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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

A Practical Irradiance Model for Bifacial PV Modules

A model, suitable for a row or multiple rows of photovoltaic (PV) modules, is presented for estimating the backside irradiance for bifacial PV modules. The model, which includes the effects of shading by the PV rows, is based on the use of configuration factors to determine the fraction of a source of irradiance that is received by the backside of the PV module. Backside irradiances are modeled along the sloped height of the PV module, but assumed not to vary along the length of the PV row. The backside irradiances are corrected for angle-ofincidence losses and may be added to the front side irradiance to determine the total irradiance resource for the PV cell. Model results are compared with the measured backside irradiances for NREL and Sandia PV systems, and with results when using ray tracing software.

14 SOLAR ENERGY↗
DOE OSTI2014

Magnetization measurements and XMCD studies on ion irradiated iron oxide and core-shell iron/iron-oxide nanomaterials

Magnetite (Fe3O4) and core-shell iron/iron-oxide (Fe/Fe3O4) nanomaterials prepared by a cluster deposition system were irradiated with 5.5 MeV Si2+ ions and the structures determined by x-ray diffraction as consisting of 100% magnetite and 36/64 wt% Fe/FeO, respectively. However, x-ray magnetic circular dichroism (XMCD) indicates similar surfaces in the two samples, slightly oxidized and so having more Fe3+ than the expected magnetite structure, with XMCD intensity much lower for the irradiated core-shell samples indicating weaker magnetism. X-ray absorption spectroscopy (XAS) data lack the signature for FeO, but the irradiated core-shell system consists of Fe-cores with ~13 nm of separating oxide crystallite, so it is likely that FeO exists deeper than the probe depth of the XAS (~5 nm). Exchange bias (Hex) for both samples becomes increasingly negative as temperature is lowered, but the irradiated Fe3O4 sample shows greater sensitivity of cooling field on Hex. Loop asymmetries and Hex sensitivities of the irradiated Fe3O4 sample are due to interfaces and interactions between grains which were not present in samples before irradiation as well as surface oxidation. Asymmetries in the hysteresis curves of the irradiated core/shell sample are related to the reversal mechanism of the antiferromagnetic FeO and possibly some near surface oxidation.

core-shell, magnetite, irradiation, XMCD↗
DOE OSTI2014

Geometric and dosimetric accuracy of dynamic tumor-tracking conformal arc irradiation with a gimbaled x-ray head

Purpose: The Vero4DRT system has the capability for dynamic tumor-tracking (DTT) stereotactic irradiation using a unique gimbaled x-ray head. The purposes of this study were to develop DTT conformal arc irradiation and to estimate its geometric and dosimetric accuracy. Methods: The gimbaled x-ray head, supported on an O-ring gantry, was moved in the pan and tilt directions during O-ring gantry rotation. To evaluate the mechanical accuracy, the gimbaled x-ray head was moved during the gantry rotating according to input command signals without a target tracking, and a machine log analysis was performed. The difference between a command and a measured position was calculated as mechanical error. To evaluate beam-positioning accuracy, a moving phantom, which had a steel ball fixed at the center, was driven based on a sinusoidal wave (amplitude [A]: 20 mm, time period [T]: 4 s), a patient breathing motion with a regular pattern (A: 16 mm, average T: 4.5 s), and an irregular pattern (A: 7.2–23.0 mm, T: 2.3–10.0 s), and irradiated with DTT during gantry rotation. The beam-positioning error was evaluated as the difference between the centroid position of the irradiated field and the steel ball on images from an electronic portal imaging device. For dosimetric accuracy, dose distributions in static and moving targets were evaluated with DTT conformal arc irradiation. Results: The root mean squares (RMSs) of the mechanical error were up to 0.11 mm for pan motion and up to 0.14 mm for tilt motion. The RMSs of the beam-positioning error were within 0.23 mm for each pattern. The dose distribution in a moving phantom with tracking arc irradiation was in good agreement with that in static conditions. Conclusions: The gimbal positional accuracy was not degraded by gantry motion. As in the case of a fixed port, the Vero4DRT system showed adequate accuracy of DTT conformal arc irradiation.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2015

WE-EF-BRA-11: Precision Partial-Tumor Irradiation of Dorsal Rodent Mammary Tumors

Purpose: To introduce a pre-clinical treatment technique on a micro-irradiator to treat specific volumes of dorsal mammary tumors in BALB/c mice while sparing lungs and spine. This technique facilitates pre-clinical investigation of tumor response to sub-optimal radiation treatments in which a portion of the tumor is unirradiated, known as a “marginal miss”. In-vitro data suggests that partial tumor radiations trigger a more aggressive phenotype in non-irradiated, regional tumor cells via bystander effects. As the lung tissue is spared, the impact of marginal miss on the development of pulmonary metastasis may be assessed. Methods: End to end test was performed on three BALB/c mice as proof of concept for larger studies. 1Gy was delivered on the micro-irradiator employing previously unexplored lateral parallel-opposed diamond and/or triangle-shaped beams. The margins of the treatment beam were defined using a combination of tumor palpation, barium fiducial markers, and real-time fluoroscopic images. The dose distribution was independently verified with kilovoltage beam Monte Carlo dose calculations with 7% statistical uncertainty and double exposure images. As a final step, the technique was used in a larger pre-clinical study (15Gy, 36 BALB/c mice) and lung metastasis in response to tumor irradiation of 100%, 50% and 0% was quantified. Results: For the Monte-Carlo dose calculations, the dose volume histograms established a maximum dose within the un-irradiated and radiated portions of the mammary tumor of 0.3Gy and 1.5Gy respectively, with a sharp gradient at the boundary. 100% of the lung volume received less than 0.5Gy. This technique proved suitable for a pre-clinical marginal miss study with 50% more lung metastases in partially-radiated mouse models compared to completely. Conclusion: We have developed a novel treatment technique for partial or full irradiation of dorsal mammary tumors incorporating lung sparing.The technique will be useful for exploring the phenomena of aggressive tumor phenotype that may arise from miss-irradiation.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Transmission electron microscopy examinations of metal-oxide interface of zirconium-based alloys irradiated in Halden reactor-IFA-638 - 2016-0039

This paper provides the results of investigations by transmission electron microscopy (TEM) on the selected materials from in-reactor oxidation tests in the Halden test reactor (Reference No. IFA-638) from 1998 to 2006. The objective of the IFA-638 test was to study the corrosion behavior of modern zirconium-based claddings to high burnup in pressurized water reactor water chemistry and thermal hydraulic conditions. The aim of this paper is to report on the microstructure of selected materials (ZIRLO{sup R}, E635, and Alloy A) after the irradiation to different burnup levels to determine the modifications induced by irradiation and to correlate results to their oxidation behavior. The TEM examinations revealed the nature of secondary phase particles (SPPs) and their modification under irradiation. Four types of SPPs were observed, namely β-niobium precipitates, Zr{sub 0.5}Nb{sub 0.3}Fe{sub 0.2} (mainly in the ZIRLO alloy), Zr(Fe,Nb){sub 2} (in E635), and (Cr,Fe){sub 2}Zr,Nb with varying niobium content (present in Alloy A: Zr-0.58Sn-0.31Nb-0.36Fe-0.26Cr). TEM observations showed that all three materials contained still several precipitates after irradiation and in the case of the ZIRLO alloy even after high burnups. Furthermore, the analysis of the metal side of the interface and its comparison with the oxide side led to the conclusion that all types of precipitates dissolved to some extent under irradiation and that their alloying element content decreased. The dissolution was intensified in the oxide. However, a more detailed examination showed that the β-niobium precipitates dissolved at a slower rate, or knowing that their composition was much richer in niobium, the time needed for the precipitates to become fully depleted from niobium was longer. Regarding the amorphization under irradiation, the b-niobium- and chromium-containing precipitates did not amorphize in the metal part of the interface. This was not the case for the other types of precipitates. Furthermore, these two types of SPP both showed delayed oxidation and due to this behavior the typical crack above the SPP in the oxide was also observed. These results are discussed to gain an improved understanding of the oxidation behavior of materials studied as a function of irradiation and residence time. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effect of hydrogen on irradiation creep and growth for ZIRLO alloy and Zr-1.0Nb - 2016-0058

The impact of hydrogen on the irradiation growth and creep of stress-relief annealed (SRA) ZIRLO{sup R} alloy and recrystallized annealed (RXA) Zr-1.0Nb cladding tubes is evaluated in this paper. The samples were charged with hydrogen in the range of approximately 160-720 ppm using the gaseous method. Biaxial in-reactor creep tests were performed after Cycle 1, Cycle 2, Cycle 3, and Cycle 4, on both as received and pre-charged hydrogen cladding tubes. Outside diameter and axial length measurements were performed on the samples. The results showed that hydrogen had no effect on the axial irradiation creep but a relatively large effect on the axial irradiation growth. Increasing hydrogen decreased the axial irradiation growth in RXA Zr-1.0 Nb, which was opposite from the behavior of SRA ZIRLO cladding. This unique hydrogen effect on the irradiation axial growth of RXA Zr-1.0Nb could be due to the different intergranular stress resulting from the fabrication process or the absence of alloying elements, including tin. The total axial strain of both Zr-1.0Nb and SRA ZIRLO cladding increased with increasing fast fluence, and Zr-1.0Nb increased at a faster rate relative to SRA ZIRLO cladding. In the diameter direction, hydrogen had a minimal effect on the total diameter strain and the diameter irradiation creep strain for both SRA ZIRLO samples and the RXA Zr-1.0Nb sample. This finding from in-reactor test is contrary to the out-reactor tests results from the literature that have shown that hydrogen significantly decreases thermal creep. The total diameter strain and diameter irradiation creep behavior for the SRA ZIRLO samples and RXA Zr-1.0Nb were similar. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2014

AGR-2 Irradiation Test Final As-Run Report

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions

As a first step toward full scale utilization of the Advanced Test Reactor (ATR) and associated post irradiation examination (PIE) equipment at INL, the NRC and INL staffs have formulated a test program that is designed to help to establish INL and the Nuclear Science User Facilities (NSUF) as a viable destination for irradiation and PIE of reactor structural materials. This initial research program utilized two materials (304 Stainless Steel weld Heat Affected Zone (HAZ) and sensitized 304L stainless steel) that have previously been irradiated at the Halden reactor in Norway and tested at Argonne National Laboratory (ANL) so that test results may be compared, thereby establishing a measure of comparability between Halden and ATR irradiations. A secondary objective of this test program was to characterize the quality of data produced using INL’s newly constructed irradiation assisted stress corrosion cracking (IASCC) test cells. The specimens were be tested in typical boiling water reactor (BWR) conditions to measure stress corrosion cracking (SCC) and fracture toughness of un-irradiated specimens and then IASCC and fracture toughness of specimens irradiated to a fluence equal to approximately 1.0 X 10 21 n/cm 2 (E > 1 MeV).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Nanoscale Consequences of Irradiation Investigated by RAD-AFM (Final Report)

To assess the feasibility of measuring material property changes induced by ion irradiation one pristine and three legacy HT-9 samples were sectioned and polished. The polish protocol was first optimized using a pristine unirradiated HT-9 sample before polishing the ion irradiated samples as the spot size was small and repeated polishing could remove the ion irradiated region entirely. AFM tapping mode was used to assess the surface quality after each polish. AFM based hardness mapping was optimized for new diamond tips specifically for use with the HT-9 samples to compensate for differences in grain size, precipitates, and heterogeneity of hardness relative to past samples. In addition, a hardness calibration protocol was developed using NIST copper, nickel, and steel reference materials. Co-located SEM-EDS and AFM analysis was used to verify the ion irradiated edge using the silver/tungsten coating as a fiducial mark. Once benchmarked, the hardness across the ion irradiated cross section, a comparison of bulk hardness according to neutron irradiation, and the hardness of the ion irradiated surface edge was measured.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A Novel Microshear Geometry for Exploring the Influence of Void Swelling on the Mechanical Properties Induced by MeV Heavy Ion Irradiation

Small disks are often the specimen of choice for exposure in nuclear reactor environments, and this geometry invariably limits the types of mechanical testing that can be performed on the specimen. Recently, shear punch testing has been utilized to evaluate changes arising from neutron irradiation in test reactor environments on these small disk specimens. As part of a broader effort to link accelerated testing using ion irradiation and conventional neutron irradiation techniques, a novel microshear specimen geometry was developed for use with heavy-ion irradiated specimens. The technique was demonstrated in pure Cu irradiated to 11 and 110 peak dpa with 10 MeV Cu ions. At 11 peak dpa, the Cu specimen had a high density of small voids in the irradiated region, while at 110 peak dpa, larger voids with an average void swelling of ~20% were observed. Micropillar and microshear specimens both exhibited hardening at 11 dpa, followed by softening at 110 dpa. The close alignment of the new microshear technique and more conventional micropillar testing, and the fact that both follow intuition, is a good first step towards applying microshear testing to a wider range of irradiated materials.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Thermomechanical Properties of Neutron Irradiated Al 3 Hf-Al Thermal Neutron Absorber Materials

A thermal neutron absorber material composed of Al 3 Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al 3 Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4T m . Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al 3 Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al 3 Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al 3 Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of –6.3% for the annealed irradiated 36.5 vol% specimen.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Investigation of High-Energy Ion-Irradiated MA957 Using Synchrotron Radiation under In-Situ Tension

In this study, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region ~7.5 μm in depth; the peak damage (~40 dpa) was estimated to be at ~7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing of the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials’ radiation damage in a dynamic manner.

36 MATERIALS SCIENCE↗
NASA NTRS2017

Optical Properties of Ar Ions Irradiated Nanocrystalline ZrC and ZrN Thin Films

Thin nanocrystalline ZrC and ZrN films (less than 400 nanometers), grown on (100) Si substrates at a substrate temperature of 500 degrees Centigrade by the pulsed laser deposition (PLD) technique, were irradiated by 800 kiloelectronvolts Ar ion irradiation with fluences from 1 times 10(sup 14) atoms per square centimeter up to 2 times 10(sup 15) atoms per square centimeter. Optical reflectance data, acquired from as-deposited and irradiated films, in the range of 500-50000 per centimeter (0.06–6 electronvolts), was used to assess the effect of irradiation on the optical and electronic properties. Both in ZrC and ZrN films we observed that irradiation affects the optical properties of the films mostly at low frequencies, which is dominated by the free carriers response. In both materials, we found a significant reduction in the free carriers scattering rate, i.e. possible increase in mobility, at higher irradiation flux. This is consistent with our previous findings that irradiation affects the crystallite size and the micro-strain, but it does not induce major structural changes.

ZrN↗
NASA NTRS2018

Late Effects of Heavy-Ion Irradiation on Ex Vivo Osteoblastogenesis and Cancellous Bone Microarchitecture

Prolonged spaceflight causes degeneration of skeletal tissue with incomplete recovery even after return to Earth. We hypothesize that heavy-ion irradiation, a component of Galactic Cosmic Radiation, damages osteoblast progenitors and may contribute to bone loss during long duration space travel beyond the protection of the Earth's magnetosphere. Male, 16 week-old C57BL6/J mice were exposed to high-LET (56-Fe, 600MeV) radiation using either low (5 or 10cGy) or high (50 or 200cGy) doses at the NASA Space Radiation Lab and were euthanized 3-4, 7, or 35 days later. Bone structure was quantified by microcomputed tomography (6.8 μm pixel size) and marrow cell redox assessed using membrane permeable, free radical-sensitive fluorogenic dyes. To assess osteoblastogenesis, adherent marrow cells were cultured ex vivo, then mineralized nodule formation quantified by imaging and gene expression analyzed by RT-PCR. Interestingly, 3-4 days post-exposure, fluorogenic dyes that reflect cytoplasmic generation of reactive nitrogen/oxygen species (DAF-FM Diacetate or CM-H2DCFDA) revealed irradiation (50cGy) reduced free radical generation (20-45%) compared to sham-irradiated controls. Alternatively, use of a dye showing relative specificity for mitochondrial superoxide generation (MitoSOX) revealed an 88% increase compared to controls. One week after exposure, reactive oxygen/nitrogen levels remained lower (24%) relative to sham-irradiated controls. After one month, high dose irradiation (200 cGy) caused an 86% decrement in ex vivo nodule formation and a 16-31% decrement in bone volume to total volume and trabecular number (50, 200cGy) compared to controls. High dose irradiation (200cGy) up-regulated expression of a late osteoblast marker (BGLAP) and select genes related to oxidative metabolism (Catalase) and DNA damage repair (Gadd45). In contrast, lower doses (5, 10cGy) did not affect bone structure or ex vivo nodule formation, but did down-regulate iNOS by 0.54-0.58 fold. Thus, both low- and high-doses of heavy-ion irradiation cause time-dependent, adaptive changes in redox state within marrow cells but only high doses (50, 200cGy) inhibit osteoblastogenesis and cause cancellous bone loss. We conclude space radiation has the potential to cause persistent damage to bone marrow-derived stem and progenitor cells for osteoblasts despite adaptive changes in cellular redox state.

LATE EFFECTS OF HEAVY-ION IRRADIATION↗
DOE OSTI2015

Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 10 11 n/cm 2 /s (>1 MeV) to fluences from 0.5 to 3.4 10 19 n/cm 2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 10 13 n/cm 2 /s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 10 13 n/cm 2 /s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 10 19 n/cm 2 . The irradiation-induced shifts of the Master Curve reference temperatures, ΔT 0 , for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T 0 , 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT 0 , were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions. These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules also increases gap size uncertainty. This study focuses on analysis of modeling assumptions and uncertainty sources to evaluate their impacts on the gap size uncertainty.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A Practical Irradiance Model for Bifacial PV Modules: Preprint

A model, suitable for a row or multiple rows of photovoltaic (PV) modules, is presented for estimating the backside irradiance for bifacial PV modules. The model, which includes the effects of shading by the PV rows, is based on the use of configuration factors (CFs) to determine the fraction of a source of irradiance that is received by the backside of the PV module. Backside irradiances are modeled along the sloped height of the PV module, but assumed not to vary along the length of the PV row. The backside irradiances are corrected for angle-of-incidence losses and may be added to the front side irradiance to determine the total irradiance resource for the PV cell. Model results are compared with the measured backside irradiances for NREL and Sandia PV systems, and with results when using the RADIANCE ray tracing program.

14 SOLAR ENERGY↗