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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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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1 st generation nuclear grade FeCrAl, (2) C26M, a 2 nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared to Zircaloy-2. C26M claddings burst at significantly higher temperatures for all tested engineering hoop stresses, had limited ballooning, and demonstrated preferential fuel retention behavior in terms of burst opening area and length. High temperature tensile data for C26M is also presented. For both un-irradiated FeCrAl alloys, it was found that a distinct “threshold” burst stress signified the transition between small and large openings. Finally, higher threshold hoop stresses were associated with higher uniaxial strength for the FeCrAl alloys, indicating that tensile data, rather than creep data, could be useful for predicting rupture size and assessing fuel dispersal concerns.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Burst characteristics of advanced accident-tolerant FeCrAl cladding under temperature transient testing

Here assessment of burst characteristics of accident-tolerant fuel (ATF) claddings is essential to evaluate the safety margins of nuclear reactors, and better understanding can enable accelerated licensing for reactor concepts that include the new materials. Therefore, this study investigated the burst behavior of an ATF candidate of C26M, which is an iron–chromium–aluminum alloy (FeCrAl) under transient testing like the simulated loss-of-coolant accident (LOCA) conditions, except the water-quenching phase, in light-water reactors (LWRs). The effect of LOCA specimen length was assessed in terms of post-burst tube parameters. No critical length effect was determined on the burst pressure, burst location, and burst size measurements. Burst temperature showed larger variation likely due to its measurement approach used in this study. The post-test diametral strain was identified as the critical parameter to reduce the specimen length as compared to tube burst length or width. Postmortem optical metrology and digital image correlation were employed to determine the local strain-state during the simulated LOCA, which showed the loading path was close to equibiaxial conditions in regions away from the burst location, rather than internally pressurized conditions (or “plane-strain” tension). Pst-test microstructural characterizations of the FeCrAl revealed that ductile damage was present at the edge of the outer radial surface while the rest of the material was ruptured via cleavage at LOCA burst temperatures. Furthermore, internal grain boundary cracks were observed at locations away from the tube burst region. This behavior was considered to occur during the cooling down period of the simulated LOCA test. Overall, this study aimed to provide essential data for fuel performance code development by considering micromechanics of the deformation and failure.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts

Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. Here, in this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO 2 formation via Cr 2 O 3 formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al 2 O 3 regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H 2 generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H 2 generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H 2 generation by orders of magnitude relative to bare Zry-4.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Measured and projected beam backgrounds in the Belle II experiment at the SuperKEKB collider

The Belle II experiment at the SuperKEKB electron–positron collider aims to collect an unprecedented data set of 50 ab -1 to study CP -violation in the B -meson system and to search for Physics beyond the Standard Model. SuperKEKB is already the world’s highest-luminosity collider. In order to collect the planned data set within approximately one decade, the target is to reach a peak luminosity of 6 x 10 35 cm −2 s −1 by further increasing the beam currents and reducing the beam size at the interaction point by squeezing the betatron function down to β $^*_y$ = 0.3 mm. To ensure detector longevity and maintain good reconstruction performance, beam backgrounds must remain well controlled. We report on current background rates in Belle II and compare these against simulation. We find that a number of recent refinements have significantly improved the background simulation accuracy. Finally, we estimate the safety margins going forward. We predict that backgrounds should remain high but acceptable until a luminosity of at least 2.8 x 10 35 cm −2 s −1 is reached for β $^*_y$ = 0.6 mm. At this point, the most vulnerable Belle II detectors, the Time-of-Propagation (TOP) particle identification system and the Central Drift Chamber (CDC), have predicted background hit rates from single-beam and luminosity backgrounds that add up to approximately half of the maximum acceptable rates.

Detector background↗
DOE OSTI2020

Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout

The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Cr-coated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. In addition, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Development of the reactor core isolation cooling system model for the extended station black-out accident

In order to accurately predict the behavior of the Reactor Core Isolation Cooling System (RCIC) in the context of the FLEX strategies under the extended station black-out accident (SBO), we present a mechanistic model of the RCIC in this work. The control volume formulation of angular momentum conservation is employed to model the Terry turbine in the RCIC system. The two-phase nozzle flow model developed in the previous work is utilized to simulate the convergent-divergent nozzles in the Terry turbine. A nozzle flow table is generated to cover the operation range of the RCIC during beyond design basis accidents (BDBAs). The mechanistic RCIC turbine-pump model and the nozzle flow table are incorporated in a TRACE BWR model. Simulations of short-term and long-term SBO transients are performed to demonstrate dynamic RCIC responses. A system analysis code with the proposed mechanistic model of the RCIC system will have a better prediction of system safety response during SBO accidents. With the addition of such more realistic models, the system code will be capable of providing expanded understanding of safety margin, which can help help to form the technical basis for the successful implementation of the FLEX and SAMG measures under Extended Loss of AC Power (ELAP) conditions for BWR or PWR reactor designs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility

The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Evaluation of flow-induced plate deflection for University of Missouri research reactor low-enriched uranium fuel element

The University of Missouri Research Reactor (MURR), located on the campus of the University of Missouri in Columbia, Missouri, is one of the six United States (U.S.) High Performance Research Reactors (USHPRR), including one critical facility, that are actively collaborating with the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program to convert from highly enriched uranium (HEU, ≥20 wt% U-235) fuel to low-enriched uranium (LEU, <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on a monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow conversion of some USHPRR, including MURR. In the design of its fuel elements, MURR is using thin parallel curved fuel plates separated by coolant channels. In this work, fluid-structure interaction (FSI) analysis of the MURR LEU fuel element is performed at the element level (as compared to the plate level analysis), which models all components of the LEU fuel element, including fuel plates and the supporting structures. Therefore, the effect of supporting structures on the flow distribution within the element and the fuel plate deflection are evaluated. In addition to the element nominal flow rate and dimensions, the tolerances in the geometry of the coolant channel and plate thickness, the effect of a comb on plate deflection, and the uncertainty of the flow rate per element are evaluated. For the LEU fuel plates, which are thinner than the current HEU plates, the predicted plate deflection is found to be small compared to the fabrication and assembly tolerances. Thus, the FSI-induced deflections are not expected to noticeably reduce the coolant flow rate or predicted safety margins in the limiting channels for the MURR LEU fuel element. In addition to the simulation work, a hydraulic performance test of the MURR LEU fuel element is currently being planned to support conversion to the use of LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Multiphysics simulation of TRISO fuel compacts and the effects of homogenization on silver release predictions

This work studies the impact of explicit and homogenized modeling approaches on the multiphysics simulation of TRistructural ISOtropic (TRISO) fuel compacts in prismatic High Temperature Gas Reactors (HTGRs) and the silver release predictions. TRISO fuel compacts exhibit complex double heterogeneity that significantly affects heat conduction, neutron transport, and fission products release. In this work, we use Cardinal, a multiphysics tool based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, to couple neutron transport and heat conduction. OpenMC is used for neutron transport and the MOOSE heat transfer module is used for the heat conduction. Then, we use BISON for the silver release predictions. Two different modeling approaches—explicit modeling of individual TRISO particles and homogenized representation using effective thermal properties—are compared at high TRISO packing fractions (20% and 40%) across varying power densities. Results demonstrate that homogenization significantly under-predicts peak temperatures, for the case of high power/TRISO, there is a difference of 83.73 K in the maximum temperature. Additionally, homogenization underestimates the silver release predictions compared to explicit modeling. Specifically, the temperature and power density differences lead to significant differences in silver release predictions. This work demonstrates the importance of accurately modeling heterogeneity of the TRISO particles to reliably predict fission product release and assess reactor safety margins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Demagnetization analysis of an 18 GHz electron cyclotron resonance ion source permanent magnet hexapole

An upgrade of ECR2 at the Argonne Tandem Linac Accelerator System is under way, focusing on increasing the intensity capabilities of the facility. ECR2 is a room temperature electron cyclotron resonance ion source, and the upgrade has strict requirements to retain radial access to the plasma chamber and keep the ion source operating without the use of superconducting magnets. The upgrade design with respect to the magnet arrangement and magnetization vectors has recently been presented [R. C. Vondrasek, J. McLain, and R. H. Scott, J. Phys.: Conf. Ser. 2743 , 012044 (2024)] using the same magnetic material as the current ECR2 hexapole. A thorough exploration of the demagnetization potential of this hexapole was carried out, and the risk of demagnetization was deemed too high, despite the magnetic performance meeting the requirements for the upgrade. Additional permanent magnet materials are considered with their respective performance evaluated. Magnet strength and demagnetization resistance are investigated and optimized with the final room temperature design demonstrating a high temperature transient demagnetization resistance and a radial magnetic field of 1.18 T at the plasma chamber wall. Finally, the simulations of this hexapole suggest that it will be sufficient to optimize intensity with a 14.5 GHz driving frequency and will allow 18 GHz operation, while keeping a high safety margin for maintaining magnet performance.

Argonne Tandem Linac Accelerator System↗
DOE OSTI2020

Dynamic PRA-Based Estimation of PWR Coping Time Using a Surrogate Model for Accident Tolerant Fuel

In this study, we propose an interpolation-based response surface surrogate methodology to manage a large number of scenarios in dynamic probabilistic risk assessment. It adopts the shape Dynamic Time Warping algorithm to cluster the interpolation neighborhood from time series sample data. The interpolation method was adapted from Taylor Kriging to allow a reduced-order model of the Taylor series. In order to demonstrate its applicability to complex issues in risk assessment for nuclear engineering, an example risk response surface to estimate emergency core cooling system (ECCS) criteria for triplex silicon carbide (SiC) accident-tolerant fuel was constructed. The response surface was exploited to estimate the cumulative failure probability of the fuel cladding structure due to the uncertainties in operator actions and safety systems. The functional failures were assessed based on a combination of individual layer failures computed by coupling Risk Analysis Virtual Environment software with a pressurized water reactor 1000-MW(electric) RELAP5 model and the in-house fuel performance assessment module. Results showed that SiC cladding failure probability spiked less than 1 min after a large-break loss-of- coolant accident whenever the current ECCS criteria for Zircaloy-4 (Zr-4) cladding was used. However, it still provides an increased safety margin of three orders of magnitude compared to Zr-4. This positive margin could be utilized to relax active ECCS requirements by allowing deviations of up to 450 s in its actuation time. The proposed surrogate methodology generated a response surface of SiC cladding failure probability reasonably well, with a significant savings of computation time. This methodology is expected to be useful in the analysis of system response with complex uncertainty sources.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Nuclear Data Impact Assessment for the HTR-10 Pebble-Bed Reactor Using SCALE

The HTR-10 was used as a representative pebble-bed high-temperature gas-cooled reactor in this assessment of nuclear data’s impact on important reactor and spent fuel metrics, including safety-related quantities such as the effective multiplication factor (k eff ), temperature reactivity feedback, spent fuel inventory, and decay heat. Using the SCALE code system tools and ENDF/B-VII.1 nuclear data libraries, we quantify the effect of nuclear data uncertainties on these key performance metrics for both fresh fuel and equilibrium core configurations. For reactor core key parameters, important contributors to uncertainty include reactions of 235 U [$\bar{v}$, fission, (n, γ)], 238 U [elastic, (n, γ)], and graphite [elastic, (n, γ)]. Additional important contributors for the equilibrium core include reactions of higher actinides ( 239 Pu, 240 Pu) and fission products ( 135 Xe, 149 Sm). For spent fuel analysis, most nuclide inventory uncertainties remain below 5%. Higher uncertainties up to 11% are being observed for minor actinides like 243 Am and 244 Cm. Additionally, fission product uncertainties in 155 Eu and 155 Gd, of 25% and 23% respectively, are also significant and have implications for burnup credit applications. 110m Ag also shows high uncertainty of up to 11%, mainly due to fission product yield uncertainties. Decay heat relative uncertainties remain below 0.6% up to 10 years’ cooling time after fuel discharge. The highest relative uncertainty of 1.5% occurs at 500 years of cooling; however, because the decay heat value is very low at that time, the absolute uncertainty is not significant. This work demonstrates that extending assessments beyond fresh fuel k eff to include irradiated cores, nuclide inventories, and decay heat is essential in understanding the behavior of uncertainties as a function of fuel burnup and can support improvements of safety margins and spent fuel management.

Nuclear data impact↗
DOE OSTI2025

Design and Analysis of an Integrated Additively Manufactured Test Article for Plasma-Facing Components

Current plasma-facing components (PFCs) used in helium-cooled divertor modules are complex structures with tungsten tile, steel sleeve components, and cartridges, all assembled in a helium-cooled multiple jet (HEMJ) structure. The goal of this project is to simplify the complex PFC design using additive manufacturing techniques to create a single integrated tungsten test article. Apart from the flexibility this opens up in exploring a wide array of geometries for the article, having a single integrated article significantly reduces the number of joints and parts in the article, thus reducing chances of leaks. A process called electron beam melting has shown to produce very high-density samples and unique geometries, enabling HEMJ or similar designs. To validate and optimize this novel design, the model underwent a series of computational fluid dynamics and finite element analysis simulations to replicate steady-state heat flux in the divertors. The simulations presented in this study consider a steady-state base heat flux of 5 MW/m 2 , with water serving as the coolant. Future research will explore the use of helium as a coolant, simulate edge-localized-mode conditions, and include experimental validation. Since 3D-printed tungsten is anisotropic, the build direction versus build plane of the article are taken into consideration for the test article strength. Because of the high operating temperatures and low ductility of tungsten, thermal creep and brittle fracture are important failure mechanisms to consider. In conclusion, the cap is evaluated with various flow velocities and nozzle diameters, and an optimal design choice is made for which this cap will survive the divertor conditions with a conservative safety margin.

Divertor↗
DOE OSTI2022

Study on Conduction Cooling of Superconducting Magnets for the ILC Main Linac

In the main linac of the International Linear Collider (ILC), superconducting magnets for beam focusing and steering will be located periodically in superconducting RF (SRF) cavity string for beam acceleration in common cryomodules. A concept of conduction cooling of the combined-functioned, splittable superconducting magnets has been proposed and investigated to adapt much different features and to meet different requirements for the superconducting magnet and SRF cavity in fabrication, assembly, and operation. It is required to integrate the superconducting magnet after the SRF cavity string assembly which completed under an ultra-clean environment. The magnet must be conductively cooled down through thermal links to a liquid helium supply pipe. According to this concept, a model magnet development was carried out in cooperation with Fermilab and KEK, and has been demonstrated in KEK superconducting RF test facility (STF). In addition, an important issue has been recently identified. High gradient SRF cavities naturally emit field emission electron flux from the inner surface, so-called dark current. It may pass through the subsequent SRF cavity string and penetrate into the superconducting magnets placed downstream. It may heat up the superconducting coils, and may cause a quench. Therefore, further study on reliable conduction cooling and to secure the superconducting magnet operation with a keeping sufficient safety margin is quite essential. In this paper, we report the installation, the improvement achieved in STF, and the R&D progress in the study on the conduction cooling of the superconducting magnet for the ILC main linac.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

A catastrophic tropical drought kills hydraulically vulnerable tree species

Drought-related tree mortality is now a widespread phenomenon predicted to increase in magnitude with climate change. However, the patterns of which species and trees are most vulnerable to drought, and the underlying mechanisms have remained elusive, in part due to the lack of relevant data and difficulty of predicting the location of catastrophic drought years in advance. In this work, we used long-term demographic records and extensive databases of functional traits and distribution patterns to understand the responses of 20–53 species to an extreme drought in a seasonally dry tropical forest in Costa Rica, which occurred during the 2015 El Niño Southern Oscillation event. Overall, species-specific mortality rates during the drought ranged from 0% to 34%, and varied little as a function of tree size. By contrast, hydraulic safety margins correlated well with probability of mortality among species, while morphological or leaf economics spectrum traits did not. This firmly suggests hydraulic traits as targets for future research.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2020

Convergent evolution of tree hydraulic traits in Amazonian habitats: implications for community assemblage and vulnerability to drought

Amazonian droughts are increasing in frequency and severity. However, little is known about how this may influence species-specific vulnerability to drought across different ecosystem types. We measured 16 functional traits for 16 congeneric species from six families and eight genera restricted to floodplain, swamp, white-sand or plateau forests of Central Amazonia. We investigated whether habitat distributions can be explained by species hydraulic strategies, and if habitat specialists differ in their vulnerability to embolism that would make water transport difficult during drought periods. We found strong functional differences among species. Nonflooded species had higher wood specific gravity and lower stomatal density, whereas flooded species had wider vessels, and higher leaf and xylem hydraulic conductivity. The P 50 values (water potential at 50% loss of hydraulic conductivity) of nonflooded species were significantly more negative than flooded species. However, we found no differences in hydraulic safety margin among species, suggesting that all trees may be equally likely to experience hydraulic failure during severe droughts. Water availability imposes a strong selection leading to differentiation of plant hydraulic strategies among species and may underlie patterns of adaptive radiation in many tropical tree genera. Our results have important implications for modeling species distribution and resilience under future climate scenarios.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

Conceptual Design of Temperature-Controlled Fueled-Salt Irradiation Experiment to Support Demonstration of Advanced Nuclear Reactors

The irradiation testing of fuel-bearing molten salts is critical to supporting the development and demonstration of molten salt reactors (MSRs). These experiments can inform several important reactor design and safety parameters, including source term modeling, evolution of thermophysical properties with burn-up, and the degradation of structural materials under reactor relevant conditions. Idaho National Laboratory is designing an instrumented and heated high temperature molten salt-fueled irradiation capsule to study the behavior of the fuel salt during in-pile irradiation. This paper details the neutronics, thermal, and mechanical analysis performed to-date. Parametric studies are performed to assess a range of material, different experiment dimensions, two in-reactor positions, and the fuel enrichment used. Principal recommendations are to opt for a peripheral reactor position to alleviate neutronic constraints, a thin salt annulus to alleviate thermal constraints, and high-temperature alloys that provide additional safety margins.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗