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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.

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623 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Proton branching ratios of 23 Mg levels

Background: The anomalous 22 Ne abundance measured in certain presolar graphite grains is thought to arise from the decay of 22 Na that was synthesized at high temperatures in core-collapse supernovae. To better interpret this abundance anomaly, the primary destruction mechanism of 22 Na, the 22 Na(p,γ) 23 Mg reaction, must be better understood. Purpose: Determine proton branching ratios of several 23 Mg excited states that play a role in the high-temperature 22 Na(p,γ) 23 Mg reaction rate. Methods: Particle decays of 23 Mg excited states populated with the previously reported 24 Mg(p,d) 23 Mg transfer reaction measurement [Kwag et al., Eur. Phys. J. A 56, 108 (2020)] were analyzed to extract proton branching ratios. The reaction was studied using a 31-MeV proton beam from the Holifield Radioactive Ion Beam Facility of Oak Ridge National Laboratory and 24 Mg solid targets. Results: Proton branching ratios of several 23 Mg excited states in the energy range of Ex = 8.044 - 9.642 MeV were experimentally determined for the first time for the p0 and p1'(p1+p2+p3) decay channels. Conclusions: These new branching ratios for 23 Mg levels can provide an experimental foundation for an improved high-temperature rate of the 22 Na(p,γ) 23 Mg reaction needed to understand production of anomalously high 22 Ne abundance in core-collapse supernovae.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Thermal Neutron Scattering Sublibrary

The thermal neutron scattering law (TSL) sublibrary aims to describe the interaction of incident neutrons at thermal or sub-thermal energies with different compound materials such fuels, moderators and special-purpose materials. In ENDF/B-VIII.1 there was a large number of new and updated TSL evaluations, including traditional moderators (light water, Beryllium metal, Beryllium Oxide, Calcium Hydride, plastics (Polystyrene and Lucite), graphite (reactor-grade and crystalline), anhydrous Hydrogen Fluoride, and heavy paraffinic oil); exotic moderators (Beryllium Carbide, Zirconium Hydride, Yttrium Hydride, Lithium-7 Hydride and Deuteride), FLiBe molten salt, structural materials and cladding (Silicon Carbide, Silicon Dioxide, Zirconium Carbide), fuels (Plutonium Dioxide, Uranium Carbide, Uranium metal, Uranium Nitride, Uranium Dioxide, Uranium Hydride), and special purpose materials. In ENDF/B-VIII.1 we also distribute alongside the evaluated files, a comma-separated file (CSV), named TSL_MAT_numbers.csv, which lists all evaluated files in the current release and their corresponding unique MAT number.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Fission Product Quantification and Burnup Analysis via Gamma Spectrometry of the AGR-5/6/7 TRISO Fuel Experiment

Five capsules containing varying configurations of TRISO fuel compacts were irradiated in the most recent Advanced Gas Reactor (AGR) TRISO test campaign, AGR-5/6/7. As part of the post-irradiation examination (PIE) of this work all the fuel compacts and capsule graphite holders were measured with the Precision Gamma Scanner (PGS) to determine isotopic activities and locations using gamma spectrometry. Measuring the compacts allowed for the comparison of physics model depletion calculations with the measured compact activity to validate models. These measurements also are used to find compacts of interest with low fission product retention that may be tested in subsequent PIE. Measurements of the holders can confirm fission product distribution and migration. Of the isotopes measured, Cs-137 and Cs-134 were used to estimate location of failed particles as well as determine a measured burnup estimation. Results from Capsule 1 compacts and the holder indicate many fuel particle failures occurred. There are high amounts of cesium in the holder near Stack 9 of Capsule 1 that corresponds to low fractions?measured-to-calculated (M/C) ratio?of cesium remaining in those compacts. All other capsules had cesium retention ratios near 1. The isotope 110mAg had less consistent results. Capsule 2 compacts exhibited a M/C of 0.20, much lower than anticipated for the temperatures these compacts underwent. Capsule 3 experienced much higher temperatures, >1300C and had an average M/C of 0.75. These are the opposite results than what was predicted by past AGR experiments for the relationship between temperature and isotope retention. Capsule 4 and 5 showed good fission product retention and agreement with the physics model predictions. The burnup results from the PGS measurements showed strong agreement between the physics models and the measured values. Using a ratio of Cs-134/Cs-137 to calculate the burnup match closely with the physics models. Additionally, the Cs-137 burnup results were less accurate but extremely close to the model results.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Electrochemical stability and corrosion behavior of Ni-Cr alloy in molten LiCl-KCl salt

This study investigated the electrochemical properties and potential-dependent stability of Cr(II) and Ni(II) in a eutectic LiCl-KCl molten salt at 500 °C under an argon atmosphere using a three-electrode cell configuration with a glassy carbon working electrode, Ag/Ag + reference electrode, and graphite counter electrode. Cyclic voltammetry was employed to characterize the Cr(0)/Cr(II)/Cr(III) and Ni(0)/Ni(II) redox transitions, yielding diffusion coefficients of 0.50 (±0.01) × 10 −5 cm 2 s −1 for Cr(II) and 0.91 (±0.16) × 10 −5 cm 2 s −1 for Ni(II) at 500 °C. The redox stability domains of these species provided insight into the corrosion mechanisms of a Ni-Cr alloy (80–20 wt %) in LiCl-KCl salt. The measured open circuit potential of the alloy corresponded to the range of potentials where the Ni(0) and Cr(II) species are stable, indicating a preferential dissolution of Cr. Immersion testing of the alloy for 336 h confirmed significant Cr depletion (or Ni-enrichment) at the surface layer, corroborating Cr dealloying as the dominant corrosion reaction under open circuit conditions. Additionally, the application of a small anodic overpotential (0.11 V) resulted in selective Cr dissolution while a large anodic overpotential (0.25 V) resulted in co-dissolution of both Cr and Ni. Furthermore, these results demonstrate that the corrosion mechanisms of a Ni-Cr alloy in LiCl-KCl depend upon the applied potential, and thus the redox conditions of the molten salt should be carefully controlled in the design of a redox buffer to mitigate corrosion in molten salt environments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Interfacial Pressure Improves Calendar Aging of Lithium Metal Anodes

Lithium metal is a very attractive anode material because its theoretical specific capacity is approximately 10 times higher than conventional graphite anodes. Despite great promise, Li anodes suffer from capacity fade due to instabilities with the electrolyte as well as stranding of active Li. We have previously shown that applied interfacial pressure improves Li anode cycling because the pressure reduces the propensity for Li isolation and enables easier reconnection. Many researchers have also shown that calendar aging can lead to Li capacity loss and this has been attributed to either electrolyte decomposition with concurrent Li corrosion or to the formation of stranded Li. Our prior research focused on calendar aging during cycling suggests the mechanism for calendar aging is largely related to stranding of Li during rest and reconnection of the stranded Li upon further cycling, evidenced by similar average Coulombic efficiencies and Li loss in cells with and without rest. Because our calendar aging studies suggest Li stranding as a major cause of Coulombic efficiency drops and our Li cycling studies suggest this can be mitigated partially through applied interfacial pressure, we hypothesized that applied pressure would improve calendar aging by reducing stranded Li and enabling reconnection. We systematically varied applied pressure (0-1000 kPa) on Li metal anodes during cycling tests with and without intermittent calendar aging periods. Though the Coulombic efficiency decreases during aging periods, the lost capacity is recovered during subsequent cycles, as shown though average Coulombic efficiency and cumulative Li capacity loss analysis. We find that application of pressure partially mitigates calendar aging, in accordance with our hypothesis that calendar aging is caused by Li standing and can be mitigated to some degree with interfacial pressure. This is further supported by our results showing that the average Coulombic efficiency and cumulative Li capacity losses are similar over 50 cycles for cells that were continuously cycled and cells with periodic calendar aging periods. This result indicates that the losses during aging are reversible, which is consistent with Li stranding and reconnection. We show that pressure is one mitigation technique that helps reduce Li calendar aging in this study, but our finding that calendar aging is primarily governed by the stranding and reconnection of dead Li has wider implications. This research suggests that other mitigations which have been shown to prevent dead Li formation or encourage reconnection during cycling would also likely be successful for the purpose of improving calendar aging. The authors were supported by a Laboratory Directed Research and Development (LDRD) program. This work was performed, in part, at the Center for Integrated Nanotechnologies, an Office of Science User Facility operated for the U.S. Department of Energy (DOE) Office of Science. Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC (NTESS), a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) under contract DE-NA0003525. This written work is authored by an employee of NTESS. The employee, not NTESS, owns the right, title and interest in and to the written work and is responsible for its contents. This work was authored in part by the National Renewable Energy Laboratory, operated by Alliance for Sustainable Energy, LLC, for the U.S. Department of Energy (DOE) under Contract No. DE-AC36-08GO28308. Any subjective views or opinions that might be expressed in the written work do not necessarily represent the views of the U.S. Government.

applied pressure↗
DOE OSTI2021

Effects of 18 O Exchange on Neutron Emission Rates of Aging 238 PuO 2 LWRHUs

The Lightweight Radioisotope Heater Unit (LWRHU) provides about 1 Watt of thermal power to critical electrical and mechanical systems in the extreme temperatures of space, maintaining the operating temperature of the spacecraft’s scientific and general operating equipment. Los Alamos National Laboratory (LANL) designed and tested the LWRHU in the 1970s through early 1980s. The first production campaign at LANL was in the early 1980s in support of the Galileo mission in which over 100 LWRHUs were manufactured. In the mid-1990s, almost 200 LWRHUs were manufactured in support of the Cassini mission. After the primary Cassini manufacturing campaign, the Department of Energy (DOE) approved the fabrication of more than 10 additional LWRHUs as spares in 1998. The LWRHU assembly contains the following components: a fuel pellet, a vented capsule, a pyrolytic graphite insulator, and a fine-weave pierced fabric (FWPF) graphite aeroshell. The fuel pellet is a hot-pressed 238 PuO 2 cylindrical pellet that is sintered to create a ceramic pellet. The pellet is encapsulated in a platinum-rhodium encapsulation. The welded fueled clad is placed inside a pyrolytic graphite thermal insulation to protect the fuel from the heat produced in reentry events. The exterior layer of the LWRHU assembly is the FWPF aeroshell. Before sintering, PuO 2 granules are heated in a furnace at temperatures greater than or equal to 700°C under an enriched 16 O atmosphere in order to reduce trace quantities of 18 O. Because 18 O undergoes an (α,n) reaction, 18 O needs to be removed as much as possible to meet radiation emission requirements set by the Jet Propulsion Laboratory (JPL) and the National Aeronautics and Space Administration (NASA), as well as other agencies. The relative abundances of the oxygen isotopes are indicated by neutron emission rate (NER) measurements. LWRHUs are vented with a sintered platinum frit vent (George 1986). During extraterrestrial operation, the vent in the cladding allows helium from the decay of plutonium to escape and relieves pressure from the system while preventing the solids from escaping (Rinehart 1996, Tate 1982 & 1985). This vent is protected during manufacture by a platinum-30 rhodium cover. The capsule vent is activated by milling a 0.025-inch diameter hole through the protective cover to a nominal depth of 0.015 inches, generally immediately before the LWRHU is loaded into the aeroshell (George 1986). Prior to long-term storage, LWRHUs are vented and thus potentially allow infiltration of 18 O from the ambient atmosphere. Although NER decreases with age due to the reduction in emitted alpha particles, it might be expected that NER reduction is mitigated by the exchange of 18 O. However, this is not necessarily the case and no trend was observed relating NER to age other than the natural decay of 238 Pu (Mulford 2021). The most probable and dominant factor for determining specific NER was manufacturing discrepancies between individual units. It is important to understand the lack of observed 16 O- 18 O exchange and to compare LWRHU pellets as much as possible.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

A Single Small-Scale Plutonium Redox Reaction System Yields Three Crystallographically-Characterizable Organoplutonium Complexes

Here, an approach to obtaining substantial amounts of data from a hazardous starting material that can only be obtained and handled in small quantities is demonstrated by the investigation of a single small-scale reaction of cyclooctatetraene, C 8 H 8 , with a solution obtained from the reduction of Cp' 3 Pu (Cp' = C 5 H 4 SiMe 3 ) with potassium graphite. This one reaction coupled with oxidation of a product has provided single-crystal X-ray structural data on three organoplutonium compounds as well as information on redox chemistry thereby demonstrating an efficient route to new reactivity and structural information on this highly radioactive element. The crystal structures were obtained from the reduction of C 8 H 8 by a putative Pu(II) complex, (Cp' 3 Pu II ) 1– , generated in situ, to form the Pu(III) cyclooctatetraenide complex, [K(crypt)][(C 8 H 8 ) 2 Pu III ], 1-Pu , and the tetra(cyclopentadienyl) Pu(III) complex, [K(crypt)][Cp' 4 Pu III ], 2-Pu . Oxidation of the sample of 1-Pu with Ag(I) afforded a third organoplutonium complex that has been structurally characterized for the first time, (C 8 H 8 ) 2 Pu IV , 3-Pu . Complexes 1-Pu and 3-Pu contain Pu sandwiched between parallel (C 8 H 8 ) 2– rings. The (Cp' 4 Pu III ) – anion in 2-Pu features three η 5 -Cp' rings and one η 1 -Cp' ring, which is a rare example of a formal Pu–C η 1 -bond. In addition, this study addresses the challenge of small-scale synthesis imparted by radiological and material availability of transuranium isotopes, in particular that of pure metal samples. A route to an anhydrous Pu(III) starting material from the more readily available Pu IV O 2 was developed to facilitate reproducible syntheses and allow complete spectroscopic analysis of 1-Pu and 2-Pu . Pu IV O 2 was converted to Pu III Br 3 (DME) 2 (DME = CH 3 OCH 2 CH 2 OCH 3 ) and subsequently Pu III Br 3 (THF) x , which was used to independently synthesize 1-Pu , 2-Pu , and 3-Pu .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tri-structural isotropic fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a three-dimensional finite element thermal model, which is subject to simulation uncertainty. The most dominant factor in the uncertainty of calculated fuel temperatures is the gas gap uncertainty due to the nub-to-shell clearance caused by a design error of AGR-5/6/7 capsules. The thermal model was revised to examine the most probable graphite offset position for six different days during the irradiation for Capsules 1 and 2. The analysis varied the offset distance and azimuthal direction at both the top and bottom of the holder. The best-fit offset was estimated based on the minimum root mean square error of the residuals (measured minus calculated) for the operational thermocouples (TCs). From these results, the following conclusions were made: (1) The holder offsets led to slightly lower average temperatures but wider temperature variations (lower minimum and higher peak fuel temperatures) for both Capsule 1 and Capsule 2. (2) During earlier cycles (162A–164B), when numerous TCs were still operational, the best-fit offset distance varied over a specific range for both the top and bottom ([0.002–0.0035 in.] for Capsule 1 and [0.003-0.004 in] for Capsule 2). In contrast, the offset azimuthal direction varied widely, especially for the offset at the bottom of the Capsule 1 holder. This is because holder movement was somewhat constrained at the top of Capsule 1 by the TC leads running through the capsule head and into the holder and by the through tubes in Capsule 2, but the Capsule 1 bottom did not have this type of constraint. (3) During later cycles, when all TCs failed, applying the maximum possible offset of 0.006 in. to the northwest direction for both the top and bottom resulted in a calculated peak fuel temperature of 1557? in Capsule 1 (i.e., a 135? increase from 1422? with zero offset on September 20, 2019 (166A)); the maximum offset of 0.0068 in. to the south for both top and bottom resulted in a calculated peak fuel temperature of 1110°C in Capsule 2 on April 20, 2020 (i.e., a 116? increase from 994? with zero offset (168A)). High peak fuel temperatures in Capsule 1 during Cycle 166A could be the cause of massive particle failure near the end of this cycle. (4) Even though the highest temperature at the tip of Type-N TCs, such as TC-1-7, slightly exceeded 1000?, the temperature along the TC wire reached as high as 1335? assuming an offset of 0.006 in. in the northwest of Capsule 1 holder near the end of Cycle 166A. This temperature significantly exceeds the temperature threshold at which TC degradation is expected to occur, ultimately contributing to considerable particle failures in Capsule 1. For eight Type-N TCs in Capsule 2, the peak TC line temperature was much lower (i.e., 1029°C for TC-2-5), assuming maximum offset of 00068 in. to the south during cycle 168A.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tri-structural isotropic fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a three-dimensional finite element thermal model, which is subject to simulation uncertainty. The most dominant factor in the uncertainty of calculated fuel temperatures is the gas gap uncertainty due to the nub-to-shell clearance caused by a design error of AGR-5/6/7 capsules. The thermal model was revised to examine the most probable graphite offset position for six different days during the irradiation for Capsules 1 and 2. The analysis varied the offset distance and azimuthal direction at both the top and bottom of the holder. The best-fit offset was estimated based on the minimum root mean square error of the residuals (measured minus calculated) for the operational thermocouples (TCs). From these results, the following conclusions were made: (1) The holder offsets led to slightly lower average temperatures but wider temperature variations (lower minimum and higher peak fuel temperatures) for both Capsule 1 and Capsule 2. (2) During earlier cycles (162A–164B), when numerous TCs were still operational, the best-fit offset distance varied over a specific range for both the top and bottom ([0.002–0.0035 in.] for Capsule 1 and [0.003-0.004 in] for Capsule 2). In contrast, the offset azimuthal direction varied widely, especially for the offset at the bottom of the Capsule 1 holder. This is because holder movement was somewhat constrained at the top of Capsule 1 by the TC leads running through the capsule head and into the holder and by the through tubes in Capsule 2, but the Capsule 1 bottom did not have this type of constraint. (3) During later cycles, when all TCs failed, applying the maximum possible offset of 0.006 in. to the northwest direction for both the top and bottom resulted in a calculated peak fuel temperature of 1557? in Capsule 1 (i.e., a 135? increase from 1422? with zero offset on September 20, 2019 (166A)); the maximum offset of 0.0068 in. to the south for both top and bottom resulted in a calculated peak fuel temperature of 1110°C in Capsule 2 on April 20, 2020 (i.e., a 116? increase from 994? with zero offset (168A)). High peak fuel temperatures in Capsule 1 during Cycle 166A could be the cause of massive particle failure near the end of this cycle. (4) Even though the highest temperature at the tip of Type-N TCs, such as TC-1-7, slightly exceeded 1000?, the temperature along the TC wire reached as high as 1335? assuming an offset of 0.006 in. in the northwest of Capsule 1 holder near the end of Cycle 166A. This temperature significantly exceeds the temperature threshold at which TC degradation is expected to occur, ultimately contributing to considerable particle failures in Capsule 1. For eight Type-N TCs in Capsule 2, the peak TC line temperature was much lower (i.e., 1029°C for TC-2-5), assuming maximum offset of 00068 in. to the south during cycle 168A.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Core Design of the Holos-Quad Microreactor

The Holos-Quad micro-reactor concept, developed by HolosGen LLC, is equipped with a 22 MWt (Mega-Watt thermal) core and an integral power conversion system converting the core thermal energy into approximately 10 MWe (Mega-Watt electric). This design can be configured to support a wide range of applications. It is a very innovative high-temperature gas-cooled reactor concept using TRI-structural ISOtropic particle fuel (TRISO) distributed in graphite hexagonal blocks, cooled with helium in a direct Brayton cycle independently executed by four Subcritical Power Modules (SPMs) fitted into a hardened 40-foot container whose dimensions are in compliance with ISO shipping containers requirements. In FY2019 HolosGen LLC was awarded by the Department of Energy Advanced Research Project Agency-Energy (DOE ARPA-E) under the MEITNER funding program. As part of the MEITNER award, the Argonne National Laboratory (ANL) contributed expertise through two specialized teams: The “Design Team” and the “Resource Team”. The Design Team was dedicated to validate feasibility of the Holos-Quad core and to optimize its core design through neutronics analyses. The Resource Team was dedicated to feasibility verification via high-fidelity codes of Holos-Quad thermal-hydraulic, heat transfer, shielding, and structural aspects. This report summarizes the activities conducted by ANL Design Team. A rigorous design approach based on multi-criteria optimization and code-to-code comparison involving stochastic and high-fidelity deterministic solutions was developed and employed at several evolutionary stages of the Holos-Quad design. Several generations of the Holos-Quad core were designed within this project before converging to the current full-scale Gen 2+ design that is detailed in this report. Figure EA-1 illustrates a cross-sectional view of Gen 2+ Holos-Quad core configuration, and Figure EA-2 provides a simplified perspective view of 1-of-4 SPMs. The Holos-Quad uses four thermal-hydraulically independent SPMs locked into stationary positions during power operation, surrounded by BeO reflector and structural component fully comprised within the dimensional constraints represented by traditional ISO containers. One of the benefits of this approach is to enable transportation of each SPM promptly after irradiation in shielded containers. The core is designed to operate for approximately 8 full-power years while the reactivity controls and power conversion system enable load-following operations. The reactivity controls are represented by independent, diversified, and redundant reactivity control systems based on control drums and redundant sets of shutdown rods. The high-fidelity simulation tools were used to assess detailed power and flux distributions of the three-dimensional full-core or quarter-core of the Gen 2+ configuration. Single-physics and multi-physics simulations of the neutronics code PROTEUS and the thermal-hydraulic code System Analysis Module (SAM) were performed to analyze the Holos design configurations with detailed high-fidelity solutions. The design work performed confirmed feasibility of the Holos-Quad concept, provided realistic design description for detailed design of the operational system, and identified several core design improvements to be further considered for future reactor development activities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

AGR-5/6/7 Final Release-to-Birth Ratio Data Analysis

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel for use in the high-temperature gas cooled reactor (HTGR). AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product (FP) transport models and codes. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. Release-to-birth ratios (R/B) for fission-gas isotopes released from each capsule are calculated from release rates, measured by germanium detectors in the Fission Product Monitoring System (FPMS) installed downstream from each capsule, and birth rates calculated using numerical models of FG generation. The R/Bs are a critical measure of the ability of the fuel kernel, the particle coating layers, and the compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, and the impact of initially defective particles and/or particle-coating failures that occur during irradiation. For fission-gas isotopes, particle failure is defined as failure of all coating layers, allowing gaseous fission atoms to escape from a particle. During the first five cycles (162B ? 165A), R/Bs were stable in the 10-8?10-6 range, and no in-pile particle failures were observed, based on the gross gamma counts. The maximum R/B value of around 2 ? 10-6 for Kr-85m resulted from the presence of as-fabricated exposed kernels (based on the high exposed kernel fraction), the dispersed uranium, and high fuel particle temperatures in Capsule 1. Comparison of capsule-measured R/Bs from these early cycles to predictions using the previously developed AGR R/B model demonstrated FG release from the AGR-5/6/7 TRISO fuel was comparable to that of previous experiments. In addition, the Kr-85m R/B per-exposed-kernel values are comparable to R/B values obtained in AGR-3/4 irradiation experiment and four irradiation experiments performed during 1980s: (1) HRB-17/18, (2) COMEDIE-BD1, (3) HFR B1, and (4) HRB-21. In contrast, all measured R/B values are lower than predictions by the commonly used Richards and German models, which are intentionally conservative. A large number of in-pile particle failures occurred in Capsule 1 by the end of Cycle 166A. During the final four cycles (166A ? 168A), apparent damage to the Capsule 1 gas line appeared to cause FG leakage from that capsule into the other four capsules, resulting in an increase in fission gas (FG) detected in the effluent for all capsules. Isolation of the Capsule 1 gas line during the last three cycles also prevented measurement of its FG release. Thus, R/Bs in all capsules after Cycle 166A are highly uncertain because of undefined amount of leakage from Capsule 1, especially for long-lived isotopes. A few hundred in-pile particle failures were estimated for Capsule 1 before the end of Cycle 166A, but the total number of failures is unknown due to the lack of FG release data in the later cycles. Based primarily on evidence from the gross gamma counts during Cycle 168A, approximately 15 particles failed in Capsule 3 and four particles failed in Capsule 2. In-pile failures in Capsule 3 were anticipated because this capsule was designed to operate beyond the HTGR normal operating temperature range. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR model, developed in INL/EXT-14-32970, for R/B of the short-lived isotopes (Kr-89 and Xe-137) with minimal leakage from Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗