1,720,995 research outputs found
Evaluation of the Thermal-Hydraulic Operating Limits of HEU-LEU Transition Cores for the MIT Research Reactor
No full textThe MIT Research Reactor (MITR) is in the process of conducting a design study to
convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU)
fuel. The currently selected LEU fuel design contains 18 plates per element, compared to
the existing HEU design of 15 plates per element. A transitional conversion strategy,
which consists of replacing three HEU elements with fresh LEU fuel elements in each
fuel cycle, is proposed. The objective of this thesis is to analyze the thermo-hydraulic
safety margins and to determine the operating power limits of the MITR for each mixed
core configuration.
The analysis was performed using PLTEMP/ANL ver 3.5, a program developed
for thermo-hydraulic calculations of research reactors. Two correlations were used to
model the friction pressure drop and enhanced heat transfer of the finned fuel plates: the
Carnavos correlation for friction factor and heat transfer, and the Wong Correlation for
friction factor with a constant heat transfer enhancement factor of 1.9. With these
correlations, the minimum onset of nucleate boiling (ONB) margins of the hottest fuel
plates were evaluated in nine different core configurations, the HEU core, the LEU core
and seven mixed cores that consist of both HEU and LEU elements. The maximum radial
power peaking factors were assumed at 2.0 for HEU and 1.76 for LEU in all the analyzed
core configurations.
The calculated results indicate that the HEU fuel elements yielded lower ONB
margins than LEU fuel elements in all mixed core configurations. In addition to full
coolant channels, side channels next to the support plates that form side coolant channels were analyzed and found to be more limiting due to higher flow resistance. The
maximum operating powers during the HEU to LEU transition were determined by
maintaining the minimum ONB margin corresponding to the homogeneous HEU core at
6 MW. The recommended steady-state power is 5.8 MW for all transitional cores if the
maximum radial peaking is adjacent to a full coolant channel and 4.9 MW if the
maximum radial peaking is adjacent to a side coolant channel.United States. Dept. of Energy. Reduced Enrichment Research and Test Reactor Progra
Pressure Drop Measurements and Flow Distribution Analysis for MIT Research Reactor with Low-Enriched Fuel
No full textThe MIT Nuclear Research Reactor (MITR) is the only research reactor in the United
States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer.
Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR,
together with the supporting thermal hydraulic analyses, propose different fuel element
designs for optimization of thermal hydraulic performance of the LEU core. Since
proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the
friction pressure drop is required to be verified experimentally.
The objectives of this study are to measure the friction coefficient in both laminar and
turbulent flow regions, and to develop empirical correlations for the finned rectangular
coolant channels for the safety analysis of the MITR. A friction pressure drop experiment
is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is
measured for both flat and finned coolant channels of various channel heights. Experiment
data show that the Darcy friction factors for laminar flow in finned rectangular channels
are in good agreement with the existing correlation if a pseudo-smooth equivalent
hydraulic diameter is considered; whereas a new friction factor correlation is proposed for
the friction factors for turbulent flow. Additionally, a model is developed to calculate the
primary flow distribution in the reactor core for transitional core configuration with
various combinations of HEU and LEU fuel elements.United States. Dept. of Energy. Reduced Enrichment for Research and Test Reactors Progra
Design and Optimization of a High Thermal Flux Research Reactor Via Kriging-Based Algorithm
No full textIn response to increasing demands for the services of research reactors, a 5 MW LEUfueled
research reactor core is developed and optimized to provide high thermal flux
within specified limits upon thermal hydraulic performance, cycle length, irradiation
utilization, and manufacturability.
A novel fuel assembly concept which makes use of integral flux traps is postulated to
meet these requirements. Each assembly can be rotated into one of three different
configurations to produce flux traps of different size, shape, and neutron energy spectrum
within the core.
A method for predicting and guiding the search for the optimum geometry was sought.
Kriging has been chosen to predict the values of eigenvalue and thermal flux at untested
geometric parameters. Because kriging treats all measurements as the sum of a global
deterministic function and a stochastic departure from that function, predictions come
with a measurement of uncertainty. As a result, the analyst can search the design space
for likely improvement, or probe areas of high uncertainty for improvements that might
have been missed using other methods. The technique is used in an algorithm for
constrained optimization of the design, and a set of best practices for use of this are
described.
The optimized design produces a peak thermal flux of 1.56 x 10[superscript 14] n/cm[superscript 2]s. Safety is demonstrated by presentation of reactivity feedback coefficients and the results of loss of flow and reactivity insertion transient analysis.
A single fission target can be used to produce 96 6-day Ci of [superscript 99]Mo per week. When the reactor is oriented to take advantage of high fast flux, steels can be subjected to damage
rates of 5.76 dpa per year. Silicon carbide can be damaged at a rate of 2.79 dpa/y. The
concept is a safe, versatile, proliferation-resistant means of supplying current and future
irradiation needs
Critical Heat Flux Enhancement via Surface Modification Using Colloidal Dispersions of Nanoparticles (Nanofluids)
No full textNanofluids are engineered colloidal dispersions of nanoparticles (1-100nm) in common fluids
(water, refrigerants, or ethanol…). Materials used for nanoparticles include chemically stable
metals (e.g., gold, silver, copper), metal oxides (e.g., alumina, zirconia, silica, titania) and carbon
in various forms (e.g., diamond, graphite, carbon nanotubes). The attractive properties of
nanofluids include higher thermal conductivity, heat transfer coefficients (HTC) and boiling
critical heat flux (CHF) than that of the respective base fluid. Nanofluids have been found to
exhibit a very significant enhancement up to 200% of the boiling CHF at low nanoparticle
concentrations.
In this study, nanofluids were investigated as an agent to modify a heater surface to enhance
Critical Heat Flux (CHF). First, the CHF of diamond, Zinc Oxide and Alumina water-based
nanofluids at low volume concentration (<1 vol%) were measured to determine if nanofluid
enhances CHF as seen in literature. Subsequently, the heaters are coated with nanoparticles via
nucleate boiling of nanofluids. The CHF of water was measured using these nanoparticle
precoated heaters to determine the magnitude of the CHF enhancement. Characterization of the
heaters after CHF experiments using SEM, confocal, and contact angle were conducted to
explain possible mechanisms for the observed enhancement. The coating thickness of the
nanoparticle deposition on a wire heater as a function of boiling time was also investigated.
Finally, theoretical analyses of the maximum CHF and HTC enhancement in term of wettability
were performed and compared with the experimental data.
The CHF of nanofluids was as much as 85% higher than that of water, while the nanoparticle
pre-coated surfaces yielded up to 35% CHF enhancement compared to bare heaters. Surface
characterization of the heaters after CHF experiments showed a change in morphology due to the
nanoparticles deposition. The coating thickness of nanoparticle was found to deposit rather
quickly on the wire surface. Within five minutes of boiling, the coating thickness of more than 1
μm was achieved. Existing CHF correlations overestimated the experimental data.Electric Power Research Institute (Contract EPP24367/C11807
Thermal Hydraulic Analysis of a Low Enrichment Uranium Core for the MIT Research Reactor
No full textThe MIT research reactor (MITR) is converting from the existing high enrichment
uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density
monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is
evolving. The objectives of this study are to benchmark the in-house computer code for
the MITR, and to perform thermal hydraulic analyses in support of the LEU design
studies. The in-house multi-channel thermal-hydraulics code, MULCH-II, was developed
specifically for the MITR. This code was validated against PLTEMP for steady-state
analysis, and against RELAP5 and temperature measurements for the loss of primary
flow transient. The benchmark analysis results showed that the MULCH-II code is in
good agreement with other computer codes and experimental data, and hence it is used as
the main tool for this study.
Various fuel configurations are evaluated as part of the LEU core design optimization
study. The criteria adopted for the LEU thermal hydraulics analysis in this study of the
limiting safety system settings (LSSS), are to prevent onset of nucleate boiling during
steady-state operation, and to avoid a clad temperature excursion during the loss of flow
transient.
In ranking the LEU core design options, the primary parameter is a low power peaking
factor in order to increase the LSSS power and to decrease the maximum clad
temperature during the transient. The LEU fuel designs with 15 to 18 plates per element,
fuel thickness of 20 mils, and a hot channel factor less than 1.76 are shown to comply
with the thermal-hydraulic criteria. The steady-state power can potentially be higher than
6 MW, which was requested in the power upgrade submission to the Nuclear Regulatory
Commission.United States. Dept. of Energy. (Reduced Enrichment for Research and Test Reactors Program
Convective Heat Transfer Enhancement in Nanofluids: Real Anomaly or Analysis Artifact?
No full textThe nanofluid literature contains many claims of anomalous convective heat transfer enhancement in both turbulent and laminar flow. To put such claims to the test, we have performed a critical detailed analysis of the database reported in 12 nanofluid papers (8 on laminar flow and 4 on turbulent flow). The methodology accounted for both modeling and experimental uncertainties in the following way. The heat transfer coefficient for any given data set was calculated according to the established correlations (Dittus-Boelter's for turbulent flow and Shah's for laminar flow). The uncertainty in the correlation input parameters (i.e., nanofluid thermo-physical properties and flow rate) was propagated to get the uncertainty on the predicted heat transfer coefficient. The predicted and measured heat transfer coefficient values were then compared to each other. If they differed by more than their respective uncertainties, we judged the deviation anomalous. According to this methodology, it was found that in nanofluid laminar flow in fact there seems to be anomalous heat transfer enhancement in the entrance region, while the data are in agreement (within uncertainties) with the Shah's correlation in the fully developed region. On the other hand, the turbulent flow data could be reconciled (within uncertainties) with the Dittus-Boelter's correlation, once the temperature dependence of viscosity was included in the prediction of the Reynolds number. While this finding is plausible, it could not be conclusively confirmed, because most papers do not report information about the temperature dependence of the viscosity for their nanofluids
Nanofluid heat transfer enhancement for nuclear reactor applications
No full textColloidal dispersions of nanoparticles are known as `nanofluids'. Such engineered fluids offer the potential for enhancing heat transfer, particularly boiling heat transfer, while avoiding the drawbacks (i.e., erosion, settling, clogging) that hindered the use of particle-laden fluids in the past. At MIT we have been studying the heat transfer characteristics of nanofluids for the past five years, with the goal of evaluating their benefits for and applicability to nuclear power systems (i.e., primary coolant, safety systems, severe accident mitigation strategies). This paper will summarize the MIT research in this area with particular emphasis to boiling behavior, including, prominently, the Critical Heat Flux limit and quenching phenomena.United States. Dept. of EnergyElectric Power Research InstituteNational Science Foundation (U.S.
Effects of Micro/Nano-Scale Surface Characteristics on the Leidenfrost Point Temperature of Water
No full textIn recent film boiling heat transfer studies with nanofluids, it was reported that deposition of nanoparticles on a surface significantly increases the nominal minimum heat flux (MHF) or Leidenfrost Point (LFP) temperature, considerably accelerating the transient cooling of overheated objects. It was suggested that the thin nanoparticle deposition layer and the resulting changes in the physico-chemical characteristics of the hot surface, such as surface roughness height, wettability and porosity, could greatly affect quenching phenomena. In this study, a set of water-droplet LFP tests are conducted using custom-fabricated surfaces which systemically separate the effects of surface roughness height (0-15 um), wettability (0-83°) and nanoporosity (∼23 nm). In addition, high-speed imaging of the evaporating droplets is used to explore the influence of these surface characteristics on the intermittent solid-liquid contacts in film boiling. The obtained results reveal that nanoporosity (not solely high surface wettability) is the crucial feature in efficiently increasing the LFP temperature by initiating heterogeneous nucleation of bubbles during short-lived solid-liquid contacts, which results in disruption of the vapor film, and that micro-posts on the surface intensify such effects by promoting intermittent liquid-surface contacts.United States. Dept. of Energy. Office of Nuclear Energy (NEUP Fellowship Program
Convective Heat Transfer in a High Aspect Ratio Minichannel Heated on One Side
No full textExperimental results are presented for single-phase heat transfer in a narrow rectangular minichannel heated on one side. The aspect ratio and gap thickness of the test channel were 29:1 and 1.96 mm, respectively. Friction pressure drop and Nusselt numbers are reported for the transition and fully turbulent flow regimes, with Prandtl numbers ranging from 2.2 to 5.4. Turbulent friction pressure drop for the high aspect ratio channel is well-correlated by the Blasius solution when a modified Reynolds number, based upon a laminar equivalent diameter, is utilized. The critical Reynolds number for the channel falls between 3500 and 4000, with Nusselt numbers in the transition regime being reasonably predicted by Gnielinski's correlation. The dependence of the heat transfer coefficient on the Prandtl number is larger than that predicted by circular tube correlations, and is likely a result of the asymmetric heating. The problem of asymmetric heating condition is approached theoretically using a boundary layer analysis with a two-region wall layer model, similar to that originally proposed by Prandtl. The analysis clarifies the influence of asymmetric heating on the Nusselt number and correctly predicts the experimentally observed trend with Prandtl number. A semi-analytic correlation is derived from the analysis that accounts for the effect of aspect ratio and asymmetric heating, and is shown to predict the experimental results of this study with a mean absolute error (MAE) of less than 5% for 4000 < Re < 70,000.United States. National Nuclear Security Administration. Global Threat Reduction Initiative (Argonne National Laboratory. Contract 25-30101-0004 A)United States. National Nuclear Security Administration. Office of Nonproliferation and International Security (Nuclear Nonproliferation Safeguards Graduate Fellowship)United States. National Nuclear Security Administration. (Sandia National Laboratories. Contract DE-AC04-94AL85000
Effects of Surface Parameters on Boiling Heat Transfer Phenomena
No full textNanofluids, engineered colloidal dispersions of nanoparticles in fluid, have been shown
to enhance pool and flow boiling CHF. The CHF enhancement was due to nanoparticle
deposited on the heater surface, which was verified in pool boiling. However, no such
work has been done for flow boiling. Using a cylindrical tube pre-coated with Alumina
nanoparticles coated via boiling induced deposition, CHF of water was found to enhance
up to 40% compared to that of the bare tube. This confirms that nanoparticles on the
surface is responsible for CHF enhancement for flow boiling. However, existing theories
failed to predict the CHF enhancement and the exact surface parameters attributed to the
enhancement cannot be determined.
Surface modifications to enhance critical heat flux (CHF) and Leidenfrost point (LFP)
have been shown successful in previous studies. However, the enhancement mechanisms
are not well understood, partly due to many surface parameters being altered at the same
time, as in the case for nanofluids. Therefore, the remaining objective of this work is to
evaluate separate surface effect on different boiling heat transfer phenomena.
In the second part of this study, surface roughness, wettability and nanoporosity were
altered one by one and respective effect on quenching LFP with water droplet was
determined. Increase in surface roughness and wettability enhanced LFP; however,
nanoporosity was most effective in raising LFP, almost up to 100ºC. The combination of
the micro posts and nanoporous coating layer proved optimal. The nanoporous layer
destabilizes the vapor film via heterogeneous bubble nucleation, and the micro posts
provides intermittent liquid-surface contacts; both mechanisms increase LFP.
In the last part, separate effect of nanoporosity and surface roughness on pool boiling
CHF of a well-wetting fluid, FC-72, was investigated. Nanoporosity or surface roughness
alone had no effect on pool boiling CHF of FC-72. Data obtained in the literature mostly
for microporous coatings showed CHF enhancement for well wetting fluids, and existing
CHF models are unable to predict the enhancement.Electric Power Research InstituteU.S. Nuclear Regulatory Commissio
- …
