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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

A User’s Guide to the PLTEMP/ANL Code

PLTEMP/ANL V4.3 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional (Appendix XV). It begins with a solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A User’s Guide to the PLTEMP/ANL Code (V.4.3)

PLTEMP/ANL V4.3 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional (Appendix XV). It should be noted that for a 3-dimensional solution the axial power shape is modeled without accounting for axial heat conduction. It begins with a solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A User’s Guide to the PLTEMP/ANL Code

PLTEMP/ANL V4.4 is a program that obtains a steady-state flow and temperature solution, in geometrical extent, for a nuclear reactor core, or a single fuel assembly, including the effect of manufacturing and modelling uncertainties by the means of hot channel factors. It is based on an evolutionary sequence of codes originally used for calculating plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over the past 30 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability for select design and safety parameters is available to greatly reduce the reactor analyst’s time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Nguyen, Kien-Cuong↗
NASA NTRS1992

A study of forced convection boiling under reduced gravity

This report presents the results of activities conducted over the period 1/2/85-12/31/90, in which the study of forced convection boiling under reduced gravity was initiated. The study seeks to improve the understanding of the basic processes that constitute forced convection boiling by removing the buoyancy effects which may mask other phenomena. Specific objectives may also be expressed in terms of the following questions: (1) what effects, if any, will the removal of body forces to the lowest possible levels have on the forced convection boiling heat transfer processes in well-defined and meaningful circumstances? (this includes those effects and processes associated with the nucleation or onset of boiling during the transient increase in heater surface temperature, as well as the heat transfer and vapor bubble behaviors with established or steady-state conditions); and (2) if such effects are present, what are the boundaries of the relevant parameters such as heat flux, heater surface superheat, fluid velocity, bulk subcooling, and geometric/orientation relationships within which such effects will be produced?

Merte, Herman, Jr.↗
NASA NTRS1990

Design and test of a compact optics system for the pool boiling experiment

The experiment described seeks to improve the understanding of the fundamental mechanisms that constitute nucleate pool boiling. The vehicle for accomplishing this is an investigation, including tests to be conducted in microgravity and coupled with appropriate analyses, of the heat transfer and vapor bubble dynamics associated with nucleation, bubble growth/collapse and subsequent motion, considering the interrelations between buoyancy, momentum and surface tension which will govern the motion of the vapor and surrounding liquid, as a function of the heating rate at the heat transfer surface and the temperature level and distribution in the bulk liquid. The experiment is designed to be contained within the confines of a Get-Away-Special Canister (GAS Can) installed in the bay of the space shuttle. When the shuttle reaches orbit, the experiment will be turned on and testing will proceed automatically. In the proposed Pool Boiling Experiment a pool of liquid, initially at a precisely defined pressure and temperature, will be subjected to a step imposed heat flux from a semitransparent thin-film heater forming part of one wall of the container such that boiling is initiated and maintained for a defined period of time at a constant pressure level. Transient measurements of the heater surface and fluid temperatures near the surface will be made, noting especially the conditions at the onset of boiling, along with motion photography of the boiling process in two simultaneous views, from beneath the heating surface and from the side. The conduct of the experiment and the data acquisition will be completely automated and self-contained. For the initial flight, a total of nine tests are proposed, with three levels of heat flux and three levels of subcooling. The design process used in the development and check-out of the compact photographic/optics system for the Pool Boiling Experiment is documented.

Ling, Jerri S.↗
NASA NTRS1989

Subcooled freon-11 flow boiling in top-heated finned coolant channels with and without a twisted tape

An experimental study was conducted in top-heated finned horizontal tubes to study the effect of enhancement devices on flow boiling heat transfer in coolant channels. The objectives are to examine the variations in both the mean and local (axial and circumferential) heat transfer coefficients for circular coolant channels with spiral finned walls and/or spiral fins with a twisted tape, and improve the data reduction technique of a previous investigator. The working fluid is freon-11 with an inlet temperature of 22.2 C (approximately 21 C subcooling). The coolant channel's exit pressure and mass velocity are 0.19 M Pa (absolute) and 0.21 Mg/sq. ms, respectively. Two tube configurations were examined; i.e., tubes had either 6.52 (small pitch) or 4.0 (large pitch) fins/cm of the circumferential length (26 and 16 fins, respectively). The large pitch fins were also examined with a twisted tape insert. The inside nominal diameter of the copper channels at the root of the fins was 1.0 cm. The results show that by adding enhancement devices, boiling occurs almost simultaneously at all axial locations. The case of spiral fins with large pitch resulted in larger mean (circumferentially averaged) heat transfer coefficients, h sub m, at all axial locations. Finally, when twisted tape is added to the tube with large-pitched fins, the power required for the onset of boiling is reduced at all axial and circumferential locations.

Smith, Alvin↗
DOE OSTI2023

A non-fueled nuclear-heated rod for in-pile transient boiling studies

Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10 B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (B nat ~ 2.05 wt %) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-a thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. In conclusion, these results give confidence that the BNHR design has successfully met experiment objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Predicting critical heat flux using localized sensing at invisible vapor-liquid interfaces

Predicting critical heat flux (CHF) in two-phase electronics cooling systems remains a significant challenge due to the sudden onset of boiling crisis and the difficulty in directly visualizing vapor-liquid interfaces. Existing sensing methods rely on lagging temperature measurements, optically accessible systems, or spatially averaged signals that cannot pinpoint CHF initiation at localized high-heat-flux regions. Here, we report a planar capacitive sensing approach that enables real-time, localized detection of vapor-liquid interface dynamics for CHF prediction in boiling heat transfer. The capacitive sensor exploits the dielectric constant difference between liquid and vapor phases to capture bubble nucleation, growth, and departure dynamics with a temporal resolution down to 2 ms. The capacitive sensing reveals distinct signals across boiling regimes: from high-frequency fluctuations during strong nucleate boiling to low-frequency fluctuations with increased amplitudes when approaching CHF. The multi-sensor array experiments demonstrate real-time localized sensing, where each sensor responds exclusively to boiling in its immediate vicinity without crosstalk from neighboring regions. This non-intrusive sensing approach provides predictive rather than lagging sensing signals of CHF occurrence, offering predictive diagnosis of two-phase liquid cooling for the thermal management of high-power-density electronics.

CHF↗
DOE OSTI2021

Transient heat transfer of superfluid He 4 in nonhomogeneous geometries: Second sound, rarefaction, and thermal layer

Transient heat transfer in superfluid 4 He (He II) is a complex process that involves the interplay of the unique counterflow heat-transfer mode, the emission of second-sound waves, and the creation of quantized vortices. Many past researches focused on homogeneous heat transfer of He II in a uniform channel driven by a planar heater. Here, we report our systematic study of He II transient heat transfer in nonhomogeneous geometries that are pertinent to emergent applications. By solving the He II two-fluid equations of motion coupled with Vinen's equation for vortex-line density, we examine and compare the characteristics of transient heat transfer from planar, cylindrical, and spherical heaters in He II. Our results show that as the heater turns on, an outgoing second-sound pulse emerges, within which the vortex-line density grows rapidly. These vortices attenuate the second sound and result in a heated He II layer in front of the heater, i.e., the thermal layer. In the planar case where the vortices are created throughout the space, the second-sound pulse is continuously attenuated, leading to a thick thermal layer that diffusely spreads following the heat pulse. On the contrary, in the cylindrical and the spherical heater cases, vortices are created mainly in a thin thermal layer near the heater surface. As the heat pulse ends, a rarefaction tail develops following the second-sound pulse, in which the temperature drops. This rarefaction tail can promptly suppress the thermal layer and take away the deposited thermal energy. The effects of the heater size, heat flux, pulse duration, and temperature on the thermal-layer dynamics are discussed. We also show how the peak heat flux for the onset of boiling in He II can be studied in our model.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
NASA NTRS2015

Numerical Simulation of Liquid Nitrogen Chilldown of a Vertical Tube

This paper presents the results of a one-dimensional numerical simulation of the transient chilldown of a vertical stainless steel tube with liquid nitrogen. The direction of flow is downward (with gravity) through the tube. Heat transfer correlations for film, transition, and nucleate boiling, as well as critical heat flux, rewetting temperature, and the temperature at the onset of nucleate boiling were used to model the convection to the tube wall. Chilldown curves from the simulations were compared with data from 55 recent liquid nitrogen chilldown experiments. With these new correlations the simulation is able to predict the time to rewetting temperature and time to onset of nucleate boiling to within 25% for mass fluxes ranging from 61.2 to 1150 kg/(sq m s), inlet pressures from 175 to 817 kPa, and subcooled inlet temperatures from 0 to 14 K below the saturation temperature.

GFSSP↗
DOE OSTI2023

Development of experimental and computational frameworks to predict subcooled flow boiling in the LANL Isotope Production Facility

Cooling is crucial to maintain the integrity of target systems in isotope production facilities. At Los Alamos National Laboratory (LANL)’s Isotope Production Facility (IPF), multiple encapsulated targets are stacked and irradiated in tandem with a 100 MeV, ~250μA proton beam. To facilitate effective heat removal, these stacked targets are separated and cooled via a series of water channels. At these beam currents, this high-energy proton beam heats the target system, likely initiating subcooled flow boiling in the cooling channels. However, in-beam monitoring of the IPF target system is not possible due to the extreme radiation environment, and the necessarily significant shielding. To better understand high-power target performance, we developed ex-situ experimental and computational frameworks to predict the behavior of subcooled flow boiling at IPF. Subcooled flow boiling experiments on Inconel 625 samples under IPF conditions (2 bar pressure, 10 GPM flow rate (i.e., 2249 kg/m 2 /s), 85 K subcooling) revealed that IPF's average operating power is at the early stage of boiling with a heat transfer coefficient of 48,000 W/m 2 /s. The proposed modeling framework enables us to predict a complete boiling curve, i.e., single-phase heat transfer, onset of nucleate boiling, two-phase heat transfer, and critical heat flux (CHF), with specification of input boiling parameters up to intermediate heat flux levels. The estimated CHF under IPF conditions is 5.2 MW/m 2 . Experimental data under reduced conditions (2 bar pressure, 1.5 GPM flow rate (i.e., 337 kg/m 2 /s), 45 K subcooling) served as validation cases for the computational modeling. This computational model can be further extended to more complicated systems replicating the real IPF configuration, for instance, to study void distribution as a function of the incident proton beam profile and coolant velocity profile of multiple cooling channels. Finally, the proposed experimental and computational frameworks provide a means to better understand cooling systems in the isotope production facilities at different accelerators, where in-beam monitoring of the cooling process is not available.

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS2023

Universal Two-Phase Convection Heat Transfer Correlations for Cryogenic Pipe Chilldown

This paper presents a set of universal two-phase convection heat transfer correlations for modeling boiling heat transfer during pipe chilldown fit over the widest available range of cryogenic fluids and thermodynamic conditions. The correlations improve upon prior correlations that were developed separately for liquid nitrogen (LN2) and liquid hydrogen (LH2) pipe quenching datasets. The new correlations include equations to calculate the single-phase vapor heat transfer, film boiling heat transfer, transition boiling heat transfer, nucleate boiling heat transfer, single-phase liquid heat transfer, bulk vapor temperature during high quality film boiling, Leidenfrost temperature, critical heat flux, critical heat flux temperature, and the onset of nucleate boiling temperature. The correlations were validated against LH2, LN2, liquid methane, liquid oxygen, and liquid argon pipe quenching datasets. The eight datasets cover the following parameter ranges: pipe lengths of 0.1 to 6.5m; outer pipe diameters of 12.7 to 25.4 mm; pipe wall thicknesses from 0.51 to 1.64 mm; flow directions of upward, downward, and horizontal; gravity levels of 1g and 0g±0.01g. A numerical model assumes homogeneous mixing between the vapor and liquid and implicitly integrates the coupled energy equations for the pipe and fluid, as well as continuity for the fluid. The model was used to estimate the equilibrium quality and fluid mass accumulation along the pipes during chilldown from estimates of the fluid-pipe heat transfer extracted from temperature measurements. The correlations can be implemented in lumped parameter codes such as SINDA/FLUINT and the Generalized Fluid System Simulation Program (GFSSP) to improve the accuracy of chilldown time and chilldown boiloff mass predictions. Such predictions are useful for designing ground or in-space cryogenic liquid transfer systems.

Universal Correlations↗
NASA NTRS2000

Heat Transfer Performances of Pool Boiling on Metal-Graphite Composite Surfaces

Nucleate boiling, especially near the critical heat flux (CHF), can provide excellent economy along with high efficiency of heat transfer. However, the performance of nucleate boiling may deteriorate in a reduced gravity environment and the nucleate boiling usually has a potentially dangerous characteristic in CHF regime. That is, any slight overload can result in burnout of the boiling surface because the heat transfer will suddenly move into the film-boiling regime. Therefore, enhancement of nucleate boiling heat transfer becomes more important in reduced gravity environments. Enhancing nucleate boiling and critical heat flux can be reached using micro-configured metal-graphite composites as the boiling surface. Thermocapillary force induced by temperature difference between the graphite-fiber tips and the metal matrix, which is independent of gravity, will play an important role in bubble detachment. Thus boiling heat transfer performance does not deteriorate in a reduced-gravity environment. Based on the existing experimental data, and a two-tier theoretical model, correlation formulas are derived for nucleate boiling on the copper-graphite and aluminum-graphite composite surfaces, in both the isolated and coalesced bubble regimes. Experimental studies were performed on nucleate pool boiling of pentane on cooper-graphite (Cu-Gr) and aluminum-graphite (Al-Gr) composite surfaces with various fiber volume concentrations for heat fluxes up to 35 W per square centimeter. It is revealed that a significant enhancement in boiling heat transfer performance on the composite surfaces is achieved, due to the presence of micro-graphite fibers embedded in the matrix. The onset of nucleate boiling (the isolated bubble regime) occurs at wall superheat of about 10 C for the Cu-Gr surface and 15 C for the Al-Gr surface, much lower than their respective pure metal surfaces. Transition from an isolated bubble regime to a coalesced bubble regime in boiling occurs at a superheat of about 14 C on Cu-Gr surface and 19 C on Al-Gr surface.

Zhang, Nengli↗
NASA NTRS2023

Effects of Heating Configuration and Operating Parameters on Heat Transfer and Interfacial Physics of Microgravity Flow Boiling With Subcooled Inlet Conditions –Experiments Onboard the International Space Station

This study is part of the Flow Boiling and Condensation Experiment (FBCE), a collaborative effort between the Purdue University Boiling and Two-Phase Flow Laboratory (PU-BTPFL) and the NASA Glenn Research Center. The FBCE fitted with the Flow Boiling Module (FBM) was launched to the International Space Station (ISS) in August 2021 and experiments were successfully performed from February to July 2022 to amass a large microgravity-flow-boiling database. This study is focused on heat transfer and flow visualization of microgravity flow boiling of n-Perfluorohexane in a rectangular channel of 5.0 mm height, 2.5 mm width (heated), and 114.6 mm length, with subcooled inlet conditions. High-speed-video photography is utilized to present flow patterns and temporal interfacial behavior. Heat transfer results are presented in the form of flow boiling curves and both parametric curves and streamwise profiles of wall temperature and heat transfer coefficient. Firstly, the parametric effects of mass velocity (199.4 – 3200.0 kg/m 2 s), inlet subcooling (0.2 – 46.0°C), and inlet pressure (124.2 – 176.7 kPa), on the aforementioned aspects are assessed for double-sided heating to establish them for a microgravity environment. Of these three parameters, mass velocity and inlet subcooling mostly determine the microgravity flow boiling behavior, while inlet pressure plays an insignificant role. Flow patterns for double-sided heating are more complex than those for single-sided heating due to interaction between the two vapor layers. Vapor interaction is minimized at high subcoolings and high mass velocities due to strong condensation offered by the subcooled bulk liquid layer separating them. Despite the different flow patterns, both single- and double-sided heating generally result in similar parametric trends and local heat transfer coefficients for similar operating conditions. Flow instabilities manifest as temporal flow anomalies and temperature oscillations, and their severity increases with increasing boiling number. Secondly, the effects of heating configuration are analyzed by comparing and contrasting several aspects of single- and double-sided heating data. The heat fluxes at which onset of nucleate boiling degradation (ONBD) and critical heat flux (CHF) occur are distinctly different for single- and double-sided heating. There exists a threshold inlet subcooling demarcating the dominance of flow acceleration and condensation effects in vapor removal from the near-wall region and replenishment of fresh liquid for boiling. Above the threshold, condensation from the near-wall region is dominant and single-sided heating yields higher heat fluxes, and below it, acceleration is dominant and double-sided yields higher heat fluxes. At mass velocity in the range of 200 – 2400 kg/m 2 s, the threshold inlet subcooling lies in the approximate range of 20 – 30°C (corresponding inlet quality of roughly -0.40 – -0.20).

Microgravity↗
DOE OSTI2025

Impact of U-10Mo HALEU fuel element tolerances on the Massachusetts Institute of Technology reactor safety and operational performance – Thermal hydraulics

The U.S. is coordinating efforts for the conversion of six U.S. High Performance Research Reactors (USHPRR) including one critical facility from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts of Institute of Technology Reactor (MITR), and achieve similar performance, high-assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being evaluated. The impact of the fabrication specification and tolerances was assessed following the preliminary design of the MITR LEU fuel elements using the U-10Mo monolithic alloy. This research focuses on the analysis of fabrication specification impact on thermal hydraulics (TH) characteristic of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., coolant channel gap thickness, fuel plate thickness, etc.). The analyses are performed based on an all-fresh LEU fuel conversion plan identified in a preliminary safety analysis report submitted to the Nuclear Regulatory Commission. The reactor power margin to the onset of nucleate boiling (ONB) is assessed under the limiting safety system settings (LSSS), where a scram occurs, to ensure there is sufficient margin to the reactor safety limit, which is defined by the onset of flow instability that occurs after the ONB. The best estimate plus uncertainty approach is employed to analyze this TH characteristic, which yields realistic results while maintaining adequate conservatism, utilizing a statistical uncertainty propagation method with the STAT7 code. The TH characteristic is analyzed as a function of the variability of the specification parameters resulting from the fabrication process. The main findings of this study show that the MITR core can meet the TH safety and operational requirements at the all-LEU initial core startup (cycle 1), selected transition cycles (most reactive cycle and most limiting cycle: cycle 3 and 5, respectively) and equilibrium (cycle 14) cores under all limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core power margin to ONB on those specification parameters that have the most direct impact on TH performance is non-linear but monotonically decreasing within the specification tolerances. The third order polynomial fit curves are reported in detail for selected limiting cases and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.

Conversion↗
NASA NTRSDate not supplied

Experimental Studies on the Effect of Inlet Liquid Subcooling on the Chilldown Characteristics of Stainless Steel Tubes

Understanding transfer line chilldown process under microgravity is important for the efficient transfer of cryogenic propellants in space fuel depots to facilitate future long duration space missions. The present work is part of the ongoing efforts to develop and test a two-phase flow chilldown test section to study the complete chilldown process under sustained microgravity conditions onboard the International Space Station. In this study, ground-based chilldown experiments are carried out on a 60 cm long SS-316 test section with PF-5060 as the working fluid. The complete chilldown curve was obtained including the film, transition and nucleate boiling regimes along with the temperature transition points. The effect of inlet liquid subcooling on the behavior of the chilldown curves are presented. Further, the chilldown and heat flux curves are analyzed to obtain re-wetting/Leidenfrost and onset of nucleate boiling temperature transition points as well as the critical heat flux values. The effect of inlet liquid subcooling on regime-specific heat flux and heat transfer coefficients are also examined.

CFD↗
NASA NTRS1973

Transient boiling heat transfer in saturated liquid nitrogen and F113 at standard and zero gravity

Transient and steady state nucleate boiling in saturated LN2 and F113 at standard and near zero gravity conditions were investigated for the horizontal up, vertical and horizontal down orientations of the heating surface. Two distinct regimes of heat transfer mechanisms were observed during the interval from the step increase of power input to the onset of nucleate boiling: the conduction and convection dominated regimes. The time duration in each regime was considerably shorter with LN2 than with F113, and decreased as heat flux increased, as gravity was reduced, and as the orientation was changed from horizontal up to horizontal down. In transient boiling, boiling initiates at a single point following the step increase in power, and then spreads over the surface. The delay time for the inception of boiling at the first site, and the velocity of spread of boiling varies depending upon the heat flux, orientation, body force, surface roughness and liquid properties, and are a consequence of changes in boundary layer temperature levels associated with changes in natural convection. Following the step increase in power input, surface temperature overshoot and undershoot occur before the steady state boiling temperature level is established.

Oker, E.↗