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At least 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Verification of severe accident management guideline (SAMG) entry condition for OPR1000 - 367

The current severe accident prevention and mitigation strategies for OPR 1000 have emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs), respectively. The main objective of SAMG is to prevent a release of radioactive materials from releasing into environment and mitigate severe accident phenomena during severe accident. The current SAMG entry condition is met when the core exit temperature (CET) reaches 650 deg. C in OPR 1000. Peak cladding temperature (PCT), which represents core integrity more precisely than CET does, cannot be used since it is not measurable in the main control room (MCR). Current SAMG entry condition for every kind of accident scenarios is not reasonable, since entry condition is determined without considering operator action time. Therefore, SAMG entry condition need to be verified with consider operator action time. Thus, an analysis methodology for SAMG entry condition was developed to verify the SAMG entry condition. A severe accident DB were also developed in order to analyze the SAMG entry condition. The analysis results of DB show that available operator action time from the SAMG entrance to RV fail is shorter than the required time in most cases. As a result, the proper SAMG entry condition was suggested through the results of DB analysis. When the SAMG entry condition is changed from CET to suggested condition, the available action time is increased about 350 to 4300 sec. Also, the suggested entry condition was verified through the simulation which considered operation action time and reflecting the suggested entry condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

SU-E-T-186: Feasibility Study of Glass Cherenkov Detector for Prompt Gamma Detection in Proton Therapy

Purpose: To simulate a Cherenkov glass detector system utilizing prompt gamma (PG) technique to quantify range uncertainties in proton radiation therapy. Methods: A simulation of high energy photons typically produced in proton interactions with materials incident onto a block of Cherenkov glass was performed with the Geant4 toolkit. The standard electromagnetic package was used along with several decay modules (G4Decay, G4DecayPhysics, and G4RadioactiveDecayPhysics) and the optical photon components (G4OpticalPhysics). Our setup included a pencil beam consisting of a hundred thousand 6 MeV photons (approximately the deexcitation energy released from 16O) incident onto a 2.5 ⊗ 2.5 ⊗ 1.5 cm3 of a Cherenkov glass (7.2 g of In2O3 + 90 g cladding, density of 2.82 g/cm3, Zeff = 33.7, index of refraction 1.56). The energy deposited from incident 6 MeV photons as well as secondary electrons and resulting optical photons were recorded. Results: The energy deposited by 6 MeV photons in glass material showed several peaks that included the photoelectric, the single and double escape peaks. About 11% of incident photons interacted with glass material to deposit energy. Most of the photons collected were in the region of double escape peak (approximately 4.98 MeV). The secondary electron spectrum produced from incident photons showed a high energy peak located near 6 MeV and a sharp peak located ∼120 keV with a continuous distribution between these two points. The resulting Cherenkov photons produced showed a continuous energy distribution between 2 and 5 eV with a slight increase in yield beginning about 3 eV. The amount of Cherenkov photons produced per interacting incident 6 MeV photon was ∼240.7. Conclusion: This study suggests the viability of utilizing the Cherenkov glass material as a possible prompt gamma photon detection device. Future work will include optimization of the detector system to maximize photon detection efficiency.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Hydrogen pickup mechanism in zirconium alloys - 2016-0055

Because hydrogen ingress into zirconium cladding can cause embrittlement and limit cladding lifetime, hydrogen pickup during corrosion is a critical life-limiting degradation mechanism for nuclear fuel. However, mechanistic knowledge of the oxidation and hydrogen pickup mechanisms is still lacking. In an effort to develop such knowledge, we conducted a comprehensive study that included detailed experiments combined with oxidation modeling. We review this set of results conducted on zirconium alloys herein and articulate them into a unified corrosion theoretical framework. First, the hydrogen pickup fraction (f{sub H}) was accurately measured for a specific set of alloys specially designed to determine the effects of alloying elements, microstructure, and corrosion kinetics on f{sub H}. We observed that f{sub H} was not constant and increased until the kinetic transition and decreased at the transition. F{sub H} depended on the alloy and was lower for niobium-containing alloys. These results led us to hypothesize that hydrogen pickup during corrosion results from the need to balance the charge during the corrosion reaction such that f{sub H} decreases when the rate of electron transport through the protective oxide increases. To assess this hypothesis, two experiments were performed: (1) micro-X-ray absorption near-edge spectroscopy (μ-XANES) to investigate the evolution of the oxidation state of alloying elements when incorporated in the growing oxide and (2) in situ electrochemical impedance spectroscopy (EIS) to measure oxide resistivity as a function of exposure time on different alloys. With the use of these results, we developed an analytical zirconium alloy corrosion model based on the coupling of oxygen vacancies and electron currents. Both modeling and EIS results show that as the oxide electric conductivity decreases the f{sub H} increases. These new results support the general hypothesis of charge balance. The model quantitatively and qualitatively predicts the differences observed in oxidation kinetics and hydrogen pickup fraction between different alloys. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

A hydride reorientation model for irradiated zirconium alloy cladding

This paper presents a hydride reorientation model for irradiated, but de-fueled, zirconium-based alloy cladding tubes for water cooled reactors subjected to temperature and stress conditions of interest to storage of commercial spent nuclear fuel under dry, inert conditions. The present development builds upon and updates an existing hydride precipitation model developed in 2003, and was motivated by recent findings that the old model underestimated the amount of radial hydrides for applied hoop stresses below 100 MPa. The data used to support the development of the new model were obtained from an NFIR-sponsored experimental program on de-fueled cladding tubes irradiated in light-water reactors. The irradiated material consisted of three types of zirconium-based alloys sampled from high burnup fuel rods: cold-worked stress-relieved (CWSR) Zircaloy-4 (Zr-4), CWSR Zirlo{sup TM}, and recrystallized annealed (RXA) Zircaloy-2 (Zr-2). The majority of the hydride re-orientation tests for the irradiated cladding specimens were conducted under constant pressure with a mean cooling rate of about 1 Celsius degree/mn. The remaining tests were conducted at slower cooling rates with the pressure being kept proportional to the temperature change during the cooling process to simulate dry storage conditions. The parameters of the hydride re-orientation model were fitted to the test data. Fitting parameters were determined for (CWSR) Zircaloy-4 (Zr-4) and CWSR Zirlo{sup TM} respectively. The cooling rate effect was empirically accounted for in the model. General good agreement between model prediction and measured radial hydride content is achieved, with a standard error less than 12 ppm. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Status Update on the SCIP-IV Program in 2020

Phase 1 of the Studsvik Cladding Integrity Project (SCIP) was launched in 2004. It was a 5-year Organisation for Economic Co-operation and Development (OECD) / Nuclear Energy Agency (NEA) joint project operated by Studsvik with about 30 participating organizations, including regulatory bodies, research institutions, utilities, and fuel suppliers from 13 different countries. The initial SCIP program focus was directed toward studying basic phenomena of fuel rod failures driven by pellet cladding mechanical interaction to get a better understanding of fundamental failure mechanisms. SCIP IV is a continuation of the SCIP programs and is planned to be another five-year project that started in July 2019. It is organized and managed in a manner similar to the earlier SCIP programs, including both a technical group and a management board. The work of SCIP-IV includes significant research into issues related to the back end of the fuel cycle, including spent nuclear fuel (SNF) storage and transportation issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Initial Application of the Reactivity Method to Analyze KSU TRIGA Fuel

The Kansas State University (KSU) TRIGA Mk II reactor was refueled over 40 years ago with fuel having stainless steel cladding. Much of that fuel had significant burnup, and little work has been done since then to quantify the core composition. The present work aims to provide data in support of efforts to understand this composition. Currently available data consist of 26 critical configurations, including the location of fuel elements, the position of control rods, and various temperatures that have been documented in operational log books. Additional measurements will be made using nondestructive techniques to quantify the burnup of fuel elements irradiated in the KSU reactor. Gamma spectroscopy is a well-established method for determining the burnup of spent fuel, and plans are in place at KSU to perform such measurements. Another option explored here is use of the reactivity method, which relates reactivity worths to burnup using linear or other correlations. Reactivity measurements of fuel elements with wide ranges of burnup were conducted to help identify the present compositions of those fuel elements and to verify the historical inventories of those fuel elements. Significant biases were observed between measured and modeled reactivities for several elements in the KSU reactor. It is expected future measurement of burnups using gamma spectroscopy will help to reduce these biases and produce a best-estimate model of the core composition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Neutron damage reduction in a traveling wave reactor

Traveling wave reactors are envisioned to run on depleted or natural uranium with no need for enrichment or reprocessing, and in a manner which requires little to no operator intervention. If feasible, this type of reactor has significant advantages over conventional nuclear power systems. However, a practical implementation of this concept is challenging as neutron irradiation levels many times greater than those in conventional reactors appear to be required for a fission wave to propagate. Radiation damage to the fuel and cladding materials presents a significant obstacle to a practical design. One possibility for reducing damage is to soften the neutron energy spectrum. Here we show that using a uranium oxide fuel form will allow a shift in the neutron spectrum that can result in at least a three fold decrease in dpa levels for fuel cladding and structural steels within the reactor compared with the dpa levels expected when using a uranium metal fuel. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Effectiveness of a dopant in U-Zr metallic fuel to prevent lanthanide migration

The advanced fast reactor concepts to achieve ultra-high burnup (about 50%) without requiring refueling by way of using metallic alloy fuel have gained interest. Fission product lanthanide accumulation at high burnup is substantial and its migration to cladding and reaction with cladding is a potential life-limiting phenomenon. As a means to solve this problem, adding an element that forms stable compounds with lanthanides to immobilize them has been proposed. The theoretical assessment shows that indium, thallium, gallium, and antimony are good candidates. Except for Sb, because these elements are low-melting temperature elements, liquid metal embrittlement of cladding is a concern if large sized agglomerates exist contacting the cladding. Alloy characterization of as-fabricated samples was performed to examine the effectiveness of the dopant addition method using optical microscopy and scanning electron microscopy. Although preliminary, the present results showed that indium is a better dopant to immobilize lanthanides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bending Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.

cladding failure↗
DOE OSTI2012

Assessment of possible cycle lengths for fully-ceramic micro-encapsulated fuel-based light water reactor concepts

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with accident-tolerant fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rate of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Incorporation of Fines and Noble Metals into HLW Borosilicate Glass: Industrial Responses to a Challenging Issue - 13056

During the early stages of spent fuel reprocessing, the fuel rods are cut and dissolved to separate the solid metallic parts of the rods (cladding and end pieces) from the radioactive nitric acid solution containing uranium, plutonium, minor actinides and fission products (FP). This solution contains small, solid particles produced during the shearing process. These small particles, known as 'fines', are then separated from the liquid by centrifugation. At the La Hague plant in France, the fines solution is transferred to the vitrification facilities to be incorporated into borosilicate glass along with the highly radioactive FP solution. These fines are also composed of Zr, Mo and other noble metals (i.e. Ru, Pd, Rh, etc.) that are added before vitrification to the the FP solution that already contained noble metals. As noble metals has the potential to modify the glass properties (including viscosity, electrical conductivity, etc.) and to be affected by sedimentation inside the melter, their behavior in borosilicate glass has been studied in depth over the years by the AREVA and CEA teams which are now working together in the Joint Vitrification Laboratory (LCV). At La Hague, the R7 vitrification facility started operation in 1989 using induction-heated metallic melter technology and was quickly followed by the T7 vitrification facility in 1992. Incorporating the fines into glass has been a challenge since operation began, and has given rise to several R and D studies resulting in a number of technological enhancements to improve the mixing capability of the melters (multiple bubbling technology and mechanical stirring in the mid-90's). Nowadays, the incorporation of fines into R7T7 glass is well understood and process adaptations are deployed in the La Hague facilities to increase the operating flexibility of the melters. The paper will briefly describe the fines production mechanisms, give details of the resulting fines characteristics, explain how the metallic particles can influence the glass properties and challenge vitrification technologies and finally, give a rundown of the R and D efforts which have produced innovative solutions contributing to the operational excellence of AREVA's vitrification facilities in the La Hague reprocessing plant. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Aluminum cladding oxide growth prediction for high flux research reactors

Aluminum cladding oxidation of research-reactor fuel elements at high power conditions has a disadvantageous effect on fuel performance due to the lower thermal conductivity of the oxide. The oxide growth prediction models available in the literature were mostly developed for low power conditions. To examine the applicability of the models to high power and high temperature test conditions, the models were studied by coupling with the most frequently employed heat transfer coefficient (HTC) correlations including the Dittus-Boelter correlation, the Colburn correlation, the Sieder-Tate correlation, and KAERI-developed correlation. The Griess model over-predicted the oxide growth while the KAERI-Griess model under-predicted the oxide growth for high power tests. The Kim model, coupled with the Colburn correlation, gave most consistent results with the measured data from two BR2 experiments. However, the Kim model was found to be inapplicable to the EUHFRR conditions at the peak power locations if it was coupled with the Dittus-Boelter correlation. A revision of the prediction models to more closely agree with the measured data was recommended. Furthermore, AG3NE and AlFeNi cladding types were tested in the E-FUTURE experiment, and a noticeable (although small) reduction in oxide thickness on the AlFeNi cladding was observed. However, this difference was believed to be only a secondary effect considering other uncertainties in model predictions, so no attempt was made to model the alloying effect.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

BISON Investigation of the Effect of the Fuel- Cladding Contact Irregularities on the Peak Cladding Temperature and FCCI Observed in AFC-3A Rodlet 4

The primary objective of this report is to document results of BISON analyses supporting Fuel Cycle Research and Development (FCRD) activities. Specifically, the present report seeks to provide explanation for the microstructural features observed during post irradiation examination of the helium-bonded annular U-10Zr fuel irradiated during the AFC-3A experiment. Post irradiation examination of the AFC-3A rodlet revealed microstructural features indicative of the fuel-cladding chemical interaction (FCCI) at the fuel-cladding interface. Presence of large voids was also observed in the same locations. BISON analyses were performed to examine stress and temperature profiles and to investigate possible correlation between the voids and FCCI. It was found that presence of the large voids lead to a formation of circumferential temperature gradients in the fuel that may have redirected migrating lanthanides to the locations where fuel and cladding are in contact. Resulting localized increase of lanthanide concentration is expected to accelerate FCCI. The results of this work provide important guidance to the post irradiation examination studies. Specifically, the hypothesis of lanthanides being redirected from the voids to the locations where the fuel and the cladding are in contact should be verified by conducting quantitative electron microscopy or Electron Probe Micro-Analyzer (EPMA). The results also highlight the need for computer models capable of simulating lanthanide diffusion in metallic fuel and establish a basis for validation of such models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Sister Rod Destructive Examinations (FY20) Appendix B: Segmentation, Defueling, Metallograhic Data and Total Cladding Hydrogen

As a part of the US DOE-NE High Burnup Spent Fuel Data Project, ORNL is performing destructive examinations (DEs) of high burnup (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO®, and M5®. The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding are being tested. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities.This report documents the status of the ORNL Phase 1 DE activities related to: rough segmentation, defueling, optical microscopy, and cladding total hydrogen measurements for 7 Phase 1 sister rods and outlines the DE tasks performed and the data collected to date. The results of these detailed examinations and others are summarized in the Sister Rod DE Status Report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Chemistry and Reactivity of Oxide Films on Aluminum Clad Test Reactor Fuels

This paper describes recent work performed to assess the degradation of aluminum spent nuclear fuel (ASNF) as part of an ongoing Department of Energy project. As part of a multi-task effort, this paper specifically describes efforts initiated in 2018 to better understand chemical phenomena and degradation in ASNF). In this paper experimental and modeling activities are described. The modeling effort presents initial results in two areas potentially influencing stability. Using an established electrolyte model based on experimental data, the solubility of aluminum as a function of temperature and pH was examined. This work describes interesting results for the minimum pH providing relevant data that was not available from the literature. It was found that solubility increases with temperature and that the minimum pH shifts to low pH with increase in temperature. The second area involves understanding the metallurgy of aluminum alloys during fuel drying. In this study, the temperature range from 100°C to 300°C was analyzed. Thermodynamic models predict that AA1100 will not experience significant phase changes while the more highly alloyed AA6061 and AA5052 may experience significant phase changes. These thermodynamic measurements establish a potential for phase changes during drying, however they do not predict the kinetics for such processes. Phase 1 of the experimental program has initiated and results are forthcoming. The test program will examine aluminum oxidation under anticipated dry storage conditions (temperature and humidity). Examination of localized corrosion due to tight crevices and galvanic contact (with stainless steel) will be included. Finally, water deliquescence will be assessed using thermogravimetric analysis with dynamic humidity profiles.

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

Recovery and Recrystallization Behavior of Nb-Containing Fe-Cr-Al Alloys

Nuclear-grade wrought Fe-Cr-Al alloys are currently under development as promising accident-tolerant fuel cladding materials for light water reactors (LWRs) due to their superior oxidation and corrosion resistance in high temperature steam environments when compared with current commercial Zr-based alloys. Sufficient amounts of Cr and Al additions in Fe-Cr-Al alloys provide the improved surface protection effects at both the service temperature and elevated temperatures in the accident case scenario. Minor alloying additions (such as Nb in the current study) also promote the second-phase precipitates to strengthen the alloys; e.g. sub-micron size Fe{sub 2}Nb-type Laves phase precipitates are likely to disperse at both inter- and trans-granular in BCC-Fe matrix after applying proper heat treatments. In LWRs, a seamless tube product is required for the fuel cladding which typically has ∼ 4 m length with an outside diameter of ∼ 10 mm and a wall thickness of less than 0.5 mm, and it imposes considerable challenges for tube fabrication process (e.g. tube drawing) of Fe-Cr-Al alloys. Research has been performed to achieve a balance between high-temperature performance and tube fabricability in newly developed Fe-Cr-Al alloys. BCC-Fe materials with Cr and Al additions typically suffer from poor ductility because of relatively high ductile-brittle transition temperature. In this sense, the microstructural stability of refined grain or subgrain structure, even at high temperature, is critical to maintain the deformability of the alloys during tube fabrication process, as well as sufficient mechanical properties at service condition. In this study, the recovery and recrystallization behaviors of Nb-containing Fe-Cr-Al alloys are studied. A specific focus was to evaluate the stability of the subgrain structure of the deformed Fe-Cr-Al alloys, and seek optimized processing parameters for seamless tube fabrication. The present results clearly demonstrated that the Nb addition stabilizes the refined subgrain microstructures of Fe-Cr-Al alloys below 700 deg. C due to Laves phase precipitations along sub-boundaries when compared with the Nb-free alloys, while maintaining sufficient strengths needed for fuel cladding in LWRs. At the same time, the results also showed that Nb-containing alloys can be softened through suitable annealing to improve the processability of tube drawing process. Detailed characterization efforts of subgrains and Laves phase precipitates are currently in progress. Completion of this work will deepen the understanding of the recovery and recrystallized behavior of Nb-containing Fe-Cr-Al alloys and provide guidelines to tube drawing process in the alloys. (authors)

36 MATERIALS SCIENCE↗