Hollow organic capsules assemble into cellular semiconductors
Explore the source record for details and available documents.
99 GENERAL AND MISCELLANEOUS↗
SEARCH · Engineering Papers
Results for “capsule”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Beryllium capsule implosions at a case-to-capsule ratio of 3.7 on the National Ignition Facility
Beryllium is a candidate ablator material for indirect-drive inertial confinement fusion experiments, motivated by its high mass ablation rate, which is advantageous for implosion coupling efficiency and stabilization of the ablation-front instability growth. In this paper, we present new data on the shock propagation, in-flight shape, and hot spot self-emission shape from gas-filled capsules that demonstrate the feasibility of predictable, symmetric, controllable beryllium implosions at a case-to-capsule ratio of 3.7. The implosions are round (Legendre mode 2 amplitude ≲5%) at an inner beam power and the energy fraction of 26%–28% of the total, indicating that larger beryllium capsules could be driven symmetrically using the National Ignition Facility.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Science Capsule: Capturing the Data Life Cycle (Science Capsule) v0.1.0
Science Capsule captures the processing and data life cycle across machines, organizations, people, and science domains. Our approach combines user engagement methods from social sciences with innovations in workflow and data management.
Essiari, Abdelilah↗
Destructive PIE and Safety Testing of Six AGR-2 UO 2 Capsule 3 Compacts
The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2) was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from June 2010 to October 2013 (Collin 2014). The fuel compacts in this experiment held either tristructural isotropic (TRISO)-coated spherical kernels of uranium oxide (UO2) or TRISO-coated kernels containing both uranium carbide and uranium oxide phases (UCO). There were six separately monitored and controlled capsules in the AGR-2 test train. Capsule 3 held twelve compacts containing UO2-TRISO particles fabricated by BWX Technologies Nuclear Operations Group. The AGR-2 TRISO particles were fabricated in a pilot-scale fluidized-bed chemical vapor deposition (FB-CVD) furnace with a coating chamber inner diameter of 150 mm (Phillips, Barnes, and Hunn 2010), which was a change from the first irradiation experiment (AGR-1) particles that had been coated in a lab-scale FB-CVD coating system with a chamber inner diameter of 50 mm (Lowden 2006). The TRISO particles were overcoated with resinated graphite flake at Oak Ridge National Laboratory (ORNL), and the overcoated particles were pressed into one-inch-long, half-inch-diameter cylinders (Hunn, Montgomery, and Pappano 2010). Each cylindrical compact held an average of 1,543 TRISO particles with 9.6% enriched UO2 kernels that had a nominal diameter of 500 μm (Hunn, Savage, and Silva 2012). Capsule 3 compacts were irradiated to average calculated burnups of 9.01–10.69% fissions per initial metal atom (FIMA), and the average calculated fluences of fast neutrons with energies E n > 0.18 MeV were 3.05–3.53×10 25 n/m 2 (Sterbentz 2014). The calculated time-average, volume-average Capsule 3 compact temperatures were 996–1,062°C. However, Capsule 3 compact temperatures varied several hundred degrees across each compact, and the timeaverage minimum (TA min ) and time-average maximum (TA max ) temperatures were between 889–999°C and 1,072–1,105°C, respectively (Hawkes 2014). After irradiation, the AGR-2 test train was transferred from ATR to the INL Materials and Fuels Complex for inspection and disassembly (Ploger, Demkowicz, and Harp 2015). The initial inspection included dimensional metrology of the compacts and graphite fuel holders. Like all the AGR-2 compacts, the compacts in Capsule 3 shrank slightly during irradiation, as expected, with an average length reduction of 1.07–1.24% and an average diameter reduction of 0.13–0.41%. Post-irradiation examination (PIE) of the capsule components was completed to measure select fission products ( 90 Sr, 110m Ag, 134 Cs, 137 Cs, 144 Ce, and 154 Eu) outside the compacts (Stempien and Demkowicz 2020). This involved gamma counting of the graphite and graphoil spacers at the top and bottom of each capsule, acid leaching for radiochemical analysis of fission products on the metallic capsule components, and burn-leach analysis of the graphite holders. The total amount of 110mAg measured on the Capsule 3 components was 13% of the calculated capsule inventory. This was significantly lower than the amount of 110m Ag measured on the three UCO capsule components, which ranged from 32–70%. The lower 110m Ag release in Capsule 3 was likely due to lower peak temperatures compared with the UCO fuel capsules (Hawkes 2014). Measured inventories of the other select fission products were also lower in Capsule 3.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Incomplete tumor capsule on preoperative imaging reveals microvascular invasion in hepatocellular carcinoma: a systematic review and meta-analysis
Purpose: Microvascular invasion (MVI), which is difficult to diagnose before surgery, is a major factor affecting postoperative recurrence in patients with hepatocellular carcinoma (HCC). The relationship between the radiological tumor capsule and MVI is controversial. This study aimed to evaluate the association between the tumor capsule and MVI. Methods: We searched Medline (by PubMed) and Embase (by OvidSP). Two review authors independently screened titles and abstracts, selected studies about MVI prediction with radiologic tumor capsule and studies with enough data for extracted, assessed the methodological quality and collected data. Summary results were presented as the diagnostic odds ratio (DOR), sensitivity, specificity, and 95% confidence interval. Results: Fifteen studies with 2038 patients were included; fourteen studies, including 1331 patients, with no significant heterogeneity indicated no relationship between absent tumor capsule and MVI [DOR = 0.90 (0.64, 1.26)]. Six studies, including 541 patients, with no significant heterogeneity showed incomplete capsule could be used to predict MVI of HCC preoperatively [DOR = 1.85 (1.13, 3.04)]. The overall sensitivity and specificity estimate were 0.50 (0.37, 0.64) and 0.64 (0.53, 0.74), respectively. Eight studies, including 1349 patients, with highly significant heterogeneity revealed that complete capsule could be a protective factor for MVI [DOR = 1.97 (1.01, 3.86)]. Conclusions: For MVI of HCC, incomplete tumor capsule is a risk factor, while a complete tumor capsule might be a protective factor. However, absent capsule doesn’t show significant relationship with MVI. This might be due to combination of the risk and protective effects of present capsule in MVI.
62 RADIOLOGY AND NUCLEAR MEDICINE↗
Review of hydro-instability experiments with alternate capsule supports in indirect-drive implosions on the National Ignition Facility
Hydrodynamic instability growth of capsule support membranes (or “tents”) has been recognized as one of the major contributors to the performance degradation in high-compression plastic capsule implosions at the National Ignition Facility (NIF). The capsules were supported by tents because the nominal 10-μm diameter fill tubes were not strong enough to support capsules by themselves in indirect-drive implosions on NIF. After it was recognized that the tents had a significant impact of implosion's stability, new alternative support methods were investigated. While some of these methods completely eliminated tent, other concepts still used tents, but concentrated on mitigating their impact. The tent-less methods included “fishing pole” reinforced fill tubes, cantilevered fill tubes, and thin-wire “tetra cage” supports. In the “fishing pole” concept, a 10-μm fill tube was inserted inside 30-μm fill tube for extra support with the connection point located 300 μm away from the capsule surface. The cantilevered fill tubes were supported by 12-μm thick SiC rods, offset by up to 300 μm from the capsule surfaces. In the “tetra-cage” concept, 2.5-μm thick wires (carbon nanotube yarns) were used to support a capsule. Other concepts used “polar tents” and a “foam-shell” to mitigate the effects of the tents. The “polar tents” had significantly reduced contact area between the tents and the capsule compared to the nominal tents. In the “foam-shell” concept, a 200-μm thick, 30 mg/cc SiO 2 foam layer was used to offset the tents away from the capsule surface in an attempt to mitigate their effects. These concepts were investigated in x-ray radiography experiments and compared with perturbations from standard tent support. The measured perturbations in the “fishing pole,” cantilevered fill tube, and “tetra-cage” concepts compared favorably with (were smaller than) nominal tent perturbations and were recommended for further testing for feasibility in layered DT implosions. The “polar tents” were tested in layered DT implosions with a relatively-stable “high-foot” drive showing an improvement in neutron yield in one experiment compared to companion implosions with nominal tents. Furthermore, this article reviews and summarizes recent experiments on these alternate capsule support concepts. In addition, the concept of magnetic levitation is also discussed.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Quantum Levitation of Fuel Capsules for Inertial Confinement Fusion
A major current challenge of the inertial confinement fusion (ICF) campaign is implosion perturbation due to fuel capsule support. Our current approach to hold ICF fuel capsules inside hohlraums is to sandwich a capsule between two thin polymeric films, referred to as tents. In addition to tents, variants of thin fibers have been used in some target designs to hold the capsule. Both tents and fibers cause strong implosion perturbations. An ideal solution to this problem is to hold the capsule inside the hohlraum by levitation, eliminating the perturbation-causing mechanical support. The goal of this project was to demonstrate a path to such an ideal capsule support solution based on quantum (superconducting) levitation. In our approach, the outer surface of the capsule is coated with a thin (about 100 nm) superconducting layer of MgB2, which is currently the most promising superconducting material candidate for this application. Magnetic coils outside the hohlraum are used to levitate and position the capsule in the center of the hohlraum. The scope of this project was limited to the first and arguably most critical step toward quantum levitation of ICF fuel capsules: the demonstration of thin-film superconductor coatings through a better understanding of the physics of defect- and interface-dominated superconducting behavior of ultrathin films. Specific objectives were (i) to develop a method for deposition of smooth and uniform thin films of MgB2 on ICF-scale spherical capsules; (ii) to better understand effects of film thickness, interfaces, and lattice defects on superconducting properties of MgB2; (iii) to perform cryogenic levitation experiments; and (iv) to develop a levitation model. All these objectives were met.
36 MATERIALS SCIENCE↗
Report on Retrieval of the Reactor Pressure Vessel A-60 Surveillance Capsule from Palisades Nuclear Generating Station
Located on the shores of Lake Michigan, the Palisades Nuclear Generating Station (PNGS) was a nuclear power plant that operated in Covert Township, Michigan. The plant had a single pressurized water reactor that produced electricity for the region. The PNGS was shut down in 2022 after more than four decades of service. The PNGS included in its surveillance program a surveillance capsule, designated A-60, containing specimens of a weld metal with nickel content of about 1.36 wt% and copper content of about 0.20 wt%. The capsule was removed from its surveillance position in the early 1995 and has been resident in the spent fuel pool since that time. This capsule was irradiated to a fluence of 1.87×10 20 n/cm 2 (E> 1MeV) that is equivalent for more than 120 effective full power year (EFPY) for the US reactor pressure vessel (RPV) fleet. The material is also of special interest because of its very high nickel content and potential for development of NiMnSi (nickel-manganese-silicon) precipitates. Combination of very high fluence and very high Ni and Cu content makes the material in this capsule of the great interest as benchmark for currently developing embrittlement trend curves (ETC) aiming to predict embrittlement at high fluences. Thus, several years ago the Light Water Reactor Sustainability Program (LWRSP) initiated negotiations with Entergy to harvest this high fluence capsule since it did not present any regulatory interest for PNGS but could play a very important role for LWRSP efforts for developing ETC for very long-term operation. The negotiations resumed once PNGS ownership moved to the Holtec International in June 2022 which was very supportive to the LWRSP efforts to harvest the A-60 capsule from the spent fuel pool. As a result, the contract was placed with the Westinghouse Electric Company (WEC) to come to PNGS site, retrieve the A-60 capsule, bring it to WEC Churchill hot cell facility, open the capsule and send all surveillance specimens in the capsule to ORNL for future characterization. These specimens from the PNGS A-60 capsule have arrived to ORNL in July 2023 and this report summarizes activities to retrieve and deliver specimens to ORNL site for future evaluation.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR
The High Flux Isotope Reactor (HFIR) is an ideal tool for materials irradiation testing because of its intense steady-state neutron flux. Many programs take advantage of HFIR’s central flux trap for irradiation experiments using capsules, also known as rabbits, to support advanced materials development and reactor design. The facility that makes up the HFIR flux trap has recently undergone a design change that increases the HFIR primary coolant volumetric flow rate by removing restrictions in the system. As a result, the usable cross-sectional area within the facility increased, opening the door to increase the cross-sectional area of the rabbit capsules that fill the facility. This report documents a new large-diameter rabbit housing that increases the usable volume within the rabbit capsule by 22.6%. However, challenges arise with increasing the capsule size, such as establishing a new maximum capsule operating pressure and determining the thermal-hydraulic characteristics. This report addresses those challenges with previously adopted HFIR safety methods. The rupture pressure of the rabbit housings is demonstrated while verifying that capsule swelling during and after rupture will not block coolant flow. Then, a safety factor is applied to ascertain an administrative operating pressure. Additionally, the thermal-hydraulic performance of the HFIR facility filled with large-diameter rabbit capsules is shown to not violate previously determined safety criteria. Next, heat transfer coefficients are determined for use in design calculations. Furthermore, this report gives an example of internal configurations for the new, larger rabbit capsules that use relevant boiling water reactor (BWR) cladding geometry. Finally, this report documents an example thermal design performance for a rabbit capsule containing six gauge-curved tensile tube specimens. The thermal performance gives predicted temperature distributions within the capsule and shows the expected temperature of the passive thermometers for post-irradiation temperature comparisons.
99 GENERAL AND MISCELLANEOUS↗
Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR
The High Flux Isotope Reactor (HFIR) is an ideal tool for materials irradiation testing because of its intense steady-state neutron flux. Many programs take advantage of HFIR’s central flux trap for irradiation experiments using capsules, also known as rabbits, to support advanced materials development and reactor design. The facility that makes up the HFIR flux trap has recently undergone a design change that increases the HFIR primary coolant volumetric flow rate by removing restrictions in the system. As a result, the usable cross-sectional area within the facility increased, opening the door to increase the cross-sectional area of the rabbit capsules that fill the facility. This report documents a new large-diameter rabbit housing that increases the usable volume within the rabbit capsule by 22.6%. However, challenges arise with increasing the capsule size, such as establishing a new maximum capsule operating pressure and determining the thermal-hydraulic characteristics. This report addresses those challenges with previously adopted HFIR safety methods. The rupture pressure of the rabbit housings is demonstrated while verifying that capsule swelling during and after rupture will not block coolant flow. Then, a safety factor is applied to ascertain an administrative operating pressure. Additionally, the thermal-hydraulic performance of the HFIR facility filled with large-diameter rabbit capsules is shown to not violate previously determined safety criteria. Next, heat transfer coefficients are determined for use in design calculations. Furthermore, this report gives an example of internal configurations for the new, larger rabbit capsules that use relevant boiling water reactor (BWR) cladding geometry. Finally, this report documents an example thermal design performance for a rabbit capsule containing six gauge-curved tensile tube specimens. The thermal performance gives predicted temperature distributions within the capsule and shows the expected temperature of the passive thermometers for post-irradiation temperature comparisons.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Initial Observations from AGR 5/6/7 Capsule 1
The fourth and final irradiation experiment in the Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) fuel development and qualification program is designated as AGR-5/6/7. Data collected from the fabrication, irradiation, and post-irradiation examination (PIE) of this tristructural isotropic (TRISO) fuel are intended to serve as the primary data set for the qualification of this fuel for use in high-temperature gas-cooled reactors (INL 2021, Collin 2018b). However, data collected from the three preceding irradiations (i.e., AGR-1, AGR-2, and AGR-3/4) may also be used to supplement data collected from AGR-5/6/7. All components of the AGR-5/6/7 fuel (i.e., UCO kernels, TRISO coatings, and fuel compacts) were produced on an engineering scale at BWXT (Lynchburg, Virginia USA) according to the fuel specification (Marshall 2017). This fuel was irradiated in the northeast flux trap (NEFT) at the Advanced Test Reactor (ATR) at Idaho national Laboratory (INL) from February 16, 2018 to July 22, 2020 (Pham et al. 2021). Measurements in the fission product monitoring system (FPMS) indicated unexpected and significant numbers of failures of TRISO particles in Capsule 1 near the end of the sixth irradiation cycle (ATR Cycle 166A). In the fourth cycle (ATR Cycle 164B) and beyond, the sweep gas flow became very low (presumably from degradation of the capsule gas outlet line via an unidentified mechanism), and the program deliberately isolated Capsule 1 from gas flow periodically. In later cycles, attempts to reestablish any kind of flow in Capsule 1 were unsuccessful. With little or no flow through Capsule 1, FPMS measurements and enumerations of failed particles in Capsule 1 were difficult or impossible as was the ability to control the helium/neon gas mixture used for temperature control. Gas flows and fission gas activity in the effluent gas from the other AGR-5/6/7 capsules were also impacted by the Capsule 1 gas flow issues and the large increase in fission gas released from the Capsule 1.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT
The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
First graded metal pushered single shell capsule implosions on the National Ignition Facility
Graded metal pushered single shell (PSS) capsules are predicted to be a viable alternative to low-Z capsule indirect drive inertial confinement fusion (ICF) implosions for achieving high fusion yields [MacLaren et al., Phys. Plasmas 28, 122710 (2021)]. The first experiments with Be/Cr-graded metal PSS capsules indicate that the implementation of the principle design feature, the graded density inner metal layer, has succeeded in producing a stable implosion with performance in agreement with predictions. With 50% Cr concentration in the pusher, PSS capsules have greater than ~2× higher shell densities during stagnation for enhanced core confinement and radiation trapping at ~35% lower shell implosion velocities than low-Z capsules. High-energy >30 keV inflight shell radiography recorded 215 km/s implosion velocities and show that implosion Legendre mode P 2 asymmetry can be tuned via inner-to-outer beam wavelength separation, similar to other implosions. Shell radiographs and neutron core images show similar P 2 asymmetry, suggesting no symmetry swings between peak implosion velocity and stagnation times. Despite the modest implosion velocities, gas-filled deuterium–tritium capsule implosions generate 10 15 neutron yields at relatively modest core ion temperatures of 2.75 keV, indicating that in spite of the high-density inner layer, the implosions have been stabilized by the design density gradient. When compared with hydrodynamic simulations, the measured yield-over-simulated is 35% due to fuel–pusher mix and other perturbations such as the capsule fill tube. Finally, simple analytical scalings of hot spot pressure and neutron yield show that PSS implosions reach similar performance at lower implosion velocities and higher shell densities to low-Z ICF capsules.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Charpy Impact Characterization of Surveillance Specimens Harvested from Palisades High Fluence A-60 Capsule
Located on the shores of Lake Michigan, the Palisades Nuclear Generating Station (PNGS) was a nuclear power plant that operated in Covert Township, Michigan. The plant had a single pressurized water reactor that produced electricity for the region. The PNGS was shut down in 2022 after more than four decades of service. The PNGS included in its surveillance program a surveillance capsule, designated A-60, containing specimens of a weld metal with nickel content of about 1.36 wt% and copper content of about 0.25 wt%. The capsule was removed from its surveillance position in early 1995 and has been resident in the spent fuel pool since that time. This capsule was irradiated to a fluence of 1.96×10 20 n/cm 2 (E>1MeV) that is equivalent for more than 150 effective full power years (EFPYs) for the US reactor pressure vessel (RPV) fleet. The material is also of special interest because of its very high nickel content and potential for development of NiMnSi (nickel-manganese-silicon) precipitates. Combination of very high fluence and very high Ni and Cu content makes the material in this capsule of great interest as benchmark for currently developing embrittlement trend curves (ETC) aiming to predict embrittlement at high fluences. Multi-year efforts by the Light Water Reactor Sustainability (LWRS) Program personnel from the Materials Research Pathway (MRP) to harvest this capsule finally succeeded in 2023. As a result, the Westinghouse Electric Company (WEC) through contract with Oak Ridge National Laboratory (ORNL) came to PNGS site, retrieved the A-60 capsule, brought it to WEC Churchill hot cell facility, opened the capsule and sent all surveillance specimens in the capsule to ORNL for future characterization by July 2023. Total of 9 tensile and 48 Charpy specimens were inventoried in the ORNL hot cells. Testing plan has been developed based on available specimens. It includes hardness, tensile, Charpy impact, Mini-CT fracture toughness testing, in-situ thermal annealing, and microstructural characterization, including Atom Probe Tomography (APT). Charpy impact testing has been completed and results are presented in this report. Moreover, negotiations with Pressurized Water Reactors Owners Group (PWROG) and Westinghouse resulted in Westinghouse donating to ORNL a piece of archive weld and base metal such that unirradiated characterization of these materials can be performed as part of this project. Comparison of measured hardening and embrittlement has been performed and data are compared to available large database.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
A standard capsule design for structural material testing in the Advanced Test Reactor
Nuclear materials testing is commonly completed in various material test and research reactors throughout the world, including the Advanced Test Reactor (ATR), but the current capsule design, analysis, and fabrication process can take years to complete. To decrease the costs and time associated with materials testing, a standard capsule has been designed which houses various specimen geometries and allows irradiation in virtually any ATR test position. The standard capsule features locking end caps which connect and lock together to form the capsule stack, eliminating the need for a basket and maximizing the quantity of specimens which are contained within the capsule. A customizable internal gas gap provides thermal resistance between the specimens and reactor coolant, making specimen temperatures from approximately 370 to 1000 K achievable. The flexibility of the capsule design allows experimenters to choose irradiation positions based off desired neutron flux, with typical fluences per cycle ranging from 8.8x10 19 to 2.3x10 21 n/cm 2 , depending on experiment position. Here this paper presents and discusses the standard capsule design and analysis.
36 MATERIALS SCIENCE↗
Capsule network-based semantic segmentation model for thermal anomaly identification on building envelopes
Thermography technology is widely used to inspect thermal anomalies in building façade systems. Computer vision-based techniques provide opportunities to autonomously detect such heat anomalies to significantly improve the efficiency of decision-making for building envelope retrofitting and maintenance. Here, in this work, we propose a novel Capsule Network-based deep learning model – CapsLab – that detects and identifies thermal anomalies by semantic segmentation. CapsLab is built based on our proposed prediction-tuning capsule (PT-Capsule) layer. Different from a traditional capsule layer, which consists of part-whole transformation and capsule-routing process, the proposed layer is composed of a prediction and tuning process, which helps decreasing the number of model parameters significantly. While the applicability of traditional Capsule Networks (CapsNets) has been limited to simpler tasks and smaller datasets due to their scalability issue, we can leverage the lightweight of the proposed PT-Capsule layer, and apply it to the semantic segmentation task. In this work, we also employ our previously presented performance metric, referred to as the Anomaly Identification Metric (AIM) (Kakillioglua et al. 2021), to evaluate the segmentation outputs. Traditional performance metrics do not accurately reflect the true performance of the segmentation models in thermal anomaly identification due to the high subjectivity in the annotation process and higher overlap ratio sensitivity of the standard metrics. AIM, on the other hand, is robust to these drawbacks. Experimental results show, both qualitatively and quantitatively, that our proposed segmentation method can effectively segment the thermal anomalies. Specifically, our model provides 9.38% and 13.53% improvements over the baseline model – DeepLabV3+ – based on traditional mIoU score and the AIM score, respectively, while requiring less model parameters and less computation at the same time. In addition, the scores that the AIM metric generates better align with the scores provided by building performance experts.
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗