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When the Halden Boiling Water Reactor closed down in 2018, a need to restore the capability for in-reactor power ramp testing arose. Such testing is valuable for studying pellet-clad interaction phenomena in nuclear fuels. The data from these studies is of great interest to a number of research programs, including the accident-tolerant fuel (ATF) program at Idaho National Laboratory (INL). In 2022, Woolstenhulme et al. proposed several power ramp testing ideas using facilities at INL, including irradiation in the Transient Reactor Test Facility (better known as TREAT) and the Advanced Test Reactor (ATR) [1]. Worrall et al. [2] and Labossiere-Hickman et al. [3] subsequently performed feasibility studies for the ATR testing options in 2023. This summary further investigates the three-pin trefoil design (Fig. 1) for the proposed ATF-2Ramp Experiment discussed in Labossiere-Hickman et al. [3]. ATF-2Ramp is designed to operate in the center flux trap (CFT) of the ATR during a powered axial locator mechanism (PALM) cycle: a short, variable-powered cycle with an asymmetric power distribution. Previously, it was shown that tailoring the thickness of the hafnium (Hf) neutron shields (“mini-shrouds”) surrounding each pin offered a degree of control sufficient to achieve the programmatic linear heat generation rate (LHGR) targets for ATF-2Ramp during the high-power period of a PALM cycle. New work involves shortening the experiment test train for consistency with the fuel pins in ATF-2D [4] and then shaping the axial power profile of the three test pins.
The recent trend towards a high burn-up discharge spent nuclear fuel necessitates a thorough understanding of hydrogen uptake in Zr-based cladding materials that encapsulate spent nuclear fuel. Although it is challenging to experimentally replicate exact conditions in a nuclear reactor that lead to hydrogen uptake in claddings, in this study we have attempted to understand the kinetics of hydrogen uptake by first electrolytically charging Zircaloy-2 (Zr-2) cladding material for various durations (100 to 2,600 s), and subsequently examining hydrogen ingress with elastic recoil detection (ERD) and transmission electron microscopy (TEM). To understand the influence of irradiation damage defects on hydrogen uptake, an analogous study was performed on ion - irradiated (0.1, 1 and 25 dpa) Zr-2. Analysis of ERD data from the un-irradiated Zr-2 suggests that the growth of the hydride layer is diffusion controlled, and preliminary TEM results support this assertion. In un-irradiated Zr-2, the diffusivity of hydrogen in the hydride phase was found to be approximately 1.1 Multiplication-Sign 10{sup -11} cm{sup 2}/s, while the diffusivity in the hydride phase for lightly irradiated (0.1 and 1 dpa) Zr-2 is an order of magnitude lower. Irradiation to 25 dpa results in a hydrogen diffusivity that is comparable to the un-irradiated Zr-2. These results are compared with existing literature on hydrogen transport in Zr - based materials.
Process development studies are being conducted to recover, purify, and reuse the zirconium (about 98.5% by mass) in used nuclear fuel (UNF) zirconium alloy cladding. Feasibility studies began in FY 2010 using empty cladding hulls that were left after fuel dissolution or after oxidation to a finely divided oxide powder (voloxidation). In FY 2012, two industrial teams (AREVA and Shaw-Westinghouse) were contracted by the Department of Energy Office of Nuclear Energy (NE) to provide technical assistance to the project. In FY 2013, the NE Fuel Cycle Research and Development Program requested development of a technology development roadmap to guide future work. The first step in the roadmap development was to assess the starting point, that is, the current state of the technology and the end goal. Based on previous test results, future work is to be focused on first using chlorine as the chlorinating agent and secondly on the use of a process design that utilizes a chlorination reactor and dual ZrCl{sub 4} product salt condensers. The likely need for a secondary purification step was recognized. Completion of feasibility testing required an experiment on the chemical decladding flowsheet option. This was done during April 2013. The roadmap for process development will continue through process chemistry optimization studies, the chlorinated reactor design configuration, product salt condensers, and the off-gas trapping of tritium or other volatile fission products from the off-gas stream. (authors)
Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. This allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to utilize lower length scale approaches (e.g., density functional theory, cluster dynamics, rate theory, phase field, and Visco-Plastic- Self-Consistent (VPSC)) to develop more physically informed models at the engineering scale for use in the BISON [9] fuel performance code.
This study investigates the hydride rim microstructure at the metal-oxide interface of Zircaloy-4 cladding segment removed from H.B. Robinson Nuclear Reactor by utilizing high resolution electron microscopy techniques with energy dispersive x-ray spectroscopy at Oak Ridge National Laboratory under the NSUF Rapid Turnout Experiment program. A complex stacking and orientation of hydride platelets has been observed below the sub-oxide layer. Furthermore, radial hydride platelets have been observed. EDS signals of both Fe and Cr has been reduced within hydrides whereas EDS signal of Sn is unaffected.
In this paper we report molecular dynamics simulation results of iron (cladding) corrosion in interaction with lead coolant of fast nuclear reactor. The goal of this work is to study effect of oxygen injection to the coolant to reduce iron corrosion. By evaluating diffusion coefficients, radial distribution functions, mean-square displacement curves and observation of crystal structure of iron before and after oxygen injection, we concluded that a significant reduction of corrosion can be achieved by issuing about 2% of oxygen atoms into lead coolant.
There is a need in nuclear reactors to inspect irradiated materials and structures. A portable scanning infrastructure for a material characterization technique called vibro acoustography (VA) is being developed by the Idaho National laboratory for nuclear applications to characterize fuel, cladding materials, and structures. The proposed VA technology is based on ultrasound and acoustic waves; however, it provides information beyond what is available from the traditional ultrasound techniques and can expand the knowledge on nuclear material characterization and microstructure evolution. VA technique has several advantages over traditional characterization techniques. VA is a non-contact submersion technique that is inherently wireless transmission and reception (from the ultrasonic transmitter to the acoustic receiver). The transmission medium can be most liquids used for reactor cooling such as water and sodium. For nuclear in-pile applications, the Ultrasonic Transducer (UT) will need to have a line of sight location that is within lower radiation flux zones. Multiple UT sources can be frequency multiplexed to allow a single hydrophone receiver to monitor several locations. The reactor structure can also be an integral part of the transmission system acting as waveguides to deliver the acoustic signal to an acoustic receiver. Because of the long wavelength of the acoustic signal, the acoustic receiver does not have to be line of sight with respect to the component under test. The focusing and generation of the acoustic beat frequency from two ultrasonic waves allows for the potential to ultrasonically image complex shapes. These features will significantly reduce experiment complexity, installation times and cost while being safe and easy to use.
The BR2 nuclear reactor is a material testing reactor (MTR) located in Mol, Belgium, and operated by the Belgian Nuclear Research Centre (SCK CEN) since 1963. The reactor is highly versatile as the number and location of fuel elements and control rods can change significantly from cycle to cycle to accommodate different needs. Argonne National Laboratory (ANL or Argonne) Reactor Conversion (RC) team has collaborated with SCK CEN for over a decade on the conversion of domestic and international research reactors from highly enriched uranium (HEU, ≥20 wt.% of 235 U) to low enriched uranium (LEU, <20 wt.% of 235 U) fuel. The U.S. High-Performance Research Reactor (USHPRR) project within the M3 Reactor Conversion Program aims at converting five U.S. high performance research reactors (MITR, MURR, NBSR, HFIR, and ATR) and one critical facility (ATR-C) to LEU fuel. These USHPRRs still use and regularly refuel with HEU fuel. Each facility has a unique reactor design, operating conditions, and fuel element design to accomplish its mission. The goal of the USHPRR project is to convert the USHPRRs and the critical facility to LEU fuel while maintaining experimental performance and ensuring safe facility operation. The current technical report focuses on two reactors requiring very high-density LEU fuel: the Massachusetts Institute of Technology Reactor (MITR) and the National Bureau of Standards Reactor (NBSR). To support the conversion of these reactors, so-called design demonstration elements (DDE) are planned to be irradiated in the BR2 reactor under conditions similar to the targeted reactors and using a prototypic geometry. In support of this experiment, SCK CEN studied and modeled the DDE irradiations using MCNP6.2 to investigate the feasibility of irradiating the MITR DDE and NBSR DDE in BR2. Argonne reviewed and confirmed the conclusions of this study. Structural analysis is another step toward converting USHPRR to LEU fuel. The objective of the current report is to provide information useful to the structural analysis of the NBSR & MITR DDEs to support its irradiation in BR2. Specifically, the goal is to provide the fast neutron (E>0.1MeV) fluence in the cladding of the fuel plates in BR2 for the whole period of irradiation (8 cycles for MITR DDE and 10 cycles for NBSR DDE). Additionally, fast neutron fluences in the side plates and in the NBSR DDE’s outside plates were calculated and reported. Neutronic calculations were performed using MCNP6.2 on the RTRHPC cluster.
The Fukushima Daiichi nuclear accident in 2011 stimulated materials development efforts to improve the current fuel design for Light Water Reactors (LWRs). From fuel element to fuel cladding, new concepts in advanced nuclear fuel development have been proposed, studied, and tested. The new fuel systems are expected to have better resistance to damage, melting, and chemical reaction during accident scenarios. To test the performance and accident tolerance of the advanced fuel, it is highly preferable to simulate off-normal transients in addition to standard in-core fuel performance tests (available at High Flux Isotope Reactor at Oak Ridge National Laboratory and Advanced Test Reactor at Idaho National Laboratory). The Transient Reactor Test Facility (TREAT) is now being prepared to restart its operation, performing transient testing to assist advanced fuel development. TREAT was operated from 1959 to 1994, and was extensively used to generate thousands of reactor transients for different programs. The reactor has not operated since 1994. The original fuel used at TREAT was graphite-based dispersion fuel with highly-enriched uranium (HEU, 93.1% enriched uranium). For nonproliferation purposes, a fuel conversion program, supported by U.S. Department of Energy, National Nuclear Safety Administration (NNSA), Office of Material Management and Minimization (NA-23) Reactor Conversion Program, is underway to develop a replacement core in which the UO{sub 2} contains only low-enriched uranium (LEU). The preliminary LEU core design has shown a much lower carbon/uranium atomic ratio and enrichment than in the HEU core design. To achieve performance similar to the HEU core, the prospective LEU core will contain a much higher density of UO{sub 2} particles. As a result, a higher level of radiation damage, particularly from fission fragments emerging from the UO{sub 2} particles, can be expected dependent on the UO{sub 2} particle size within the fuel. To support LEU fuel design, the present study investigates the effect of reactor radiations on the thermal conductivity of graphite-based dispersion fuel. Both neutron- and fission-radiation induced degradations of the fuel are modeled, and validated based on the available experimental data. In this study, the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel was analyzed. A semi-empirical method to model the relative thermal conductivity with reactor radiations was proposed and validated. The proposed method can be further applied to predict the changes in the thermal conductivity of the prospective LEU TREAT fuel during long-term operation. (authors)
Accurate, real-time monitoring of strain in fuel, cladding, and structural components of nuclear reactors is critical to better understand radiation induced phenomena during reactor tests and operations. The data provided is crucial to verify physics-based, multiscale modeling and simulation efforts, which aim to shorten the timeline for the development of new nuclear materials. Resistive strain gauges have limited performance during in-pile experiments due to the harsh operating conditions and limited physical space between fuel and cladding components. In this work, aerosol jet printing using silver nanoparticle inks was used to fabricate interdigitated electrode capacitive strain gauges on aluminum alloy 6061 tensile specimens. Here to simulate the temperatures of a traditional light water reactor, the capacitive strain gauges were tested with a mechanical test frame up to 300 °C and compared to commercially available bondable resistive strain gauges. The printed capacitive strain gauges exhibited a gauge factor of 1.0 and showed higher reproducibility and predictability of strain sensing performance than the resistive strain gauge. The results demonstrate the potential of aerosol jet printing to fabricate strain sensors with predictable performance and reduced invasiveness for high-temperature applications with confined spacing.
The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)
Nuclear fuel encounters severe thermomechanical environments in which its mechanical response is determined by its microstructure, temperature and stress level histories. Simulating the response of such microstructures is crucial for predicting both performance and transient fuel mechanical responses and experimental verification of such predictions is therefore of great interest. While most of the deformation in a nuclear fuel rod occurs in the cladding, deformation of the fuel itself is still of interest with deformation mechanisms at operating temperature and above including creep, swelling, cracking as well as pellet-clad interaction. Characterization of these properties and understanding of the underlying deformation phenomena at operating or excursion temperatures is therefore of great importance for development and ultimately licensing of improved and novel nuclear fuel forms. Diffraction techniques offer unique insight on the atomistic (e.g. crystal structure) and microstructure (e.g. phase transformations, texture, defects) length scales and have a long history of providing unique data to inform relevant deformation models that enable the required predictive capabilities. For example, dislocations lead to diffraction peak broadening that can be characterized to estimate the dislocation density and study the role of dislocations on the deformation while measuring lattice strains allows to studie load sharing in two phase materials. In this report the requirements for a sample environment for high temperature deformation of nuclear fuels are defined. The HIPPO neutron time-of-flight diffractometer at LANSCE will host this sample environment and is also described. This instrument covers diffraction angles from 140° to 40° and is also equipped with an event-mode neutron imaging detector system, enabling energy-resolved neutron imaging in parallel with the diffraction that could measure sample temperature from Doppler broadening of neutron absorption resonances or measure pore densities from changes in the attenuation. Designs of devices to characterize thermomechanical properties of nuclear fuel without diffraction are also considered to guide the design. While this report is focused on applications for nuclear fuels, the device can also characterize cladding, moderator or structural materials and therefore contribute to other fields of research and development for advanced reactors. The temperatures planned to be reached are above 2000℃, thus enabling characterization of LWR reactor fuels under accident scenarios but also reaching temperatures of fuels developed for nuclear thermal propulsion and providing opportunities to characterize those. In conjunction with the energy-resolved neutron imaging detector, this setup would allow to measure neutron cross-sections at high temperatures, filling a gap towards development of reactors operating at high temperatures.
Nuclear fuel rod is composed of cladding tube and a specified number of fuel pellets contained. In the United States, spent nuclear fuel (SNF) is expected to be transported to at least one storage facility before permanent disposal. The fatigue behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending must be understood in order to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup SNFs (>45 GWd/MTU). This report presents the experimental results related to Zircaloy (Zry)-4 - based surrogate rods and high-burnup SNFs, based on recent work performed at Oak Ridge National Laboratory (ORNL). The surrogate rod was made of Zry-4 cladding and alumina pellets, and high-burnup fuel rods were discharged from H.B. Robinson pressurized water reactor. The reversed cyclic bending testing was conducted at 5 Hz under loading control. The effect of pre-hydriding and burnup or irradiation on the flexural rigidity and fatigue life of cladding-pellet system were discussed. The fatigue data obtained are extremely useful to the future certification of SNF storage and transportation cask.
The mechanical interaction between the fuel and cladding that occurs during operation of a nuclear reactor is important to understand as it can lead to cladding failures and release of radioactive material into the coolant. Additionally, in order to develop better models of the pellet-clad mechanical interactions, the mechanical properties of the fuel at relevant operating temperatures, like the elastic moduli, are needed for current and advanced accident tolerant fuels (ATFs). In this work, elevated temperature nanoindentation and resonant ultrasound spectroscopy were used to measure the moduli and hardness of several fluorite materials (CeO 2 , ThO 2 , UO 2 ) and several ATF candidates (ATF) (U 3 Si 2 , UN, UB 2 ). In addition, a comparison of the two techniques was performed in this study to independently validate the mechanical properties.
The development and deployment of accident-tolerant cladding materials in light water nuclear reactors requires the collection of neutron-irradiated properties of these materials with specific consideration to the material properties orientation (i.e., tube geometry). To that end, a test plan to irradiate several cladding tube and ring specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was proposed, along with the development of a new capsule design that features ring specimens with a preexisting gauge section. This ring specimen geometry removes the need to machine a gauge section on the irradiated specimen and thus allows for direct application of a ring-pull test post-irradiation. The test matrix includes uncoated and chromium-coated Zircaloy-4 and advanced zirconium alloy cladding specimens, pre-characterized at ORNL. Plans are to irradiate the specimens at about 4 and 10 dpa—corresponding to two and five cycles at the HFIR centerline—with a target specimen gauge section temperature of 330°C.
Portable Vibro-acoustic Testing System for In situ Microstructure Characterization and Metrology James A. Smith, Corrie I. Nichol Idaho National Laboratory, Idaho Falls ID Mostafa Fatemi, Mayo Clinic, Rochester MN There is a need in research reactors like the INL to inspect irradiated materials and structures. The goal of this work is to develop a portable scanning infrastructure for a material characterization technique called vibro-acoustography (VA) that has been developed by INL [1] for nuclear applications to characterize fuel, cladding materials, and structures. The proposed VA technology is based on ultrasound and acoustic waves; however, it provides information beyond what is available from the traditional ultrasound techniques and can expand the knowledge on nuclear material characterization and microstructure evolution. VA is a three-dimensional (3D) imaging modality based on ultrasound-stimulated acoustic emission. VA uses the force caused by the beating of two frequencies to generate an acoustic emission signal. Due to absorption or reflection, the energy density in the object at an acoustic focal point changes to produce a “radiation force.” This force locally vibrates the object, which results in an acoustic field that depends on the characteristics of the object at that point. This acoustic field is detected for every point in the object; the resulting data is used to make an image of the object’s mechanical properties. This technology is fundamentally different from traditional ultrasound imaging. More specifically, VA is: (a) based on the “radiation force,” a nonlinear phenomenon in sound propagation, and (b) sensitive to both high and low frequency dynamic responses of the object. The combination of these two response types provides information about characteristics of the object beyond what is possible by traditional ultrasound imaging. High spatial resolution is achieved, since the radiation force is localized to the focal point of the ultrasound beam. Low-frequency information, which is a representative of structural characteristics, is gained through the radiation force mechanism at a low (beat) frequency. This paper will report on the development of a portable scanning that can be set up to characterize materials and component in open water reactors and canals in situ. We will show some initial laboratory results of images generated by vibro-acoustics of surrogate fuel plates and graphite structures and discus the design of the portable system. 1. Smith, J. A., and Kunerth, D. C., “Vibro-acoustic Sensors for Materials Characterization and Related Methods and Systems,” Patent Application 13226280, 09/06/2011.
After uranium, the second largest mass contained in most used nuclear fuel (UNF) is the zirconium-based Zircaloy{sup R} cladding in the fuel cladding which accounts for about 25% of the mass. In current practice, the cladding requires disposal in a geologic repository. Zirconium may be recovered from the Zircaloy cladding of UNF for recycle. Alternatively, it may be decontaminated to recategorize it as low-level waste. Either approach reduces the quantities of high-level waste destined for a geologic repository, thereby reducing the overall cost of recycling UNF. Process development studies are under way in the United States and Republic of Korea based on dry chlorination using a mixture of chlorine (Cl{sub 2}) and an inert carrier such as argon (Ar) to convert the zirconium in the alloy, which is greater than 98 wt% of the alloy mass, to zirconium tetrachloride (ZrCl{sub 4}). The ZrCl{sub 4} is volatile and is easily separated from the non-volatile components. The residues include non-volatile chlorides such as CrCl{sub 3}, oxychlorides and oxides. The product ZrCl{sub 4} is recovered as a solid salt by condensation when cooled to less than about 300 deg. C. Feasibility tests with actual UNF cladding have demonstrated that the recovered zirconium chloride salt is sufficiently separated from fuel and fission product radionuclides that it emanates no significant dose and can be handled without shielding. Several methods for capturing tritium released from cladding during the chlorination process have been identified. A detailed description of the methodologies, advantages, drawbacks, and conceptual flowsheets will be presented. (authors)