Towards Reentry Capsule Dynamic Stability Prediction: Recent Experience using PyFR for Capsule Wake Flow
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Rathakrishnan Bhaskaran↗
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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↗
Analysis of Fission Products on the AGR-1 Capsule Components
The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (1) the metal capsule components, (2) the graphite fuel holders, (3) the graphite spacers, and (4) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs-134, Cs-137, Eu-154, and Sr-90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 -2 (Capsule 3) to 3.8×10 -1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag-110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Biocompatible Capsules and Methods of Making
Embodiments of the invention include capsules for containing medical implants and delivery systems for release of active biological substances into a host body. Delivery systems comprise a capsule comprising an interior enclosed by walls, and a source of active biological substances enclosed within the capsule interior, wherein the active biological substances are free to diffuse across the capsule walls. The capsule walls comprise a continuous mesh of biocompatible fibers and a seal region where two capsule walls overlap. The interior of the capsule is substantially isolated from the medium surrounding the capsule, except for diffusion of at least one species of molecule between the capsule interior and the ambient medium, and prevents cell migration into or out of the capsule. Methods for preparing and using the capsules and delivery systems are provided.
Loftus, David J.↗
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↗
The effects of convergence ratio on the implosion behavior of DT layered inertial confinement fusion capsules
The wetted foam capsule design for inertial confinement fusion capsules, which includes a foam layer wetted with deuterium-tritium liquid, enables layered capsule implosions with a wide range of hot-spot convergence ratios (CR) on the National Ignition Facility. In this paper, we present a full-scale wetted foam capsule design that demonstrates high gain in one-dimensional simulations. In these simulations, increasing the convergence ratio leads to an improved capsule yield due to higher hot-spot temperatures and increased fuel areal density. High-resolution two-dimensional simulations of this design are presented with detailed and well resolved models for the capsule fill tube, support tent, surface roughness, and predicted asymmetries in the x-ray drive. Our modeling of these asymmetries is validated by comparisons with available experimental data. In 2D simulations of the full-scale wetted foam capsule design, jetting caused by the fill tube is prevented by the expansion of the tungsten-doped shell layer due to preheat. While the impacts of surface roughness and predicted asymmetries in the x-ray drive are enhanced by convergence effects, likely underpredicted in 2D at high CR, simulations predict that the capsule is robust to these features. Nevertheless, the design is highly susceptible to the effects of the capsule support tent, which negates all of the one-dimensional benefits of increasing the convergence ratio. Indeed, when the support tent is included in simulations, the yield decreases as the convergence ratio is increased for CR > 20. Finally and nevertheless, the results suggest that the full-scale wetted foam design has the potential to outperform ice layer capsules given currently achievable levels of asymmetries when fielded at low convergence ratios (CR < 20).
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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↗
Processes to Open the Container and the Sample Catcher of the Hayabusa Returned Capsule in the Planetary Material Sample Curation Facility of JAXA
Japanese spacecraft Hayabusa, which returned from near-Earth-asteroid Itokawa, successfully returned its reentry capsule to the Earth, the Woomera Prohibited Area in Australia in Jun 13th, 2010, as detailed in another paper [1]. The capsule introduced into the Planetary Material Sample Curation Facility in the Sagamihara campus of JAXA in the early morning of June 18th. Hereafter, we describe a series of processes for the returned capsule and the container to recover gas and materials in there. A transportation box of the recovered capsule was cleaned up on its outer surface beforehand and introduced into the class 10,000 clean room of the facility. Then, the capsule was extracted from the box and its plastic bag was opened and checked and photographed the outer surface of the capsule. The capsule was composed of the container, a backside ablator, a side ablator, an electronic box and a supporting frame. The container consists of an outer lid, an inner lid, a frame for latches, a container and a sample catcher, which is composed of room A and B and a rotational cylinder. After the first check, the capsule was packed in a plastic bag with N2 again, and transferred to the Chofu campus in JAXA, where the X-ray CT instrument is situated. The first X-ray CT analysis was performed on the whole returned capsule for confirming the conditions of latches and O-ring seal of the container. The analysis showed that the latches of the container should have worked normally, and that the double Orings of the container seemed to be sealed its sample catcher with no problem. After the first X-ray CT, the capsule was sent back to Sagamihara and introduced in the clean room to exclude the electronic box and the side ablator from the container by hand tools. Then the container with the backside ablator was set firmly to special jigs to fix the lid of container tightly to the container and set to a milling machine. The backside ablator was drilled by the machine to expose heads of bolts, which combined the ablator to the outer lid of the container, and after the drilling had been finished, all the bolts were unscrewed and the backside ablator was removed from the container. Then, the container was sent to the Chofu X-ray facility again to examine in detail by a micro X-ray CT instrument in order to reconfirm that the condition of the latches of the lid of container was normal and that its double O-ring seemed to have been sealed after the last X-ray CT analysis.
Fujimura, A.↗
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↗
Numerical Investigation of Roughness Effects on Transition on Spherical Capsules
The state of the boundary layer on space re-entry vehicles significantly affects the design of the thermal protection system. However, the physical mechanism that leads to the laminar-turbulent boundary-layer transition on blunt spherical capsules remains an open question in literature. This work numerically assesses the potential of roughness-induced non-modal disturbance growth on re-entry capsules with a spherical-section forebody by optimal transient-growth theory and direct numerical simulation. Two different sets of wind-tunnel experiments are considered. Optimal transient-growth studies have been performed for the blunt capsule experiments at Mach 5.9 in the Hypersonic Ludwieg tube Braunschweig (HLB) of the Technische Universitat Braunschweig. In some of these measurements, the capsule model was equipped with a specifically designed patch of distributed micron-sized surface roughness. The transient-growth results for the HLB capsule are compared to corresponding numerical data for a Mach 6 blunt capsule experiment in the Adjustable Contour Expansion (ACE) facility of the Texas A&M University (TAMU) at lower Reynolds number. Similar trends are observed for both configurations. In particular, a rather low maximum energy gain is noted for the surface temperature conditions of the experiments. It is shown that the surface temperature dependence of the optimal transient-growth results is very similar for both capsule configurations. Moreover, the generation of stationary disturbances by well-defined roughness patches on the capsule surface is studied for the conditions of the HLB experiment using direct numerical simulations (DNS). To help explain the observed laminar-turbulent transition downstream of the roughness patch in some of the HLB capsule experiments, additional simulations were carried out to study the evolution of unsteady perturbations within the steady disturbance flow field due to the roughness patch. However, the DNS did not provide any indication of modal or non-modal disturbance growth in the wake of the roughness patch, and hence, the physical mechanism underlying the observed onset of transition remains unknown.
Hein, Stefan↗
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↗
Cross-type optical separation of elastic oblate capsules in a uniform flow
The dynamic behavior of an elastic capsule with an initially oblate spheroidal shape during cross-type optical separation was numerically investigated. The penalty immersed boundary method was adopted for the fluid-membrane interaction, and the optical force calculation was conducted by using the ray optics method including the ray-surface intersection algorithm. The oblate elastic capsule of b/a = 0.5 with different surface Young's moduli and different initial inclination angles was considered. The oblate capsule with higher surface Young's moduli was less deformed, and was more migrated for each initial inclination angle. Unlike the oblate rigid particle, the initially inclined capsules with moderate inclination angles were similarly migrated since the oblate elastic capsule was deformed during rotation near the laser beam axis. The oblate capsules can be separated according to the surface Young's modulus, except for nearly non-inclined capsules. As the fluid velocity decreased, the migration distance increased. The maximum deformation parameter was insensitive to the fluid velocity. Furthermore, a new dimensionless number (S{sub ec}) was introduced to predict the migration distance of the oblate elastic capsule.
71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
Capsule Shimming Developments for National Ignition Facility (NIF) Hohlraum Asymmetry Experiments
Capsule drive in National Ignition Facility (NIF) indirect drive implosions is generated by x-ray illumination from cylindrical hohlraums. The cylindrical hohlraum geometry is axially symmetric but not spherically symmetric causing capsule-fuel drive asymmetries. We hypothesize that fabricating capsules asymmetric in wall thickness (shimmed) may compensate for drive asymmetries and improve implosion symmetry. Simulations suggest that for high compression implosions Legendre mode P 4 hohlraum flux asymmetries are the most detrimental to implosion performance. General Atomics has developed a diamond turning method to form a GDP capsule outer surface to a Legendre mode P 4 profile. The P 4 shape requires full capsule surface coverage. Thus, in order to avoid tool-lathe interference flipping the capsule part way through the machining process is required. This flipping process risks misalignment of the capsule causing a vertical step feature on the capsule surface. Recent trials have proven this step feature height can be minimized to ~0.25 µm.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗