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DOE OSTI2024

An In Situ transmission electron microscopy study on the synergistic effects of Au-ion irradiation and high temperature on nuclear graphite microstructure

Abstract The combined effects of high-temperature and heavy-ion irradiation on Mrozowski cracks (MC) and nuclear graphite crystallographic dimensions have been studied using in situ heating and in situ ion-irradiation in the transmission electron microscope (TEM). Electron transparent lamella of nuclear graphite, IG-110, was irradiated using a 2.8 MeV Au beam at an ion flux of 3.991 ×10 10 ion cm −2 s −1 for 70 min at 800 °C. Upon high-temperature irradiation, Mrozowski crack closure was studied quantitatively. The analysis showed linear, positive expansion of nuclear graphite which is significantly different from the dimensional changes previously reported for low-dose neutron irradiation of nuclear graphite in which the material undergoes negative to positive expansion via a turnaround radiation dose. The trend of the thermal expansion coefficient (CTE) of pristine IG-110 in this study is consistent with previous reports in the 100 °C–800 °C temperature region in which the dimensional change ranges from negative to positive values.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Experimentation in Exploring Photovoltaic Inverter Dynamics Under Different Irradiance Levels Through a Data-Driven Approach

As conventional direct connections of synchronous generators are being phased out, inverter-based resources (IBRs) with grid support functions are increasingly being integrated into power systems. This transition requires the development of accurate dynamic models for IBRs to predict how power systems will adapt to varying levels of IBRs penetration, establish grid code requirements, and ensure compliance. Here, this study introduces an active probing signal-based data-driven modeling technique to accurately derive the dynamics model of a smart photovoltaic inverter operating in Volt-Watt and Freq-Watt modes, in compliance with the IEEE 1547–2018 standard. The paper focuses on investigating how the dynamics of the PV inverter model respond to fluctuations in solar irradiance, utilizing real-time digital simulator experimentation. The experimental analysis demonstrates that the amplitude of dynamics fluctuates with changes in irradiance across both operational modes and confirms the active power’s dependence on irradiance levels. Furthermore, the nature of inverter dynamics varies distinctly between the different modes of activation. Critically, our findings indicate that dynamic models require DC-gain adjustments to accommodate contrasting irradiance levels, highlighting a negative gradient linear relationship between the DC-gain of each model and the irradiance.

14 SOLAR ENERGY↗

DOE OSTI2022

In situ ion irradiation of amorphous TiO 2 nanotubes

Understanding of structural and morphological evolution in nanomaterials is critical in tailoring their functionality for applications such as energy conversion and storage. Here, we examine irradiation effects on the morphology and structure of amorphous TiO 2 nanotubes in comparison with their crystalline counterpart, anatase TiO 2 nanotubes, using high-resolution transmission electron microscopy (TEM), in situ ion irradiation TEM, and molecular dynamics (MD) simulations. Anatase TiO 2 nanotubes exhibit morphological and structural stability under irradiation due to their high concentration of grain boundaries and surfaces as defect sinks. On the other hand, amorphous TiO 2 nanotubes undergo irradiation-induced crystallization, with some tubes remaining only partially crystallized. The partially crystalline tubes bend due to internal stresses associated with densification during crystallization as suggested by MD calculations. Furthermore, these results present a novel irradiation-based pathway for potentially tuning structure and morphology of energy storage materials.

36 MATERIALS SCIENCE↗

DOE OSTI2022

FFT-based model for irradiated aggregate microstructures in concrete

The concrete biological shield of light water reactors is exposed to neutron and gamma irradiation throughout its lifetime, which results in the long-term degradation of the concrete’s mechanical properties. Under neutron irradiation, the concrete’s aggregates are subjected to radiation-induced volumetric expansion (RIVE), which strongly depends on the mineral content of the aggregate and exhibits the largest expansion in silicate-bearing minerals. In this work, the authors used the fast Fourier transform-based code Microstructure-Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) in 2D to model the expansion of five different aggregates provided by the Japan Concrete Aging Management Program (JCAMP). Comparable rock specimens were irradiated at the JEEP-II test reactor. The model uses realistic aggregate microstructure reconstruction based on high-resolution characterization images. The model accounts for anisotropic RIVE, thermal expansion, and the associated initiation and propagation of damage. The RIVE models are calibrated based on expansion data in the literature. The authors assume that damage occurs exclusively at interfaces between the particles that compose an aggregate and that these interfaces also exhibit swelling. Using a micromechanical model, the evolution of Young’s modulus with RIVE is calculated for each aggregate and compared with Russian irradiation data. The modeled linear expansion agrees well with the experimentally measured expansion. Furthermore, the model also predicts that anisotropic RIVE and thermal expansion result in an earlier onset of damage with neutron fluence than in the isotropic case.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities

All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

An Evaluation of the Spectral Irradiance Data from the NSRDB

A recent effort by the National Renewable Energy Laboratory (NREL) has led to a new radiative transfer model, the Fast All-sky Radiation Model for Solar applications with Narrowband Irradiances on Tilted surfaces (FARMS-NIT), to efficiently compute spectral irradiances in the plane-of-array (POA). This model has been implemented in the National Solar Radiation Data Base (NSRDB) to provide photovoltaic (PV) resource in both narrowband and broadband wavelengths. This study conducts an evaluation of the spectral irradiances from the PV Resource dataset using surface-based observations at NREL’s Solar Radiation Research Laboratory (SRRL) and University of Oregon (UO). The results demonstrate that the PV resource has a generally good agreement with the long-term observations in both clear-sky and cloudy-sky conditions. Further research is needed to reduce the overestimation of visible irradiances in clear sky conditions and underestimation of near-infrared irradiances in cloudy-sky conditions.

14 SOLAR ENERGY↗

DOE OSTI2023

Challenges and successes in establishing a multipurpose fuels and materials irradiation capsule [Slides]

This presentation discusses the challenges and success related to designing and fabricating multipurpose capsules for irradiation experiments for irradiation within the Advanced Test Reactor (ATR) to be presented at tje DEVICE-MTR irradiation experiment design working group meeting hosted by the Japanese Atomic Energy Agency (JAEA). The information herein is limited to the design of the capsules only and no programmatic information from any project supporting the design of the multipurpose capsule is discussed. The intent of the design team was establish a capsule design that could accommodate multiple different irradiation project needs for the irradiation of nuclear fuel and materials of interest in the nuclear field. The main points of the presentation are as follows: use of tin as a molten thermal bond material was selected for the fueled capsule design and was found to be inadequate as it would corrode the stainless steel at high temperatures, this capsule design has high sensitivity to manufacturing of the gas gap and control of the gas mixture, and the fueled capsule has experienced welding challenges with the inner capsule material 800H Incoloy.

42 ENGINEERING↗

DOE OSTI2024

Summary Report on Ion Irradiation Study of Ceramic Coating on Suppressing FCCI

Advanced cladding is critical for advanced nuclear reactors with an enhanced performance in radiation tolerance and neutron transparency. Using advanced cladding will ensure the adequate thermal conductivity and mechanical stability of the cladding base material, corrosion resistance, high-temperature coolant compatibility of the cladding surface, and chemical stability in the cladding inner wall against fuel cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a modified surface, a clad base material, and a modified inner wall) promises to meet all these requirements. Initial research and development (R&D) regarding this innovative cladding seek to demonstrate the effectiveness of a thin ceramic coating in suppressing FCCI under ion irradiation to high dose. This United States (U.S.) Department of Energy (DOE)–Office of Nuclear Energy (NE) project report summarizes the results of recent ion irradiation studies of diffusion-couple samples with details of an ion irradiation experiment, the characterization of the interface microstructure of cerium (Ce)/titanium nitride (TiN)/oxide-dispersion strengthened (ODS), Ce/TiN/iron (Fe), Ce/ODS, and Ce/Fe samples irradiated with 80 million electron-volts (MeV) of xenon (Xe) ions to 100 displacements per atom (dpa) at 500?C using the Argonne Tandem Linac Accelerator System (ATLAS) facility at Argonne National Laboratory (ANL). The results of the ODS-substrate sample are encouraging and demonstrate the effectiveness of a thin ceramic coating on FCCI mitigation under irradiation. The results from an Fe-substrate sample revealed a complex microstructure, and the root cause is discussed.

36 MATERIALS SCIENCE↗

DOE OSTI2026

AGR-5/6/7 Irradiation Experiment Fission Product Mass Balance

This report presents the fission product mass balance for the AGR-5/6/7 TRISO fuel irradiation experiment. The fission product inventories deposited on capsule components outside of the fuel (e.g., stainless-steel shells, graphite holders, Grafoil disks, and associated hardware) were quantified as part of the post-irradiation examination (PIE) to assess the performance of this fuel. Comparisons were made between these inventories and depletion calculations, non-destructive measurements of fuel fission product inventories in the intact fuel compacts, prior AGR experiments, and results from among each of the five distinct AGR-5/6/7 capsules. The data served as estimates of the condensable fission products released from the fuel during irradiation. Excluding Capsule 1 (which experienced accidental damage during irradiation) and Capsule 3 (which was tested at very high irradiation temperatures of >1300°C), the results indicate that the AGR-5/6/7 fuel performed comparably to fuel from earlier AGR experiments. The mass balance results were also used to estimate the number of particles with in-pile SiC failures. Subject to the assumptions made in these estimates, and excluding Capsules 1 and 3, the in-pile SiC failure rates for AGR-5/6/7 are comparable to those observed in AGR-2.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Prediction of Solar Irradiance and Photovoltaic Solar Energy Product Based on Cloud Coverage Estimation Using Machine Learning Methods

Cloud cover estimation from images taken by sky-facing cameras can be an important input for analyzing current weather conditions and estimating photovoltaic power generation. The constant change in position, shape, and density of clouds, however, makes the development of a robust computational method for cloud cover estimation challenging. Accurately determining the edge of clouds and hence the separation between clouds and clear sky is difficult and often impossible. Toward determining cloud cover for estimating photovoltaic output, we propose using machine learning methods for cloud segmentation. We compare several methods including a classical regression model, deep learning methods, and boosting methods that combine results from the other machine learning models. To train each of the machine learning models with various sky conditions, we supplemented the existing Singapore whole sky imaging segmentation database with hazy and overcast images collected by a camera-equipped Waggle sensor node. We found that the U-Net architecture, one of the deep neural networks we utilized, segmented cloud pixels most accurately. However, the accuracy of segmenting cloud pixels did not guarantee high accuracy of estimating solar irradiance. We confirmed that the cloud cover ratio is directly related to solar irradiance. Additionally, we confirmed that solar irradiance and solar power output are closely related; hence, by predicting solar irradiance, we can estimate solar power output. This study demonstrates that sky-facing cameras with machine learning methods can be used to estimate solar power output. This ground-based approach provides an inexpensive way to understand solar irradiance and estimate production from photovoltaic solar facilities.

14 SOLAR ENERGY↗

NASA NTRS2012

Nucleobases and other Prebiotic Species from the Ultraviolet Irradiation of Pyrimidine in Astrophysical Ices

Nucleobases are N-heterocycles that are the informational subunits of DNA and RNA, and are divided into two families: pyrimidine bases (uracil, cytosine, and thymine) and purine bases (adenine and guanine). Nucleobases have been detected in meteorites and their extraterrestrial origin confirmed by isotope measurement. Although no Nheterocycles have ever been observed in the ISM, the positions of the 6.2-m interstellar emission features suggest a population of such molecules is likely to be present. In this work we study the formation of pyrimidine-based molecules, including nucleobases, as well as other species of prebiotic interest, from the ultraviolet (UV) irradiation of pyrimidine in combinations of H2O, NH3, CH3OH, and CH4 ices at low temperature, in order to simulate the astrophysical conditions under which prebiotic species may be formed in the interstellar medium and icy bodies of the Solar System. Experimental: Gas mixtures are prepared in a glass mixing line (background pressure approx. 10(exp -6)-10(exp -5) mbar). Relative proportions between mixture components are determined by their partial pressures. Gas mixtures are then deposited on an aluminum foil attached to a cold finger (15-20 K) and simultaneously irradiated with an H2 lamp emitting UV photons (Lyman and a continuum at approx.160 nm). After irradiation samples are warmed to room temperature, at which time the remaining residues are recovered to be analyzed with liquid and gas chromatographies. Results: These experiments showed that the UV irradiation of pyrimidine mixed in these ices at low temperature leads to the formation of several photoproducts derived from pyrimidine, including the nucleobases uracil and cytosine, as well as their precursors 4(3H)-pyrimidone and 4-aminopyrimidine (Fig. 1). Theoretical quantum calculations on the formation of 4(3H)-pyrimidone and uracil from the irradiation of pyrimidine in pure H2O ices are in agreement with their experimental formation pathways. In those residues, other species of prebiotic interest such as urea and the amino acids glycine and alanine could also be identified. However, no pyrimidine derivatives containing CH3 groups, including the third nucleobase thymine, could be identified, suggesting that the addition of methyl groups to pyrimidine is not an efficient process.

DNA and RNA↗

DOE OSTI2022

Effect of Water Chemistry on Crack Growth Rate in Neutron Irradiated X-750 and XM-19

The Advanced Test Reactor (ATR) National Science User Facility (NSUF) was used to study the effect of neutron damage on crack growth rates (CGR) under simulated boiling water reactor (BWR) water chemistry conditions using a fast neutron spectrum to accelerate the damage process. Nickel based Alloy X-750 and XM-19 were irradiated in the ATR center flux trap (CFT), with this experiment marking the first civilian project to utilize this position within the reactor. 0.4T-CT specimens were used to study crack growth rates for two sample fluences (~1.9 x 1020 n/cm2 and ~9.5 x 1020 n/cm2, E > 1MeV)) in both normal water chemistry (NWC) at 2.5 ppm O2 and hydrogen water chemistry (HWC) at ~80 ppb H2. CGR testing was conducted at a temperature of 288 °C. A decrease in CGR was observed for both alloys with the average ratio (CGR in NWC/CGR in HWC) being a factor of 281 for the XM-19 samples and a factor of 27.8 for the X-750 samples. Observed CGRs in this study showed little difference compared to non-irradiated samples (the non-irradiated XM-19 was 20% cold worked), suggesting limited sensitivity of these susceptible alloys to irradiation effects. Detailed results of CGR response to varying environmental electrochemical potential (ECP) and water purity will be presented in the context of crack growth mitigation in different water chemistries and the change in material behavior relative to non-irradiated samples of the material.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Design and Fabrication of a Temperature-Controlled Fueled Molten Salt Capsule Irradiation Experiment

Idaho National Laboratory (INL) is developing a Molten salt Research Temperature-controlled Irradiation (MRTI) experimental capsule for the Neutron Radiography (NRAD) reactor. The experiment is envisaged to be versatile and able to host a variety of salt-wall material combinations. The first test is designed to contain actinide-bearing chloride salt, which previously has not been subjected to a neutron field. The main objectives of the experiment are to produce irradiated salt samples for Post Irradiation Examination (PIE), to evaluate salt behavior under irradiation, and to test in-situ sensors/instrumentation. The final design for the experiment completed in 2022, and fabrication of the experiment is underway. The salt is double encapsulated providing insulation and redundancy in the case of leakage. A gas-gap between the inner capsule and the outer container enables tuning of thermal conduction to the NRAD coolant by altering its thickness and its gas-mixture. An inert gas mixture ratio in the gas-gap ensures the configurations reached a targeted average salt temperature of 600oC, while avoiding salt freezing or exceeding material limits. A resistive heater placed within a thermowell and controls the temperature of the experiment (both when the reactor is on or off). The selected configurations can contain over 10 cm3 of fuel-bearing chloride salt, which is sufficient for the purposes of the post-irradiation examination (PIE) and provides sufficient volume for two submerged thermocouple probes to monitor salt temperature. The capsule wall material is Inconel 625, and weld qualifications for the assembly have been developed. A glove-box laser weld technique was refined as part of this process to ensure a leak-tight assembly for the salt-bearing assembly and reduce the risks of impurities permeating into the salt. This work was supported through the INL Laboratory Directed Research & Development (LDRD) Program under DOE Idaho Operations Office Contract DE-AC07-05ID14517.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

ENYGF 2023 - Design of Temperature-controlled Fuel-salt Irradiation Experiment

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

High-Throughput Ion Irradiation and Microstructural Characterization of Multi-Principal-Element Alloys

Multi-principal-element alloys (MPEAs) have garnered interest in the nuclear community for their apparent resistance to microstructural damage under irradiation. While both experimentation and simulation have demonstrated reduced hardening and void swelling in MPEAs, the relationship between composition and irradiation response is still poorly understood since a rich experimental database for MPEAs does not yet exist. In this work, to accelerate the generation of experimental irradiation-response data for MPEAs, high-throughput synthesis, irradiation, and characterization techniques have been employed. Specifically, additive manufacturing has been used to produce arrays of various MPEAs (1-cm2 coupons), which have been subsequently irradiated using 4-MeV Ni2+ ions to a peak damage of 200 dpa at 500 °C. Hardening and void swelling have been assessed using nanoindentation and trenching via plasma focused ion beam (PFIB), respectively. Trends in the experimental data relating hardening and void swelling behavior to composition have been analyzed using different machine learning approaches.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Physical Properties of Hilti HIT-RE 500 V3 After Irradiation: Slant Shear, Tensile, Flexural, and Compressive

There are various applications in which epoxy adhesives may be used within hot-cell environments. These epoxies currently see limited use within hot cells since it is known that organics see degradation after small doses even as low as 50 kilogray (kGy). These losses in physical property strengths limit the use of these epoxies to non-structural applications. In modern construction, however, the use of epoxies to bond structural components such as rebar or threaded anchors is becoming increasingly prevalent. To employ these types of materials the Idaho National Laboratory (INL) has begun investigating the specific degradation properties of epoxies for use in structural applications in environments where radiation is present. Hilti HIT-RE 500 V3 is used extensively in new construction and renovation projects across the world. The INL has conducted various tests on this epoxy for low dose environments. These tests include tensile, flexural, compression, and slant shear tests following the American Society for Testing Material International (ASTM) standards D638, D790, D695, and C882/882M respectively. Two sample sets for each test were created where the baseline specimens were not irradiated, and another set was irradiated to an accumulated dose of approximately 50-60 kGy. All the samples were destructively tested and analyzed focusing on the change in physical properties, examples of these tests are shown in Figures 1-4. Tensile testing results indicate a 0.9% loss in maximum load capacity, a 0.5% loss in tensile stress at maximum load, a 1.4% loss in tensile strain at break, and a 15% decrease in modulus of elasticity. Flexural testing results display an increase in all the following: 12% in strain at maximum load, 10% in maximum load capacity, 5% in stress at maximum load, and 3% in young’s modulus. The compressive physical properties after irradiation display a 1% increase in maximum load, a 3% decrease in stress at yield, and a 5% decrease in modulus of elasticity. The maximum compressive stress saw no measurable change after irradiation. The slant shear specimens after irradiation displayed a 2% increase in maximum load capacity and maximum compressive stress. Though a structural analysis would have to be completed as is normal for building design the increases and decreases in the measured physical properties indicate that this material may be used for structural applications in low dose hot-cell environments without adhesive failure.

36 MATERIALS SCIENCE↗

DOE OSTI2023

A Fast Reactor Irradiation Experiment Design in the ATR

Modern modeling techniques were used to investigate a proposed method for fast neutron irradiations in an existing thermal-spectrum reactor, the Advanced Test Reactor (ATR). This method builds upon pre-existing ideas, where fast flux is increased by surrounding the specimens with fissionable “booster fuel” but diverges from historical approaches by using an already developed fuel element design used in the Belgian Reactor 2 (BR2) as the booster fuel while leveraging modern 3-D modeling and simulation techniques. Design evaluations and neutronics simulations were performed to evaluate the performance of a BR2 fuel element irradiated in an ATR flux trap with test pins in the central channel of the BR2 fuel element. These efforts have yielded promising results. Adding a BR2 fuel element in the northeast (NE) flux trap of ATR was predicted to result in a 150% increase to the incident fast neutron flux with a fast (>0.1 MeV) to thermal (<0.625 eV) neutron flux ratio ranging from approximately 50 to 150, dependent on the material used for thermal neutron filtering and volume of moderator within the central channel of the BR2 element. The predicted annual fast neutron fluence (>0.1 MeV) ranges from 7.9 × 1021 to 9.1 × 1021 n/cm2. Given the relatively large fast to thermal neutron flux ratio, the calculated radial power profiles within 4.3 mm outer diameter U10Zr fueled specimens irradiated within the BR2 booster fuel element are adequate representations of those within fast neutron reactors. The predicted radial power profiles are not flat, but they are more prototypic than those seen in advanced fuels tests which began in ATR in 2003. Another distinct advantage of this experiment design is that full-scale test pins can be irradiated to augment the ongoing series of reduced scale advanced fuels tests. The proposed experiment design irradiated within a BR2 fuel element in a flux trap of ATR offers an improved alternative to the current testing of advanced reactor fuels in ATR. Selected results from this design evaluation are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗