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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to provide hermetic seal. The replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Analysis of the absorbed hydrogen in cladding tubes applied in the QUENCH-LOCA tests

A shortened loss of coolant scenario was simulated experimentally at fuel rod bundle scale in the QUENCH-LO test. All pressurized rods burst at temperatures between 790 and 855 deg. C. The hydrogen distribution in the cladding of chosen fuel rod simulators was studied by neutron radiography and tomography. It was shown that the time between burst and quenching as well as the temperatures occurring during this period, and hence the degree of inner oxidation, are important parameters for the hydrogen uptake. No hydrogen uptake was measured for the specimen with 63 s or less between burst and quenching. For longer time intervals, hydrogen enriched bands were formed. These bended bands are oriented non-symmetrically to the tube axis. Micro-structure investigations give hints that the hydrogen keeps in solution in the a-zirconium lattice. The hydrogen enriched bands influence the mechanical properties of the rods, characterized by a higher hardness. Brittle fracture occurs in these bands in the tensile test. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. In this paper, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8–1.9 dpa in a temperature range of 195–559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Finally, both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Systematic Technology Evaluation Program for SiC/SiC Composite-based Accident-Tolerant LWR Fuel Cladding and Core Structures: Revision 2015

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

In Situ Sample Environment for Studying Corrosion of Nuclear Materials

Currently there is lack of fundamental understanding of kinetics and reaction mechanisms that occur at nuclear cladding-steam interface under extreme environments in the nuclear reactors. Such understanding is vital to design and develop advanced cladding materials for more economical and safer nuclear power. We present advanced in situ sample environment for probing solid-fluid interfaces using high resolution synchrotron methods for studying early stages of corrosion including development of oxide layer as corrosion proceeds. Both X-ray diffraction and X-ray fluorescence methods will be discussed to probe structural and elemental composition changes under steam environment using corrosion resistant cell at high temperatures. High intensity and brightness of synchrotron coupled with advanced in situ sample environment developed, presents an opportunity to understand the fundamental interfacial processes involved in degradation of materials. The in situ sample environment is optimized for XRD and XRF data collection at NSLS-II and will be available to all users through the proposal process. The in situ sample environment is built to study the interfacial interactions at temperatures and pressures up to 400 deg. C and 1500 psi. It consists of three main parts; the distribution manifold, the pressure vessel, and the reaction cell as shown in Figure 1. Water is allowed to heat up to 350 deg. C inside the pressure vessel to produce steam which is delivered to reaction cell using insulated delivery system. The cell design is adapted from J. Diefenbacher, et. al. However, it is modified for corrosion applications in steam environments. The reaction cell and sample holder are made from hastelloy due to its high corrosion resistance. The reaction cell is surrounded by copper heating block to control its temperature independently from the rest of the system. Lab view based platform is developed for remote monitoring of pressure and temperature. The sample environment was commissioned at 6-IDD beamline in the Advanced Photon Source, for grazing incidence XRD corrosion studies. Metallic Zr sample (200 μm thickness) was used due to its faster oxidation kinetics as compared to Zircaloy. Sample was exposed to steam at 350 deg. C and 50 psi for a total time of 64 hours. Results showed the presence of α-Zr as the dominant phase. After 32 hours of exposure to steam, three oxide peaks in the low diffraction angle side started to develop with time. These peaks are attributed to the growth of substoichiometric zirconium oxide phases (Zr{sub 3}O and Zr{sub 3}O{sub 1-x}). (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2019

Comprehensive investigation of the role of Nb on the oxidation kinetics of Zr-Nb alloys

Zirconium oxidation, and the associated hydrogen pickup, has long been considered the life-limiting degradation mechanism of fuel elements in pressurized-water reactor (PWR) systems. Currently, PWRs use Zr-Nb alloys as the cladding material of choice. However, oxidation kinetics for Zr-Nb alloys can vary greatly depending on the Nb content; moreover, even Zr-Nb with the same composition can exhibit radically different oxidation kinetics if heat treated at different temperatures. To improve the fundamental understanding of the corrosion of Zr-Nb, it is necessary to determine mechanistically how Nb affects the oxidation kinetics of the overall system. To achieve this, nano-beam X-ray absorption near-edge spectroscopy (XANES), energy-dispersive spectroscopy (EDS), transmission electron microscopy (TEM), and modelling using the physically based C4 model have been performed in concert on samples of Zr-1.0Nb annealed at two different temperatures, 580 °C and 720 °C, and oxidized in an autoclave environment. Through TEM/EDS, it has been seen that β-Zr precipitates dissolved upon oxidation, releasing Nb to the surrounding environment, while β-Nb precipitates remain intact and retard oxidation of Nb. Comparison of the XANES results and C4 model predictions illustrate that space charge generated in the oxide can lead to subparabolic oxidation kinetics in Zr-Nb alloys, which have been shown to negatively affect the hydrogen-pickup fraction. However, sufficient Nb in solid solution can become oxidized to compensate this mechanism to maintain parabolic kinetics in select alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Fast Spectrum Experiment Fuel Pin Payload Specification for Package 9330

This specification supports work to be conducted by AREVA Federal Services under Idaho National Laboratory (INL) Subcontract No. 45559, Release 50, including Statement of Work (SOW) -11417 addressing package transportation licensing. This document provides transportation payload specification for experiment fuel pins produced in the performance of INL CRADA 13-CR-15. The payload description provided in this document will be used to obtain licensing necessary to enable transportation of unirradiated experiments in ATR Fresh Fuel Shipping Containers (FFSCs). The FFSC has a current U. S. Department of Energy (DOE) certificate, USA/9330/AF-96, revision 0, which is available on the DOE RAMPAC website or upon request from the INL or DOE. Its desired payload will consist of individual experiment fuel pins (a.k.a. rods) loaded in a FFSC’s internal fuel handling enclosure (FHE), a.k.a. internal canister. Each package and its FHE arrangements should be prepared for a payload that contains less than critical mass and is certified for ground and air transportation per NRC 10 CFR 71 and International Atomic Energy Agency (IAEA) criteria. The unirradiated experiment fuel pins are all fabricated with uranium-based fuel and ferritic-martensitic stainless steel, HT-9, cladding and end plugs. The payload may include pins with and without spacer-wires. Solid pins will have minimal amounts of sodium bonding between the fuel and inside cladding. Annular pins will contain He gas installed at atmospheric pressure (~12.7 psia). Uranium fuel will be of varying enrichments with a maximum bounding enrichment of 94% U-235. Blanket material consists of depleted uranium and zirconium of varying weight percent. Section 2 provides pin parameters for NRC 10CFR71 and IAEA evaluations. Supplemental criteria titled functional and operating requirements is intended to convey other requirements, including those that protect the payload, rather than to demonstrate package compliance with regulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fuel elements of research reactor CM

In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Severe Accident Analysis Of BWR Core Fueled With UO 2 /FeCrAl with Updated Materials and Melt Properties From Experiments

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. FeCrAl alloys have slower oxidation kinetics and higher strength compared to Zr-based alloys at high temperatures. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident tolerant fuel concept over the existing zirconium based materials employed today.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

First-Principles Studies of Tritium Species Dissociability & Diffusivity Across the Interface of Nickel-Plated Zircaloy-4

Zirconium (Zr) and its alloys (Zircaloy-4) are widely used in nuclear reactors due to their low neutron adsorption cross-section and excellent corrosion resistance. In tritium-producing burnable absorber rods (TPBARs), the metal getter tube located between the cladding and the γ-LiAlO 2 pellets is composed of nickel (Ni)-plated Zircaloy-4, which is used to capture tritium ( 3 H) species (mainly 3 H 2 and 3 H 2 O) generated from γ-LiAlO 2 pellets during irradiation. The 3 H-related products transfer to the surface of metal Ni upon adsorption and dissociation to form new 3 H species and diffuse into the Zircaloy-4 getters to form metal hydrides (Zr 3 H x ). Therefore, exploring 3 H species ( 3 H 2 , 3 H 2 O) dissociation on the surface of Ni and diffusion across the interface of Ni-plated Zircaloy-4 getters can provide a better understanding of 3H species formation and transport from pellets into the getters.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Accommodation of tin in tetragonal ZrO{sub 2}

Atomic scale computer simulations using density functional theory were used to investigate the behaviour of tin in the tetragonal phase oxide layer on Zr-based alloys. The Sn{sub Zr}{sup ×} defect was shown to be dominant across most oxygen partial pressures, with Sn{sub Zr}{sup ″} charge compensated by V{sub O}{sup ••} occurring at partial pressures below 10{sup −31 }atm. Insertion of additional positive charge into the system was shown to significantly increase the critical partial pressure at which Sn{sub Zr}{sup ″} is stable. Recently developed low-Sn nuclear fuel cladding alloys have demonstrated an improved corrosion resistance and a delayed transition compared to Sn-containing alloys, such as Zircaloy-4. The interaction between the positive charge and the tin defect is discussed in the context of alloying additions, such as niobium and their influence on corrosion of cladding alloys.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2014

Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO 2 /zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Assessment of Corrosion Resistance of Candidate Alloys for Accident Tolerant Fuel Cladding under Reactor Conditions

The environment of a light water reactor (LWR) core is a combination of high temperatures, high pressures, high neutron and gamma fluxes, mechanical stresses and chemical aggressive coolants. All of these factors combined can induce changes in the microstructure of the fuel and cladding that are very difficult to predict in a systematic fashion [1] and are major challenge for the safety and life extension of current reactors. Throughout reactor transients and accidents, the cladding may experience deterioration caused by a temperature increase, oxidation embrittlement [1], [2], or mechanical interaction with the fuel caused by stress [3]. These events may lead to cracking or rupture of the cladding, causing the release of fission products into the coolant. Such events have been observed at the Three Mile Island and Fukushima accidents. To avoid such occurrences in LWRs, the accident tolerant fuels (ATF) program was initiated to focus on the replacement of zirconium-based alloys with materials that exhibit slower steam oxidation kinetics. This project focuses on several iron-based alloys such as T91, APMT, MA956, experimental Fe-Cr alloys and one experimental nanofeatured alloy (NFA). Experiments have been conducted in both PWR primary water at 320°C and BWR normal water chemistry at 288°C, spanning a large range in electrochemical corrosion potential (ECP). Samples were exposed to either proton irradiation (University of Michigan) or electron irradiation (Notre Dame Radiation Laboratory) to independently assess the roles of displacement damage or radiolysis on the corrosion rate, oxide thickness, morphology, structure and resistivity. Post-irradiation characterization of various regions of the electronirradiated samples was completed by means of microscopy (Notre Dame Integrated Imaging Facility) and spectroscopy (Notre Dame Materials Characterization Facility) techniques to provide high resolution information regarding the oxide layer present on the surface of the material.

Grdanovska, Slavica↗
DOE OSTI2016

ROD INTERNAL PRESSURE QUANTIFICATION AND DISTRIBUTION ANALYSIS USING FRAPCON

The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified forWatts BarNuclearUnit 1 (WBN1) fuel rods by modeling core cycle design data, intercycle assembly movements, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layers is derived and applied to FRAPCON output data to quantify the RIP and CHS for these fuel rods. SCALE/Polaris is used to quantify fuel rod-specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel blankets. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP limit. Lastly, improvements to the computational methodology of FRAPCON are proposed.

cladding hoop stress↗
DOE OSTI2026

Time at Temperature Experiments on Neutron Irradiated Zircaloy-2 using Conventional & Flash DSC

Dryout events in Boiling Water Reactors (BWRs) are currently treated by NRC regulations as automatic disqualification for continued fuel rod operation, even though this criterion does not account for the rate or duration of power increases, associated changes in material behavior, or the possibility of rewetting. Operational history from Anticipated Operational Occurrences (AOOs) shows that short, transient power excursions often demand only modest heat removal, and industry experience suggests that fuel can briefly enter dryout yet return to safe, stable operation. The lack of detailed understanding of the material response during such events motivates the present series of experiments. This study uses unique and innovative methods to investigate microstructural evolution in irradiated Zircaloy-2 exposed to high temperatures in inert environments. Differential Scanning Calorimetry (DSC) and FlashDSC are combined to build a comprehensive experimental framework capable of identifying the a–ß phase transformation in zirconium and examining defect annealing under steep thermal gradients. FlashDSC enables rapid heating and cooling of Focused Ion Beam (FIB)–prepared large-area lift-outs (LALOs) of irradiated Zircaloy-2 at rates of 1,000 K/s to peak temperatures of 600°C, 750°C, and 900°C. The goal is to determine whether these conditions produce measurable microstructural changes that could influence cladding performance in typical BWR environments. Microstructural characterization includes quantifying dislocation density and assessing secondary phase particle (SPP) size and distribution using Transmission Electron Microscopy (TEM). Ongoing analysis, such as diffraction pattern indexing and 4D STEM processing, will further refine these observations.

36 - MATERIALS SCIENCE↗
DOE OSTI2013

Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO 2 – Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This article describes modifications made to a version of the MELCOR code by the Idaho National Laboratory (INL) that allows its application to alternate fuel-cladding systems. Preliminary analysis results are presented in this paper for candidate silicon carbide and FeCrAl cladding concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗