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DOE OSTI2024

Distributed Magnetic Field and Temperature Monitoring for Superconducting Radio Frequency Cavities

The overall objective of the proposed Phase I program was to design, construct and demonstrate a fiber optic sensing system capable of providing temperature and magnetic field measurements with an enhanced spatial resolution that can be implemented over a large surface area (cryomodules) to survey superconducting radio frequency cavities and magnets. A magnetic field sensor capable of detecting fluxes on the order of 1 μT is required to detect the distribution of trapped flux on the cavity surface. A unique distributed magnetic field sensor was successfully designed and constructed to demonstrate the detection of magnetic fluxes less than 500 nT. The sensor leveraged the ultra-high sensitivity of Sentek’s picoDAS to measure the magnetostriction induced vibrations in a commercially available Metglas 2605 SC ribbon that was in physical contact with sensing fiber. Static magnetic fields were detected by applying an alternating current a copper wire proximate to the Metglas 2605SC ribbon to create an AC bias magnetic field. In an alternative approach, an AC bias magnetic field was applied to a special magnetic field sensing fiber with Metglas 2605SC cladding successfully detect a magnetic field flux of a 3 μT. Exhaustive testing was performed to characterize the dependency of sensor response on the direction of the applied magnetic field. Although the special sensing fiber based magnetic field sensor did not exhibit an observable dependence on the direction of the magnetic field, the Metglas 2605SC ribbon sensor exhibited a clear directional dependence. A wide variety of polymer materials were evaluated to enhance the temperature response of an FBG based sensor at cryogenic temperatures. The processing and performance challenges provided the motivation to develop a new simple cryogenic temperature sensor that uses a commercially available fiber optic splice protector. The EVA hot melt tube that becomes adhered to the optical fiber and the polyolefin outer tube that shrinks upon heating in the fusion splicer heater provide the high thermal expansion coefficient necessary to impart a significant strain on the FBG when exposed to cryogenic temperatures. The temperature sensitivity (Δ𝜆𝐵𝑟𝑎𝑔𝑔~ 62 𝑝𝑝𝑚/℃) of the FBG-based sensor was on par with the best reported to date. The simple design, use of readily available cost-effective materials, and well-established processing techniques lends this approach to the creation of hundreds to thousands of temperature sensors on one single optical fiber length. The inherently small form factor also allows for co-location with the distributed magnetic field sensor. In preparation for field testing of the prototype sensing system at the Jefferson Labs in potential Phase II program, several different cable designs were evaluated to package the sensors. The preliminary successful demonstration of fully functional sensing cables provides the foundation for subsequent development efforts to advance the Technology Readiness Level of the technology. The technical feasibility of the proposed approach was successfully demonstrated in this Phase I effort.

43 PARTICLE ACCELERATORS↗

DOE OSTI2014

Micro-Structured Sapphire Fiber Sensors for Simultaneous Measurements of High-T and Dynamic Gas Pressure in Harsh Environments

This is the final report for the program “Micro-Structured Sapphire Fiber Sensors for Simultaneous Measurements of High Temperature and Dynamic Gas Pressure in Harsh Environments”, funded by NETL, and performed by Missouri University of Science and Technology, Clemson University and University of Cincinnati from October 1, 2009 to September 30, 2014. Securing a sustainable energy economy by developing affordable and clean energy from coal and other fossil fuels is a central element to the mission of The U.S. Department of Energy’s (DOE) National Energy Technology Laboratory (NETL). To further this mission, NETL funds research and development of novel sensor technologies that can function under the extreme operating conditions often found in advanced power systems. The main objective of this research program is to conduct fundamental and applied research that will lead to successful development and demonstration of robust, multiplexed, microstructured silica and single-crystal sapphire fiber sensors to be deployed into the hot zones of advanced power and fuel systems for simultaneous measurements of high temperature and gas pressure. The specific objectives of this research program include: 1) Design, fabrication and demonstration of multiplexed, robust silica and sapphire fiber temperature and dynamic gas pressure sensors that can survive and maintain fully operational in high-temperature harsh environments. 2) Development and demonstration of a novel method to demodulate the multiplexed interferograms for simultaneous measurements of temperature and gas pressure in harsh environments. 3) Development and demonstration of novel sapphire fiber cladding and low numerical aperture (NA) excitation techniques to assure high signal integrity and sensor robustness.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2016

Experimental test plan for PWR sister rods in the high burnup spent fuel data project

The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally defining the effects of long-term storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO, and M5. In approximately 10 years, the instrumented cask will be opened for inspection and several fuel rods from the project assemblies will be characterized. To support the project goals, the project pulled 25 'sister rods', 9 from the project assemblies and 16 from similar HBU assemblies, for characterization. The 25 sister rods are all high burnup and cover the range of cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed nondestructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of the storage period (the 'cask rods') to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup used fuel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

MP197HB; First licensed transportation package with canistered high burnup fuel after dry storage - 15580

AREVA TN has recently received the NRC 10 CFR Part 71 transportation license CoC 9302 Revision 7 for the MP197HB transportation package that allows transportation of canistered high burnup fuel assemblies. The Nuclear Horizontal Modular Storage (NUHOMS{sup R}) system canisters containing high burnup PWR and BWR fuel assemblies are now licensed for transportation as a payload in MP197HB after these canisters were in dry storage at the utility ISFSIs. Enhancements have been made to the MP197HB transportation package to improve thermal performance and also provide superior shielding capabilities. Design features of existing AREVA TN transportation packages such as radial fins, thermal shield, neutron resin and optimized gamma and neutron shielding were used to accommodate transportation of high burnup fuel assemblies with heat loads up to 32 kW. The transportation package is designed for exclusive use by rail, truck or marine transport. State-of-the-art analyses methodologies were used to demonstrate compliance with regulatory requirements of 10 CFR 71.55 for the high burnup fuel after dry storage at any ISFSI. Some of the issues addressed in the license application related to the transportation of high burnup fuel included potential degradation of fuel cladding, uncertainty associated with high burnup fuel cladding material properties, and larger radiation and thermal sources due to higher burnups. Bounding conservative assumptions were made where limited test data was available. AREVA TN's licensing approach did not rely on the cladding integrity of high burnup fuel cladding. Instead, the criticality, thermal and shielding analyses were carried out assuming high burnup fuel reconfiguration. With the approval of this transportation package for PWR and BWR canistered high burnup fuel from dry storage at ISFSIs, the utilities can now transport these canisters which are in dry storage at their sites. This is also a major step forward for the canistered fuel that is currently in dry storage at the shutdown plants because they can use this AREVA TN MP197HB transportation package to ship that fuel to any interim storage facility (ISF) or to a repository when they become available. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Vitrification of Sellafield Wastes with High Metallic Magnesium Alloy Content - 17495

The Sellafield nuclear complex, located in northwest England, is the largest such site in the United Kingdom. Sellafield Limited, which is the operating contractor for the Nuclear Decommissioning Authority, is evaluating thermal treatment technologies for a potential future facility to process a variety of Intermediate Level Waste sludges and slurries accumulated from historical site operations. In support of that effort, Atkins and the Vitreous State Laboratory (VSL) of The Catholic University of America conducted studies to evaluate the feasibility and implications of installing a joule-heated ceramic melter (JHCM) vitrification system to process a variety of sludge and slurry radioactive wastes at the Sellafield site. These sludges have the potential to contain a significant proportion of reactive metals in localized regions. The metallic fraction derives from unreacted Magnox spent fuel cladding. Thus, an important objective of this work was to investigate the performance of JHCM technology for the treatment of streams with localized regions of high reactive metal contents. The present work was focused on Site Ion Exchange Effluent Plant (SIXEP) Magnox sludges, investigated previously. Suitable waste simulants were developed for the SIXEP Magnox sludge waste stream, which included surrogates for all important radionuclides. High waste-loading, fully compliant glass formulations were designed using a combination of previous experience, extensive waste glass data bases, and small-scale crucible melts. Glasses from the test melts were characterized for properties that are important for processability and product quality. The most viable compositions were then selected for small-scale melter trials to collect information on processing rates and behavior, off-gas and secondary waste compositions, and overall system mass balance. These data were used to support the technology evaluation, which considered such factors as minimization of inactive additions, waste volume reduction, retention of radionuclides, minimization of secondary wastes, and overall process efficiency and effectiveness. The data were also used to develop sizing estimates for the full-scale facility and to estimate treatment durations under various treatment scenarios. A series of tests was then performed with partial replacement of the magnesium hydroxide with a metallic magnesium alloy that is close in composition to Magnox alloy. A commercially available magnesium alloy, Elecktron AZ31B, which has a nominal magnesium content of 95.7%, was used for this purpose. The testing was performed on a DuraMelter 10 (DM10) JHCM system to characterize the melter off-gas, glass product, and the effects of magnesium alloy particle size on feed processing. Data on processing rate and any processing issues, product glass composition, off-gas composition including hydrogen, and system operating parameters, including flows, temperatures, pressures, etc., were collected and analyzed to evaluate the effects of magnesium alloy on the vitrification process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Instrumented Lid: Full Instrumentation System Testing

This report describes full system testing of an instrumentation system for the monitoring of aluminum clad spent nuclear fuel (ASNF). The requirements for this system were laid out in a previous document, with respect to former studies, and the details of project actualization were reported. The reader is directed to these references for more background information. In short, a team composed of personnel from Savannah River National Laboratory (SRNL) and Idaho National Laboratory have been tasked with the design and fabrication of an instrumented assembly for remote sampling, monitoring, and radiation testing (SMART) for demonstration and surveillance of the dry storage environment and conditions of ASNF in a DOE standardized canister. SRNL proposed the fabrication of two systems to accomplish this goal: 1) an instrumentation system consisting of gas and particle collection, temperature, humidity, pressure, and radiation monitoring; and 2) a lid assembly consisting of a custom-designed lid and hat, together with a custom fuel basket. The corresponding instrumentation system is shown in the piping and instrumentation diagram. The individual functionality tests of the instrument system sensor components were performed, and the results were reported in the report “Instrumented Lid- System Functionality Testing.” Savannah River National Laboratory, September 1, 2019.7 The sensor components of instrumentation system have been mounted and tested in-line with a 6” diameter test canister, and the results are reported in this document.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Benchmark Exercise for the Control Rod Swelling Evaluation

The VTR core has six reactivity control assemblies and three safety assemblies. The control assemblies or primary control rods are adjusted during the normal operation to balance the core reactivity and to control the reactor power. A typical control assembly radial layout is presented in Figure 1. The figure shows the swelled absorber (B 4 C) rod. Initially, helium gas fills the gap between the pin and the cladding before irradiation swelling takes place. For VTR, HT9 steel was selected as the cladding and duct material. The main neutron absorbing material used in the VTR is B 4 C. When residing in the core, the neutronics, thermophysical, and mechanical properties of the materials used in a control assembly will degrade due to accumulated neutron damage. Material degradation limits how long a control assembly can reside in the core. Many phenomena affect the control assembly lifetime, such as the loss of reactivity worth due to B 4 C depletion, the mechanical interaction of the absorber rod and the cladding due to B 4 C swelling, the helium gas buildup in the pin due to B-10 capture, etc. B 4 C swelling, which causes closure of the gap between the absorber rod and the cladding, is usually considered as the main limiting factor from past experience. An initial study was conducted at PNNL to evaluate the irradiation behavior of a VTR control assembly. The evaluation was performed using the CNRD2 code that was initially developed for the FFTF. The study also included an assessment of the VTR control assembly and focused on a 61-pin control assembly design, which is different from that used (37-pin design) in the core design study. The study conducted by PNNL was reviewed independently by ANL. A Python script referred to as the Control Assembly Evaluation Script (CAES) was developed for the independent review and additional assessment of 37-pin control assembly design. The script has focused on the assessment of the absorber rod swelling for its importance in determining the control assembly lifetime. CAES uses geometry, neutronics, materials data as input to predict the swelling of the absorber rod during its residence in the reactor core. The results from CAES showed some non-negligible differences against the PNNL results. Some of the differences can be attributed to the different interpretation of the control rod assembly dimensions. To resolve this issue, a benchmark exercise was proposed. The benchmark specification was developed by PNNL. The benchmark exercise was performed independently at PNNL and ANL using different codes/scripts (CRND2 and CAES). This memo documents the results calculated using the different codes. However, this report is limited to presenting the results obtained. Further investigation of the cause of the observed difference will be performed as part of future activities, pending continuation of the VTR program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Scalable measurements of tow architecture variability in braided ceramic composite tubes

Abstract The performance of braided ceramic matrix composites has been shown to depend on the spatial arrangement of tows; therefore, a new class of tools is required to measure irregularities in the composite architecture for components with intricate geometries. We report a scalable and robust reconstruction technique built upon stereoscopic digital image correlation that is able to efficiently measure the position of tows in arbitrarily shaped composites. This method was applied to triaxially braided ceramic matrix composite tubes intended for use as nuclear fuel cladding, which revealed both long‐range “systematic” tow packing defects associated with the manufacturing process and short‐range “intrinsic” defects due to the braid architecture. These findings suggested that the character of tow spacing variation in braided composite tubes was substantially more complex than in planar woven composites. These measurements are expected to lead to improved processing of braided composites and to facilitate the design of statistically representative virtual specimens for finite element modeling.

Heim, Frederick M.↗

DOE OSTI2020

Monitoring the Caustic Dissolution of Aluminum Alloy in a Radiochemical Hot Cell Using Raman Spectroscopy

Chemical processing of highly radioactive materials commonly takes place in heavily shielded hot cells. The remote, real-time monitoring of chemical processing streams via optical spectroscopic techniques in hot cells may be particularly useful. Here in this paper, we describe the implementation of Raman spectroscopy and chemometric analysis to monitor the dissolution of aluminum-clad targets containing irradiated aluminum–neptunium oxide cermet pellets in caustic solutions in a hot cell environment. Partial least squares regression analysis was used to generate calibration models to quantify the concentration of dissolved aluminum, nitrate, and hydroxide in solutions within the radiochemical hot cell. This work explored a systematic approach to optimize a matrix of calibration standards using a D-optimal experimental design. The Design of Experiments-based regression model, in comparison to more traditional analytical approaches, was found to be the more practical method for building calibration models, with fewer samples, to obtain informative analytical data from Raman spectra.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Electrically-Assisted Tubing Processes for Enhancing Manufacturability of Oxide Dispersion Strengthened Structural Materials for Nuclear Reactor Applications

The overall goal of this project is to enhance the formability of Nanostructured Ferritic Alloys (NFAs) such that high-strength thin-walled claddings can be successfully and economically manufactured for use in future nuclear reactors. Specifically, four objectives were identified: Objective #1 – to achieve a good understanding of the mechanical behavior and resulting microstructure of 14YWT made from the powder form subjected to various Electrically- Assisted (EA) loading conditions. Objective #2 – to conduct EA-flat rolling of 14YWT completed and the prioritization/selection of two alternative designs of EA tube rolling processes. Objective #3 – to design and assemble a new EA tube rolling machine. Objective #4 – to conduct rolling using the new EA rolling machine. The above objectives were achieved and the findings can be further applied to establish other innovative fabrication processes. Specifically, the major accomplishments are summarized as: The use of an EA forming process was investigated to reduce the forming force of ODS ferritic steels. The effect of a continuous electric current on 14YWT during plastic deformation was evaluated. EA uniaxial tension tests show that the electric current is able to reduce the flow stress of 14YWT. 14YWT exhibits a ductility minimum at a current density of 34.5 A/mm2 which is consistent with fractography results. Hardness measurements and TEM results indicate that oxide particles are stable during the EA process. By following the tests, microstructure analysis was performed on the sample in its as-received state and after current treatment. The microstructures are stable when subjected to high current density. The related thermal model was developed to predict the Joule heating temperatures. The temperatures were compared with the experimental results. At last, a temperature-dependent hardening model was used to explore the mechanisms acting during EA deformation. Since the model can accurately predict the yield strength of 14YWT under EA conditions without the incorporation of non-thermal effects, it can be concluded that the reduction in yield strength is due to the Joule heating generated under continuous current loading conditions. An oven-heated pure thermal tension test was conducted to confirm this conclusion. The new EA tube rolling machine has been completely developed. The related work includes: (1) forming process design; (2) analysis of the effect of roll-tube contact geometry on rolling force; (3) analysis of machine performance under mechanical loading; (4) design of machine electrical setting path and insulation; (5) multi-pass process analysis; (6) design of power transmission system (axial/radial/torque feeding mechanisms), roll bearing system, cooling system, workpiece feeding mechanism, and other related mechanical connection support components; (7) design and optimization of roll profiles/dimensions; (8) analysis and simulation of the rolling process; (9) development of engineering drawings and fabrication; (10) assembly and calibration of the EA tube rolling machine; (11) design and integration of the electrical control system; (12) hardware integration for high power supply electrically-assisted rolling process; (13) induction heating system integration; (14) experimental tests for the new rolling process. Through the process from conceptual design to final machine calibration and experimental tests, there are many technical and physical works that have been implemented and improved for the best performance of the innovative prototype.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Report on High Energy Arcing Fault Experiments

This report documents an experimental program designed to investigate High Energy Arcing Fault (HEAF) phenomena for medium voltage electrical switchgear containing aluminum conductors. This report covers full-scale laboratory experiments using representative nuclear power plant (NPP) three-phase electrical equipment. Electrical, thermal, and pressure data were recorded for each experiment and documented in this report. This report covers four of the fourteen planned medium voltage electrical enclosure experiments. Subsequent reports will document the additional experiments performed in the future. The experiments were performed at KEMA Labs located in Chalfont, Pennsylvania. The experimental design, setup, and execution were completed by staff from the United States Nuclear Regulatory Commission (NRC), the National Institute of Standards and Technology (NIST), Sandia National Laboratories (SNL) and KEMA. In addition, representatives from the Electric Power Research Institute (EPRI) observed some of the experimental setups and execution. The HEAF experiments were performed on four near-identical units of General Electric metal-clad medium voltage switchgear. The three-phase arcing fault was initiated on the primary cable connection bus. All four experiments used the same system voltage (6.9 kV) but varied the current and duration. Real-time electrical operating conditions, including voltage, current and frequency, were measured during the experiments. Heat fluxes and incident energies were measured with plate thermometers and slug calorimeters at various locations around the electrical enclosures. Internal enclosure pressures were measured during the experiments. The experiments were documented with normal and high-speed videography, infrared imaging, and photography. Insights from the experimental series included timing information related to enclosure breach, event progression, mass loss measurements for electrodes and steel enclosures, peak pressure rise, particle analysis, along with visual and thermal imaging data to better understand and characterize the hazard. These results will be used in subsequent efforts to advance the state of knowledge related to HEAF.

42 ENGINEERING↗

DOE OSTI2018

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

CYRANO3 the EDF fuel code performance: Global overview and recent developments on MOX fuel

CYRANO3 is the thermal mechanical industrial code developed and used by EDF to simulate the nuclear fuel rod performance in pile, and during the disposal storage period as well. This code has already been used successfully for the last ten years by the EDF engineering teams in justifying the overall reloading options covering various fuels: UO{sub 2}, UO{sub 2} + gadolinium, MOX, UO{sub 2} plus additives, and various cladding as well proposed by the nuclear fuel suppliers. Its particularity is related to a construction around a standard one-dimensional thermal and mechanical finite elements kernel with an appropriate architecture to plug in any new functionalities or models. Since the beginning it is developed as a module able integrating within a core calculation chain. Other constrain is to always guaranty a running time consistent with the calculation of a core within a reasonable time. The objective is to provide the EDF designers with a convivial and qualified code, with robust algorithms and able to quickly evolve with computer performances. Algorithms are able to account for adiabatic transients and are robust enough to avoid any undesirable overflows. The permanent challenge is indeed to provide with the more descriptive simulations not degrading to much the running time. Developments are obviously performed within the strict Quality Insurance rules. The code is then associated to a large documentation base, including model specification, realization, validation and test procedures. Maintenance is managed through a system of anomaly and improvement forms. Computers are still evolving quickly. Therefore, after an architecture adaptation, the CYRANO3 code is now integrated in the PLEIADES platform as a PWR industrial application. The platform is an exchange space allowing sharing functionalities, models and properties. In this paper, we will give a global overview of the capabilities of CYRANO3 code to simulate the thermomechanical behavior of fuel rods, and we will focus on recent developments realized on MOX fuel. In CYRANO3 code, heterogeneous models (neutronic model, micromechanical behavior law and fission gas model) are developed for MOX fuels and integrated in order to account for differential evolutions between phases under irradiation. The methodology and these models are presented in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Overview of a Versatile Loading System for Anisotropic Material Property Characterization

Additive manufacturing, cold rolling, and other thermomechanical treatments on metals, especially HCP, can cause texture and oriented grain structures resulting in anisotropic mechanical properties. This can lead to macroscopic response that significantly deviates from isotropic assumptions which motivates studying mechanical properties in multiple directions. In addition, multiaxial loading and complex stress states overlapping the above mentioned material anisotropy can lead to unexpected outcomes. Idaho National Laboratory developed an advanced mechanical testing system to study a range of uniaxial to multiaxial stress states while measuring the anisotropic response bringing insight into the overlap between material properties and stress states. High-temperature capability and stress- or strain-controlled loading is available to enable a variety of experiment types and conditions to measure elastic, plastic, and viscoplastic properties. This poster presents the design and capabilities of this system with preliminary results highlighting the benefits of multiaxial loading and anisotropic analysis in an integrated system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

An approach to model reactor core nodalization for deterministic safety analysis

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Designed porosity materials in nuclear reactor components

A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗