretting. The vast majority of research on fretting has been for bare rods with spacer grids. There are several significant ways that wire-wrapped fuel pins differ from bare rods that may have significant impact on the vibrations present. The increased transverse flow and asymmetrical flow pattern may result in larger and imbalanced force loads on the pins. The helical bracing of each pin may result in unique vibration patterns, and gaps may form between pins affecting how they are braced. Simulation allows for the observation of the forces present, how the flow field affects these forces, and how the structure responds and can be repeated for multiple cases for relatively low cost. Here, simulation of the flow induced vibrations for a 7-pin wire-wrapped fuel pin bundle is carried out by coupling Nek5000, a spectral element LES/DNS solver with DIABLO, a finite element CSM solver. The small amplitude vibrations allow for the assumption that the structure motion has negligible effect on the fluid flow, permitting the much less computationally expensive one-direction coupling. In order to gather significant force history data for the 2.63 helical pitch structure, a scheme was developed to artificially create force history from the power spectrum density of the one-pitch CFD simulation forces. The vibrations could then be simulated in the structure multiple times varying the contact and gaps between various pins. This provides for the search of worst-case scenarios and informs about the types of fretting that might be present. The simulation geometry is based on an experiment for comparison. However, the viscosity of the simulation was increased by a factor of four to make the simulation computationally feasible. While validation is not possible, comparisons are still valuable.

42 ENGINEERING↗
DOE OSTI2024

METHOD FOR EXPOSING IRRADIATED FUEL CLADDING TO RAPID THERMAL TRANSIENTS

Reactor core materials exposed to rapid thermal transients during accidents can experience significant changes in mechanical properties as irradiation hardness is recovered. Understanding material performance during transients using computer modeling requires accurate descriptions of mechanical property evolution with time and temperature. A simple experimental capability was developed to expose irradiated Zircaloy to a simulated thermal transient using immersion in molten tin (at 675°C) to rapidly heat and hold before rapidly cooling with a water quench. Multiple heating-cooling cycles using a nonirradiated tensile specimen showed simplicity and viability. Based on preliminary test results, a robust system was established for routine evaluation of simulated thermal accident scenarios in the Low Activation Materials Design and Analysis (LAMDA) laboratory at Oak Ridge National Laboratory (ORNL). This paper summarizes system development and nonirradiated Zircaloy commissioning test results. Future experiments will conduct tensile tests on irradiated Zircaloy after various rapid temperature profiles.

Byun, TS↗
DOE OSTI2023

Interface Problem Formulation Improvements with Application to Nuclear Fuel Performance Analysis

The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation Program aims to develop predictive capabilities by applying computational methods to the analysis and design of advanced reactor and fuel cycle systems. This program has been providing engineering scale support for the development of BISON, a high-fidelity and high-resolution fuel performance tool. This report documents new developments and robustness improvements in mechanical and thermal (gap heat transfer) contact formulations. The improvements range from the migration of industrial level (“assessment”) nuclear fuel model setups to the usage of mortar formulations, the addition of frictional contact to one-dimensional layered representations of fuel and cladding components, and the addition of the Petrov-Galerkin approach to dual mortar, which improves performance on curved, relatively coarse meshes. In addition, the Lagrange-multiplier enforcement of mechanical mortar contact constraints has been extended to two additional types of enforcement: penalty and augmented Lagrange-Uzawa. We show that the latter approach yields the same interface results as dual mortar in the Multiphysics Object-Oriented Simulation Environment, with the advantage of not worsening the condition number of the system matrix—thereby enabling the use of some general implementations of iterative preconditioners, at the expense of additional system evaluations (i.e., augmentations).

42 ENGINEERING↗
DOE OSTI2017

Microstructural Characterization of Irradiated U0.7ZrH1.6 Using Ultrasonic Techniques

In recent years, there has been an increased level of effort to understand the changes in microstructure that occur due to irradiation of nuclear fuel. The primary driver for this increased effort is the potential for designing new fuels that are safer and more reliable, in turn enabling new and improved reactor technologies. Much of the data on microstructural change in irradiated fuels is generated through a host of post irradiation examination techniques such as optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM) to determine grain structure, porosity, crack geometry, etc. in irradiated fuels. Such “traditional” examination techniques were recently used to characterize a novel new fuel consisting of U0.17ZrH1.6 pellets bonded to zircaloy-2 cladded with lead-bismuth eutectic before and after irradiation. However, alternative methods such as ultrasonic inspection can provide an opportunity for nondestructively assessing microstructure in both in-pile and post-irradiation examinations. In this paper, we briefly describe initial results of ultrasonic examination of the U0.17ZrH1.6 pellets (unirradiated and irradiated), in a post-irradiation examination study. Data indicate some correlation with microstructural changes due to irradiation; however, it is not clear what the specific microstructural changes are that are influencing the ultrasonic measurements. Interestingly, specimens with nominally identical burnup show differences in ultrasonic signatures, indicating apparent microstructural differences between these specimens. A summary of the experimental study, preliminary data and findings are presented in this short paper. Additional details of the analysis will be included in the presentation.

Microstructure evaluation↗
DOE OSTI2019

Room Temperature Lasing of Single-Mode Arched-Cavity Quantum-Cascade Lasers

Single-mode lasing at room temperature in quantum-cascade lasers (QCLs) with arched cavity design has been demonstrated. The output optical power in single-mode lasing regime at ~7.7-μm lasing wavelength was above 6 mW with a side-mode suppression ratio of up to 25 dB. The QCL heterostructure for the arched cavities was grown by molecular-beam epitaxy (MBE) based on a heterojunction of In{sub 0.53}Ga{sub 0.47}As/Al{sub 0.48}In{sub 0.52}As solid alloys, lattice-matched with InP substrate, and InP layers performing the function of waveguide claddings.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2014

STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

NUMERICAL VERIFICATION OF THE RELAP-7 CORE CHANNEL SINGLE-PHASE MODEL

The RELAP-7 code is the next generation of nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). All the physics in RELAP-7 are fully coupled and the errors resulted from the traditional operator-splitting approach are eliminated. By using 2nd order methods in both time and space and eliminating operator-splitting errors, the numerical error of RELAP-7 can be minimized. Numerical verification is the process to verify the orders of numerical methods. It is an important part of modern verification and validation process. The core channel component in RELAP-7 is designed to simulate coolant flow as well as the conjugated heat transfer between coolant flow and the fuel rod. A special treatment at fuel centerline to avoid numerical singularity for the cylindrical heat conduction in the continuous finite element mesh is discussed. One steady state test case and one fast power up transient test case are utilized for the verification of the core channel model with single-phase flow. Analytical solution for the fuel pin temperature and figures of merit such as peak clad temperature and peak fuel temperature are used to define numerical errors. These cases prove that the mass and energy are well conserved and 2nd order convergence rates for both time and space are achieved in the core channel model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

An approach for spent nuclear fuel containment integrity verification using gas tagging

Verification of containment integrity is required for spent nuclear fuel (SNF) managed by the commercial nuclear industry and U.S. Department of Energy (DOE), especially after extended storage. Certain SNF storage systems, such as the DOE road-ready dry storage system, hold several packaged containments within a welded over-canister. These packaged containments are called Department of Energy Standard Canisters (DOESCs). DOESC leakage identification is challenging because their containment boundary cannot be accessed for testing and their contents (i.e., SNF and fill gas) are often similar. There are concerns that this could result in costly characterization and repackaging operations of DOE road-ready dry storage systems if compromised DOESCs are suspected. Here, to address these concerns, this paper presents an approach for applying a gas tagging process using xenon to uniquely identify compromised inaccessible containments following extended storage. The containments considered for this application are seven DOESCs, each packaged within a single over-canister. Two different SNF loading configurations from the Advanced Test Reactor and Fort Saint Vrain nuclear power plant are considered. These configurations are used to represent research reactor aluminum-clad spent nuclear fuel (ASNF) and TRi-structural ISOtropic (TRISO) SNF types. Results for this application show that for ASNF and TRISO type fuels for which the selected fuels are representative, the volume of taggant required at loading is determined primarily by the lower detection limit and leak rate of taggant from a compromised DOESC, rather than the amount of fission-generated xenon in the loaded fuel. While the application presented is suited for larger leaks, smaller leaks could be detected by modifying certain design parameters. This gas tagging approach can also be applied to other DOE containments and advanced reactor SNF storage systems.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2015

Design of water-moderated heterogeneous cores in new STACY facility through JAEA/IRSN collaboration - 14363

In the post-Fukushima prospect, Japan Atomic Energy Agency, JAEA, intends to re-open its Static Experiment Critical Facility STACY, in order to contribute to the validation of fuel debris calculations in conditions close to those affected by the Fukushima Daiichi Nuclear Power Station accident. The reactor core of STACY is being modified by JAEA from homogeneous solution fuel core to heterogeneous pin- type fuel core. The first step of critical experiments is to establish reference experiments involving lattices of UO{sub 2} rods without fuel debris. A design of such experiments is in progress for the widest range of moderation ratios in the frame of JAEA/IRSN collaboration. To anticipate the specification values of materials to be manufactured (grids, rods), an evaluation of uncertainties pertaining to the different parameters driving k{sub eff} is paramount. The following uncertainties of experimental cores are evaluated to estimate the reactivity effect of engineering tolerance/or measurements of the modified STACY: (1) Density of fuel pellets, (2) Enrichment of uranium, (3) Diameter of fuel pellets, (4) Thickness of fuel claddings, (5) Impurities of fuel pellets, (6) Impurities of fuel claddings and (7) Positioning of fuel rods in the lattice. For these calculations, three types of moderation conditions were considered within the experimental parameters of STACY: the lowest moderation (lattice pitch = 1.15 cm; V{sub m}/V{sub f}=1.2), optimum moderation (lattice pitch = 1.5 cm, V{sub m}/V{sub f} = 2.9) and an over moderated case (lattice pitch = 2.55 cm; V{sub m}/V{sub f} = 11). The calculations to derive the overall uncertainty of modified STACY were performed propagating the 1σ uncertainties of parameters of existing STACY in terms of k{sub eff}. For that purpose, a perturbation calculation (differential operator) technique and a stochastic geometry modeling technique were used with the MCNP5 continuous energy Monte Carlo code and a point-wise cross section library based on the JENDL-4.0 evaluated nuclear data library. Moreover, the results of calculations were compared with those calculated by two continuous energy Monte Carlo codes using the correlated sampling technique: - An experimental version of the Japanese MVP2 code combined to the JENDL-4.0 nuclear data library, - The MORET 5 code, developed by IRSN, combined to the JEFF-3.1 nuclear data library. The proposed paper discusses all the aforementioned issues, and confirms that the uncertainty pertaining to the critical experiments of the STACY is sufficiently small. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Preliminary HEA Solid Phase Processing Development Report

The development of high temperature fuel cladding materials to withstand a variety of extreme environments have received much attention. Several potential materials systems that have been identified for the fuel systems and core structural materials application in advanced reactor systems are ferritic/martensitic steel (e.g., HT9), austenitic stainless steels (e.g., 316 LN), oxide-dispersion strengthened steels (e.g., 12 YWT), Ni-based alloys and ceramic-based composites depending on the type of the reactors. Though these material systems have promising properties conducive for radiation-resistant performance, they suffer beyond the design-limit from one or more damage processes such as void swelling, radiation embrittlement, phase instability, corrosion, and limited creep life.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Microstructural Characterization of Irradiated U{sub 0.17}ZrH{sub 1.6} Using Ultrasonic Techniques

In recent years, there has been an increased level of effort to understand the changes in microstructure that occur due to irradiation of nuclear fuel. The primary driver for this increased effort is the potential for designing new fuels that are safer and more reliable, in turn enabling new and improved reactor technologies. Much of the data on microstructural change in irradiated fuels is generated through a host of post irradiation examination techniques such as optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM) to determine grain structure, porosity, crack geometry, etc. in irradiated fuels. Such 'traditional' examination techniques were recently used to characterize a novel new fuel consisting of U{sub 0.17}ZrH{sub 1.6} pellets bonded to zircaloy-2 cladded with lead-bismuth eutectic before and after irradiation. In addition to the conventional suite of post-irradiation examination tools, several nondestructive interrogation techniques exist which bring the potential for in-situ assessment of microstructural changes. One such method, ultrasonic examination, was tested on each of the samples. This paper describes the results of this examination. Ultrasonic measurements have a long and successful history of use for nondestructive materials characterization, including nondestructive evaluation (NDE) of degradation and damage, microstructure characterization, and measurement of physical parameters for process control. Ultrasonic NDE is critical to the nuclear power industry's in-service inspection program for maintaining pressure boundary integrity, and is being actively investigated for PIE of fuels. Ultrasonic measurements, typically performed at 10 MHz or higher, performed post irradiation show that fuel microstructural parameters, such as porosity and grain size, can be correlated to ultrasonic velocity. The objective of this examination was to assess the feasibility of using ultrasonic measurements on irradiated fuel to determine sensitivity to microstructural changes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Post Quench Ductility Evaluation of Zircaloy-4 and Select Iron Alloys under Design Basis and Extended LOCA Conditions

Oxidation experiments were conducted at 1200 C in flowing steam with tubing specimens of Zircaloy-4, 317, 347 stainless steels, and the commercial FeCrAl alloy APMT. The purpose was to determine the oxidation behavior and post quench ductility of these alloys under postulated loss-of-coolant accident conditions. The parabolic rate constant for Zircaloy-4 tubing samples at 1200 were determined to be k = 2.173 107 g2/cm4/s C, in excellent agreement with the Cathcart-Pawel correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. The ductility of post quenched samples was evaluated by ring compression tests at 135 C. For Zircaloy-4, the ductile to brittle transition occurs at an equivalent cladding reacted (ECR) of 19.3%. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR 50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours.

accident test↗
DOE OSTI2023

Addendum to Capability Needs for Irradiated and Radioactive Materials Research (Ad Hoc Committee Summary Report)

Nuclear materials and fuels studies are challenging research and engineering targets because of their inherent radioactivity as well as their heterogeneous microstructure. Advanced light sources have made significant underpinning scientific contributions to the understanding of structural and fuel cladding material microstructure and properties and the development of structure-property relationships. These facilities have the potential to help advance the Department of Energy (DOE) Office of Nuclear Energy (DOE-NE) mission priorities providing underpinning understanding of technically important challenges, including: irradiation induced embrittlement and swelling; stress corrosion cracking and corrosion in extreme environments; ageing of reactor pressure vessels steels; nuclear fuel characterization; and nuclear waste form optimization as well as playing a major role in the development of a materials performance matrix to aid in the design of new materials to meet the needs of advanced reactor concepts. However, the delivery of underpinning understanding is the mission of DOE Office of Science (DOE-SC). While filling such knowledge gaps is important, efforts in this direction should not be confused with the strategic technology focused mission goals of DOE-NE to enable the continued operation of existing U.S. nuclear reactors, to enable the deployment of advanced nuclear reactors, to develop advanced nuclear fuel cycles, and to maintain U.S. leadership in nuclear energy technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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↗