Advanced Test Reactor Research Articles

Initial phase of Pu-238 production in Idaho National Laboratory

The initial phase of Plutonium-238 (Pu-238) production for radioisotope thermal generation is described here in detail. Two dosimeters with/without cadmium sleeve containing Neptunium-237 (Np-237) and other neutron sensors were inserted into the low power Advanced Test Reactor Critical (ATRC) for Pu-238 production testing. The gamma-ray energy measurements from Np-238, the short-lived intermediate product, confirmed that sizable amount of Pu-238 can be produced in the full power Advanced Test Reactor (ATR). The Pu-238 production determined from the irradiation experiment was in accord with the modelling predictions. Detailed studies of the Au/Cu sensors for thermal and epithermal flux analysis and their consistency for sensors in different locations, bare and in Cd-sleeve, provided confidence in the Pu-238 production data.

Overcoming challenges to support us resumption of high specific activity cobalt-60

Domestic production of high specific activity 60Co was halted after a target rupture in 2012 at the Advanced Test Reactor (ATR). The Isotope Program (IP) within the US Department of Energy (DOE) Office of Science tasked a multilaboratory team of researchers and managers from Oak Ridge and Idaho National Laboratories with the redesign the radioisotope capsule. The objective of this effort was to create a more robust and reliable design, compared to the pre-2012 target. The team successfully completed this task to produce the DOE-IP cobalt (Co) production capsule design. Furthermore, 66 capsules were successfully fabricated by Oak Ridge National Laboratory (ORNL) and delivered to Idaho National Laboratory (INL) for irradiation in the ATR between January 2014 and October 2016. This paper describes the efforts of the team to prepare and disposition the two initial DOE-IP Co production capsules that were processed in March 2020. These efforts include performing accurate production predictions, experimentally validating predictions with assay measurements, shipping with the Orano-furnished Battelle Energy Alliance Research Reactor shipping package, and disassembling capsules at the isotope vendor site.

Swelling of U-Mo Monolithic Fuel: Developing a Predictive Swelling Correlation under Research Reactor Conditions

Understanding swelling behavior in monolithic, uranium-molybdenum, plate-type fuel is necessary to qualify the fuel for reactor use and for the conversion of high performance research reactors from highly enriched to low-enriched uranium. Multiple mechanisms influence plate dimensional stability, including solid and gaseous fission-product induced swelling, irradiation-assisted creep, fuel-phase transformation, and interaction-layer formation. Separating these phenomena remains a challenge, and current models do not appear to adequately predict experimental results at higher fission densities where it is most critical. To mitigate the mechanistic uncertainty, post-irradiation profilometry is used to increase the number of data points available over a range of in-reactor irradiation experiments conducted at the Advanced Test Reactor, and to provide a statistical precedent for a swelling-behavior model. This work establishes a predictive swelling correlation as a function of fission density, with associated confidence and prediction bounds, by analyzing more than 18,000 thickness data points collected on 74 irradiated U-10Mo monolithic fuel test plates over a range of irradiation conditions.

Atom probe characterisation of segregation driven Cu and Mn–Ni–Si co-precipitation in neutron irradiated T91 tempered-martensitic steel

The T91 grade and similar 9Cr tempered-martensitic steels (also known as ferritic-martensitic) are leading candidate structural alloys for fast fission nuclear and fusion power reactors. At low temperatures (300–400 °C) neutron irradiation hardens and embrittles these steels, therefore it is important to investigate the origin of this mode of life limiting property degradation. T91 steel specimens were separately neutron irradiated to 2.14 dpa at 327 °C and 8.82 dpa at 377 °C in the Idaho National Laboratory Advanced Test Reactor. Atom probe tomography was used to investigate the segregation driven formation of Mn–Ni–Si-rich (MNSPs) and Cu-rich (CRP) co-precipitates. The precipitates increase in size and, slightly, in volume fraction at the higher irradiation temperature and dose, while their corresponding compositions were very similar, falling near the Si(Mn,Ni) phase field in the Mn–Ni–Si projection of the Fe-based quaternary phase diagram. While the structure of the precipitates has not been characterised, this composition range is distinctly different than that of the typically cited G-phase. The precipitates are composed of CRP with MNSP appendages. Such features are often observed in neutron irradiated reactor pressure vessel (RPV) steels. However, the Si, Ni, Mn, P and Cu solutes concentrations are lower in the T91 than in typical RPV steels. Thus, in T91 precipitation primarily takes place in solute segregated regions of line and loop dislocations. These results are consistent with the model for radiation induced segregation driven precipitation of MNSPs proposed by Ke et al. Cr-rich alpha prime (α’) phase formation was not observed.

Advanced Manufacturing of Printed Melt Wire Chips for Cheap, Compact Passive In-Pile Temperature Sensors

Melt wires are a passive sensor used to determine peak temperatures. Traditional melt wires are commonly used in test reactor experiments, such as in the Advanced Test Reactor. However, the conditions within a reactor present significant challenges towards test design, due to space limitations and the harsh environment. For example, some test capsules have only a couple millimeters in diameter available for instrumentation, which is too small to accommodate a traditional melt wire package. To enable instrumentation for space-limited applications, peak temperature sensors have been developed using additive manufacture to form printed melt wires. This paper reports on the design and fabrication of miniaturized melt wire chips with printed melt wires. This study advances the development of unique, compact temperature sensors capable of sensing user specified temperature ranges within the harsh environment of irradiation testing.

Results of the AGR-2 TRISO fuel performance demonstration irradiation experiment in the Advanced Test Reactor

UCO and UO2 tristructural isotropic fuel compacts were irradiated in the AGR-2 experiment, conducted in the Advanced Test Reactor for 559.2 effective full power days. UCO and UO2 compacts reached calculated peak burnups of 13.15 and 10.69% fissions per initial heavy-metal atom, and fast fluences of 3.47 × 1025 and 3.53 × 1025 n/m2 (E > 0.18 MeV), respectively. The time-average volume-average temperatures ranged from 987 to 1296 °C in UCO compacts and from 996 to 1062 °C in UO2 compacts. Fission product release-to-birth (R/B) ratios remained below 2 × 10−6 in UCO compacts and 10−7 in UO2 compacts during the first three irradiation cycles. R/B data then became unreliable due to mixing of the capsule gas flows which hindered the evaluation of fuel performance for the later portion of the irradiation. Post-irradiation examination of the irradiation capsules and fuel compacts is underway and will provide additional information on fuel performance.

α-U and ω-UZr2 in neutron irradiated U-10Zr annular metallic fuel

To develop metallic fuel with ultra-high burnup of 30%-40%, an annular U-10Zr fuel with 55% smear density was fabricated through a casting route and irradiated at the Advanced Test Reactor at Idaho National Laboratory. The annular fuel design also serves as a demonstration of the feasibility to replace the sodium bond with a helium bond to benefit the geological disposal of irradiated fuel, cut the cost of fuel fabrication, and boosts the overall metallic fuel economy. This paper reports the results from transmission electron microscopy (TEM) based post-irradiation examination of this fuel type irradiated to a burnup of 3.3% fissions per initial heavy metal atoms. The low burnup was planned for initial screening of this fuel design. After irradiation, the initial U-10Zr microstructure separated into an α-U annular region and an UZr2+x center region with a nanoscale spinodal decomposed microstructure. Because of the large amount of interface areas created in this microstructure, the fission gas atoms and vacancies generated in the UZr2+x phase are possibly pinned at the interface areas, leading to ~20 times smaller fission gas bubbles than those in the neighboring α-U. The large bubbles in α-U become connected and merged into large pores that provide fast paths for fission gas release into capsule plenum which prevents further swelling of fuel slug. The fuel slug center still has open space to accommodate further fuel swelling from solid fission products at higher burnup. Other neutron irradiation induced phase and microstructure change are also characterized and compared with traditional solid fuel designs.

Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels

This work was part of a program begun in 2001 to develop advanced nuclear fuels, originally as carriers for plutonium and minor actinides (neptunium, curium, and americium) taken from spent commercial light-water reactors (LWR) so that the plutonium and minor actinides could be ‘burned’ or transmuted in an accelerator or a fast nuclear reactor. A central part of these experiment programs has been the development of advanced fast reactor fuels, because a fast reactor was considered the most efficient vehicle to transmute the actinide waste products, and metallic fuels is a central focus of these tests. An experiment design was developed in which a thermal test reactor, the Advanced Test Reactor (ATR), was used to test small fuel pin prototypes, by creating areas in the core shielded by cadmium filters to produce a largely epithermal and fast neutron spectrum environment in which the pins could be irradiated. The results of non-fertile metallic fuel (no uranium) tests are presented here.Pu-Am-Np-Zr fuels were irradiated to fission densities up to 33 × 1020 fission/cm3 and Pu-239 depletions of up to 39%. The depletions were created by roughly 2/3 by fission and 1/3 by transmutation neutron capture. Up to five fuel ‘rodlets’ were irradiated in three sealed capsules stacked axially in the core, and the peak cladding temperatures ranged from 300°C to 500°C, depending on axial location as those near the core centerline are operating hotter and to higher fission densities. Several post-irradiation examinations (precision gamma scanning and fission gas release) were similar to other historical metal fuel experiments in fast reactors. However, optical metallography indicated that two of the rodlets had breached. The exact reasons are unclear. Due to the design of this irradiation experiment a rodlet breach could have increased the temperature in others in the same capsule by contaminating the thermal gap helium with heavier and less conductive fission product gases. Some of those rodlets showed high amounts of fuel/cladding chemical interaction (FCCI).

Microstructural response of the fuel phase in U-7Mo dispersion fuel irradiated at different powers

Irradiation conditions such as fission power, fission rates, and temperature are important parameters for nuclear fuels because they influence the microstructural behavior and ultimate irradiation performance. Two U-7Mo/Al-3.5 wt.% Si dispersion fuel plates were irradiated in the Advanced Test Reactor edge-on to the core in the RERTR-9B experiment at different powers to study the effects of power on fuel phase swelling behavior. The results of non-destructive measurements have been reported in a previous paper [1], and the destructive examination results are being reported here. In particular, the microstructural evolution of the U-Mo fuel phase in irradiated fuel plates has been investigated by cutting transverse cross-sections, using generated scanning electron microscopy micrographs, and image processing methods to assess the changes in microstructural features, such as fission gas pore growth and non-recrystallized grain fraction. In this study, the focus was to examine the evolution of fission gas pores from both the constrained and unconstrained regions of U-Mo fuels irradiated at different powers. Based on an estimated linear gradient, the rate of change of fission gas pore size with locally calculated fission densities is more prominent in the higher power specimen. As such, the average porosity and pore area is approximately 11% and 2 × larger in the fuel irradiated at higher power than that of the lower power specimen. The growth of fission gas pores can accelerate the pore interconnectivity and thus fission gas outside the fuel kernel (FGO), which are two precursors for swelling in U-Mo fuels. The influence of increased power, internal plate stresses, and fission-induced creep on fuel kernel deformation and lateral mass transfer is also highlighted.

A diffusion synthetic acceleration approach to k-eigenvalue neutron transport using PJFNK

This paper investigates a diffusion synthetic acceleration (DSA) approach to the k-eigenvalue problem of the neutron transport equation. The novelties of this study reside in 1) casting the low-order diffusion equation directly as a nonlinear problem, which is solved with the PJFNK (preconditioned Jacobian-free Newton Krylov) method, circumventing the need for power iteration; 2) showing the appealing features on solving the diffusion equation with DSA compared with nonlinear diffusion acceleration (NDA) with a simplified model of ATR (Advanced Test Reactor); 3) proposing the DSA weak form for the discretization scheme with SAAF (self-adjoint angular flux) formulation that is stable with anisotropic scatterings; 4) showing the impact of the vacuum boundary coefficient of the low-order equation on the convergence of DSA. Both DSA and NDA are implemented in and tested with Rattlesnake code, the INL MOOSE-based radiation transport application for multi-physics modeling and simulations.

Characterization of a minor actinides bearing metallic fuel pin irradiated in EBR-II

The X501 transmutation experiment irradiated in Experimental Breeder Reactor II (EBR-II) was intended to evaluate the safety and performance of the addition of minor actinides (Am, Np) into the metallic nuclear fuel system.The irradiation of this experiment is also of paramount importance for comparisons between true fast spectrum irradiations (e.g. EBR-II) and shrouded irradiation capsules performed at the thermal neutron spectrum Idaho National Laboratory Advanced Test Reactor. The X501 experimental fuel pins were irradiated to a peak measured burnup of 6.2% fission per initial heavy metal atom. This work reports and discusses Postirradiation Examination results for one of two minor actinide bearing pins from the EBR-II X-501 experiment (X501-G591) with a focus on fuel microstructure evolution.Fuel swelling, fission product distribution, cladding strain, fission gas release, fuel microstructure and fuel cladding chemical interaction were all evaluated. The fission gas release was 79.9% and the microstructure along the fuel pin presented several characteristics very similar to other EBR-II U-Pu-Zr ternary metallic fuel experience. Minor actinides seem to not dramatically affect the overall performance of this candidate transmutation fuel. Scanning Electron Microscopy was also performed and an in-depth characterization of the different regions and phases is presented. Performance data from these irradiations can be used to inform the feasibility of minor actinide transmutation in future reactor systems.

The effect of composition variations on the response of steels subjected to high fluence neutron irradiation

A set of low alloy model reactor pressure vessel steels, with systematic variations in their Mn, Ni, and Si contents, were neutron-irradiated to high fluence (1.4 × 1020 n/cm2) in the Advanced Test Reactor at Idaho at 290°C and a flux of 3.6 × 1012 n/cm2s. The alloys were analysed using atom probe tomography and solute clusters were observed in each alloy, including in one alloy that contained low nominal levels of Mn (0.04 at. %) and Si (0.06 at. %). Changes in the mechanical properties of the alloys were correlated with cluster volume fractions. Whilst the effect of nominal composition was observed to influence cluster composition, cluster nucleation site was not observed to affect composition. Several grain boundaries were also analysed and the segregation behaviour of certain elements is discussed.

The correlation between microstructure and nanoindentation property of neutron-irradiated austenitic alloy D9

The microstructure and nanomechanical properties of three samples of the modified stainless steel (referred as the alloy D9) were systematically characterized after neutron irradiation in the Advanced Test Reactor. The samples were irradiated to 5.0, 8.2, and 9.2 displacements per atom at 448, 430, and 683°C, respectively. The evolutions of dislocation loops, cavities, and radiation-induced precipitates were quantitatively studied to reveal their dose and temperature dependencies. Nanohardness and nanoindentation creep tests were conducted at room temperature on the irradiated samples. Unexpected radiation hardening was observed in the highest-temperature-irradiated sample due to the formation of an unknown type of Ni- and Si-rich precipitates whose contributions to the radiation and mechanical performances of the alloy were discussed. We provide the radiation-microstructure-property correlations of alloy D9 with new insights, which can benefit the development and optimization of advanced austenitic alloys for future nuclear applications.

Microstructure evolution in MA956 neutron irradiated in ATR at 328 °C to 4.36 dpa

MA956 is an iron-chromium-aluminum (FeCrAl) based oxide dispersion strengthened (ODS) alloy produced by mechanical alloying. The alloy was irradiated in the Advanced Test Reactor (ATR) at 328 °C up to 4.36 dpa with both thermal and fast neutrons. The microstructures before and after irradiation were investigated by various TEM techniques. The size and number density of dislocation loops, voids, and oxides were analyzed. The results showed that both 1/2<111> and <100> loops were generated in irradiated materials. The number density of dislocation loops reached up to 7.17 × 1021/m3. Cavities/voids were formed during irradiation. Argon bubbles were observed in unirradiated materials attached to the surface of oxide particles. The swelling rate was estimated to be 0.08% without subtracting pre-existed Ar bubbles. The oxide particles could maintain crystal structure during irradiation. The oxides had small decrease in size but half reduction in number density after irradiation. Calculation estimated that <100> loops and dislocation lines contributed most to hardening. Alpha prime precipitates were suggested to be formed by comparing to the calculated hardening. Dislocation lines are demonstrated not forming tangles in irradiated specimens.

Performance of Custom-Made Very High Temperature Thermocouples in the Advanced Gas Reactor Experiment AGR-5/6/7 during Irradiation in the Advanced Test Reactor

The Advanced Gas Reactor-5/6/7 (AGR-5/6/7) experiment is the fourth and final experiment in the AGR experiment series and will serve as the formal fuel qualification test for the TRISO fuels under development by the U.S. Department of Energy. Certain locations in this experiment reach temperatures higher than any of the previous AGR tests, up to 1500°C. Such extreme temperatures create unique challenges for thermocouple-based temperature measurements. High-temperature platinum-rhodium thermocouples (Types S, R, and B)and tungsten-rhenium thermocouples (Type C) suffer rapiddecalibration due to transmutation of the thermoelements fromneutron absorption. For lower temperature applications, previousexperience with Type K thermocouples in nuclear reactors haveshown that they are affected by neutron irradiation only to alimited extent. Similarly, Type N thermocouples, which are morestable than Type K at high temperatures, are only slightly affectedby neutron fluence. Until recently, the use of these nickel-basedthermocouples was limited when the temperature exceeds 1050°Cdue to drift related to phenomena other than nuclear irradiation.Recognizing the limitations of existing thermometery to measuresuch high temperatures, the sponsor of the AGR-5/6/7 experimentsupported a development and testing program for thermocouplescapable of low drift operation at temperatures above 1100°C. High Temperature Irradiation Resistant Thermocouples (HTIR-TCs)based on molybdenum/niobium thermoelements have been underdevelopment at Idaho National Laboratory (INL) since circa 2004. A step change in accuracy and long-term stability of thisthermocouple type has been achieved as part of the AGR-5/6/7thermometry development program. Additionally, long-termtesting (9000+ hrs) at 1250°C of the Type N thermocouplesutilizing a customized sheath developed at the University ofCambridge has been completed with low drift results. Both theimproved HTIR and the Cambridge Type N thermocouple typeshave been incorporated into the AGR-5/6/7 test, which beganirradiation in February 2018 in INL’s Advanced

Effect of diamond additive on the fission gas release in UO2 fuel irradiated to 7.2 GWd/tHM

UO2 pellets containing 5 vol% diamond particles were irradiated in the Advanced Test Reactor at an LHGR of up to 310 W/cm to the burnup of 7.2 GWd/tHM. Fission gas release, measured during post-irradiation examination, was determined to be 1.08%. Fission gas release modeling performed using BISON fuel performance code showed that undoped UO2 fuel irradiated under the identical conditions, would have had fission gas release of 9.09%. Noting that diamond has the highest thermal conductivity of any known material, these results suggest that doping with a good thermal conductor is an effective means to reduce fission gas release in UO2 fuels.

Fission gas monitoring for the AGR-5/6/7 experiment

The AGR-5/6/7 test is a fueled multiple-capsule experiment currently being irra- diated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). AGR-5/6/7 is the fourth and last experiment in the “AGR-experiment” series which is being performed and funded by the United States Department of Energy (DOE) as part of the INL Advanced Reactor Technologies (ART) program. To derive isotopic release-to-birth ratio’s, the fission gas measured at the Fission Product Monitoring System (FPMS) must be corrected for decay during the transport from the constantly-irradiated fuel capsule to the FPMS measurement location. The decay time is dependent on the capsule effluent flow rate and transport volume. Capsule effluent flow rates are measured and recorded into the ATR Distributed Control System (DCS). Therefore, only the capsule specific transport volumes are needed for accurate transport decay corrections. This paper contains a summary of the AGR-5/6/7 experiment test train, details of the leadout flow experiment, the process of obtaining the short lived gaseous fission product data for use in determining the transport volumes and the AGR-5/6/7 transport volumes that will be used to calculate the capsule specific isotopic release activities and preliminary release-to-birth ratio’s for the AGR-5/6/7 experiment.

Design Studies for the Optimization of 238Pu Production in NpO2 Targets Irradiated at the High Flux Isotope Reactor

Efforts to reestablish a domestic 238Pu production capability in support of National Aeronautics and Space Administration mission objectives are ongoing throughout the U.S. Department of Energy complex. The Plutonium-238 Supply Project (PSP) was initiated in response to a report published by the National Research Council in 2011 stating that “without a restart of 238Pu production, it will be impossible for the United States, or any other country, to conduct certain important types of planetary missions after this decade.” The PSP is targeting a sustained, constant production rate of 1.5 kg/year of heat source PuO2 for several years. Design and optimization studies of 237Np-bearing targets are underway at Oak Ridge National Laboratory (ORNL). It is anticipated that targets will be irradiated in ORNL’s High Flux Isotope Reactor (HFIR) and in the Advanced Test Reactor (ATR) at Idaho National Laboratory. A variety of target materials, containments, arrangements, and irradiation histories have been analyzed, and the results indicate that a sufficient quantity of 238Pu can be produced in HFIR and ATR to fulfill the PSP’s constant production rate target. This paper focuses on the design and optimization of new target configurations containing pellets that are (1) ~93% of the theoretical density of NpO2, (2) loaded into pins of cladding materials that can be handled as solid waste following postirradiation 238Pu recovery operations, (3) irradiated in various vertical experiment facility (VXF) locations in the HFIR permanent beryllium reflector, and (4) rotated within and/or moved to another VXF location following each HFIR operational cycle to maximize 238Pu production and minimize peak heat generation rates.

The use of U3Si2/Al dispersion fuel for high power research reactors

In order to investigate the irradiation performance of U3Si2/Al dispersion fuel at relatively aggressive reactor conditions, compared to how this fuel is typically used in a research or test reactor, post irradiation examination was performed on two fuel plates that were irradiated at ∼270 W/cm2 average surface heat flux and to a maximum local burnup greater than 8 × 1021 fissions/cm3 in the Advanced Test Reactor as part of the RERTR-8 experiment. As part of the non-destructive examinations that were performed on both fuel plates, detailed thickness measurements using a high-fidelity measurement bench were performed on one fuel plate and a segment of the second fuel plate. This paper provides the first reported results of such high-resolution thickness measurements for U3Si2/Al dispersion fuel plates, containing primarily the U3Si2 fuel phase, irradiated at such power levels to high burnup. For destructive examination using optical metallography, samples were generated from both irradiated plates. Localized regions of high swelling (characteristic of internal blistering) were observed on the surface of both fuel plates, suggesting that the performance limits had been reached for each plate. Destructive examinations showed that the fuel particles that are primarily U3Si2, which become amorphous during irradiation, in both plates contained relatively large fission gas bubbles with evidence of fission gas bubble interconnection.

A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

The Advanced Test Reactor (ATR) has been used successfully for the testing of fast reactor fuel for nearly two decades. These successes have been in spite of numerous challenges for testing fast reactor fuel in the ATR (a thermal spectrum reactor), but the solutions to those challenges have resulted in excessively long irradiation times (~10 years) for high-burnup targets as well as experiments that are highly sensitive to fabrication tolerances and eccentricities. This paper presents a solution to the problems of extended irradiation times and fabrication sensitivities. Thermal and neutronic analyses were performed to show that a reduced-diameter fuel pin with an equivalent linear heat generation rate can provide a prototypic thermal profile (peak centerline and inner clad temperature) along with a near-prototypic power profile within the ATR thermal spectrum. This allows the experiment to reach a high burnup in an expeditious timeframe compared to traditional ATR fast fuel irradiations. In addition, problems with fabrication sensitivities were addressed by introducing a double-encapsulated experiment that pushes the high heat flux helium gap farther away from the fuel pin. Fuel pin position eccentricities are also mitigated by using a large sodium bond between the pin and capsule fuel. The advantages and potential pitfalls of this revised design are discussed, including the effect of length scales on fuel system behavior.