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Articles 1 - 30 of 1441

Full-Text Articles in Nuclear Engineering

Isotopically-Resolved Neutron Cross Sections As Probe Of The Nuclear Optical Potential, Cole Davis Pruitt May 2019

Isotopically-Resolved Neutron Cross Sections As Probe Of The Nuclear Optical Potential, Cole Davis Pruitt

Arts & Sciences Electronic Theses and Dissertations

Neutron scattering experiments provide direct access to the forces experienced by nucleons in the nuclear environment. Due to the experimental difficulty of cross section measurements with neutrons, isotopically-resolved neutron scattering cross sections are sorely needed as inputs for many nuclear models. This dissertation presents the results from a campaign of isotope-specific neutron total cross section measurements on 16,18O, 58,64Ni, 112,124Sn, and 103Rh from 3-450 MeV and elastic scattering differential cross section measurements on 112,nat,124Sn at 11 and 17 MeV. Equipped with these new data and with computational improvements to the Dispersive Optical Model (DOM), we ...


Tertiary Safety System For Nuclear Spent Fuel Pool, Jonathan Farmer, Amanda Bachmann, Taylor Adams, Eissa Altalahlah, Trina Garrett, Jillian Newmyer, Drew Shayotovich May 2019

Tertiary Safety System For Nuclear Spent Fuel Pool, Jonathan Farmer, Amanda Bachmann, Taylor Adams, Eissa Altalahlah, Trina Garrett, Jillian Newmyer, Drew Shayotovich

Chancellor’s Honors Program Projects

No abstract provided.


Calculation And Tabulation Of Efficiencies For Tungsten Foil Positron Moderators, R. Alsulami, M. Albarqi, S. Jaradat, Shoaib Usman, Joseph T. Graham May 2019

Calculation And Tabulation Of Efficiencies For Tungsten Foil Positron Moderators, R. Alsulami, M. Albarqi, S. Jaradat, Shoaib Usman, Joseph T. Graham

Mining and Nuclear Engineering Faculty Research & Creative Works

Monte Carlo radiation transport simulations were used to calculate the positron stopping profiles in tungsten positron moderator foils. Stopping profiles were numerically integrated with efficiency kernels derived from Green's function solutions of the 3D diffusion equation to determine the moderation efficiency in both the backscattering and transmission geometries. Stopping profiles and efficiencies were calculated for positron energies from 10 keV to 10 MeV and incident angles from 0° to 75°. The resulting efficiencies agreed with other calculated and measured values in the literature, especially when similar values of the positron diffusion length and surface emission branching ratio were used ...


Manufacture Of Dual Sided Microstructured Semiconductor Neutron Detectors, Jared Medina Apr 2019

Manufacture Of Dual Sided Microstructured Semiconductor Neutron Detectors, Jared Medina

Kansas State University Undergraduate Research Conference

The world is in need of a new way to detect neutrons. The best current detectors rely on 3He, which is in short supply. The 3He detectors are extremely expensive. The goal of this project is to produce inexpensive and robust detectors that do not rely on 3He. Instead of using gas, the Dual Sided Microstructured Neutron Detectors (DS-MSNDs) are made from a semiconductor material. The DS-MSNDs have been simulated to have up to 70% efficiency, which is comparable to the 3He detectors efficiency of about 80%. The DS-MSNDs have micro-trenches that are back filled with ...


Investigation Of The Coupled Nuclear, Thermal-Hydraulic, And Thermomechanical Response Of A Natural Circulation Research Reactor Under Severe Reactivity-Initiated Accident Transients, Darren G. Talley Mar 2019

Investigation Of The Coupled Nuclear, Thermal-Hydraulic, And Thermomechanical Response Of A Natural Circulation Research Reactor Under Severe Reactivity-Initiated Accident Transients, Darren G. Talley

Mechanical Engineering ETDs

Research reactors play an important role in higher education, scientific research, and medical radioisotope production around the world. It is thus important to ensure the safety of facility workers and the public. This work presents a new reactor transient analysis code, referred to as Razorback, which computes the coupled reactor kinetics, fuel element heat transfer, fuel element thermal expansion and thermal stress, and thermal-hydraulic response of a natural circulation research reactor. The code was developed for the evaluation of large rapid reactivity addition in research reactors, with an initial focus on the Annular Core Research Reactor (ACRR) at Sandia National ...


Developing A Simulation Tool For Evaluating In-Motion Detector Systems (La-Ur-18-29139), Brian Jennings Mar 2019

Developing A Simulation Tool For Evaluating In-Motion Detector Systems (La-Ur-18-29139), Brian Jennings

Nuclear Engineering ETDs

Available simulation software lacks the ability to produce in-motion detector responses for detector systems that may be used to detect the illicit trafficking of nuclear materials. In this thesis, a simulation tool is developed that uses static measured data as a basis set for producing in-motion detector responses with the ability to vary many parameters in each simulated trial. Once the basis set is measured and loaded into the simulation tool, the user interface allows the user to enter variations to speed, source height, source-to-detector distance, background exposure rate, which source(s) are present, their relative strength and shielding configuration ...


Development Of Wireless Pebble For Packed Bed Heat Transfer Measurements And Machine Learning-Aided Accident Diagnosis For Loss Of Flow Accident (Lofa), Dongjune Chang Jan 2019

Development Of Wireless Pebble For Packed Bed Heat Transfer Measurements And Machine Learning-Aided Accident Diagnosis For Loss Of Flow Accident (Lofa), Dongjune Chang

Nuclear Engineering ETDs

In the first study, a novel wireless pebble for scale experiments is developed, and a simple heat transfer experiment is conducted to determine the difference in the local heat transfer coefficient. Based on the fact that the use of Dowtherm A between approximately 57–87 °C is an alternative to the normal use of the FliBe temperature range of 600–700°C, a new-concept wireless device in a scaled experiment is introduced. This device consists of a 63.5 mm diameter metal shell and contains a built-in customized circuit board and battery for driving temperature measurements and wireless data transfer ...


Application Of Machine Learning To Chf Modelling, Mingfu He Mr Jan 2019

Application Of Machine Learning To Chf Modelling, Mingfu He Mr

Nuclear Engineering ETDs

Accurate prediction of CHF is still a challenging issue in the study of boiling heat transfer. Many factors contribute to the occurrence of CHF and the various trigger mechanisms are proposed to unravel physical phenomena behind CHF. However, those mechanisms cannot cover the multiple primary factors simultaneously and even some of them still remain controversially unresolved. In light of the complexity and difficulty of CHF modelling, hereby an ensemble-learning based framework is proposed to model and predict CHF based on the databank of CHF. Some prior trials have been done for three primary aspects of dominant factors, that is, surface ...


An Experimental And Numerical Investigation Of Flow Accelerated Flibe Corrosion, David B. Weitzel Jan 2019

An Experimental And Numerical Investigation Of Flow Accelerated Flibe Corrosion, David B. Weitzel

Nuclear Engineering ETDs

Renewed interest in molten salt reactor technology has brought to light the need for a better understanding of FLiBe corrosion. To this end a flowing FLiBe corrosion test loop was designed to test the flow effects of FLiBe corrosion. The loop consists of a pump, melt tank, and stainless-steel tubing assembly that heats the molten salt to high temperatures and circulates it over test specimens. The experiment has been constructed and has completed initial shakedown testing.

To support the flowing FLiBe experiment, a numerical corrosion model that couples FLiBe electrochemistry, solid metal diffusion, and mass transport was implemented. The model ...


Measurement And Analysis Of The Extreme Physical Shock Environment Experienced By Crane-Mounted Radiation Detection Systems, Matthew Boyd, Jennifer Erchinger, Craig M. Marianno, Gene Kallenbach Jan 2019

Measurement And Analysis Of The Extreme Physical Shock Environment Experienced By Crane-Mounted Radiation Detection Systems, Matthew Boyd, Jennifer Erchinger, Craig M. Marianno, Gene Kallenbach

International Journal of Nuclear Security

At ports of entry, radiation detectors could be mounted on container gantry crane spreaders to monitor cargo containers entering and leaving the country. These detectors would have to withstand the extreme physical conditions experienced by these spreaders during normal operations. Physical shock data from the gable ends of a spreader were recorded during the loading and unloading of a cargo ship by two hard mounted PCB Piezotronics model 340A50 accelerometers and two Lansmont SAVER 9X30 units (with padding). The majority of large shocks were observed in the vertical direction. The Lansmont units recorded mean shocks of 22.215 ± 1.174 ...


Compositional Analysis Of Cerium And Cesium In Rapid Setting Cement As An Immobilization Agent For Nuclear Waste, Riyadh M. Motny Jan 2019

Compositional Analysis Of Cerium And Cesium In Rapid Setting Cement As An Immobilization Agent For Nuclear Waste, Riyadh M. Motny

Theses and Dissertations

A feasibility of rapid setting cement (RSC) as an agent of immobilization for certain elements such as fission products or radioactive materials was explored. Cerium (Ce) and cesium (Cs) have been selected as a surrogate for U and/or Pu and fission products, respectively, in this study in three phases. In Phase I, RSC was evaluated for physical properties (e.g., porosity, density, pH values, etc.) using two groups methods—the cement powder at different concentrations of Ce (2 – 10 wt%) with deionized water (DIW) and artificial seawater (ASW). The results showed that the final setting time and compressive strength ...


Experimental Investigation Of Liquid Contact In The Developing Post-Dryout Chf Flow Boiling Regime Using Surface Mounted Thermistors, Hiralkumar Harshadbhai Patel Jan 2019

Experimental Investigation Of Liquid Contact In The Developing Post-Dryout Chf Flow Boiling Regime Using Surface Mounted Thermistors, Hiralkumar Harshadbhai Patel

Doctoral Dissertations

"Understanding heat transfer in the post-critical heat flux (CHF) flow boiling regime is important for determining the performance of the heat transfer equipment for various industrial applications requiring high heat transfer rates, e.g., heat exchangers, boilers, chemical and nuclear reactors. Liquid can be present in the core of the flow channel in the form of entrained liquid droplets, especially immediately downstream of film dryout. These droplets are suspected to provide an important heat transfer mechanism as they impinge on the heated wall. The objective of the current study is to investigate liquid contact with the heated wall in this ...


Holt (Maria) Papers, 1962-2016, Special Collections, Raymond H. Fogler Library, University Of Maine Jan 2019

Holt (Maria) Papers, 1962-2016, Special Collections, Raymond H. Fogler Library, University Of Maine

Finding Aids

Born and raised in Farmington, Maria Glen Holt studied nursing at Cornell University in New York. She worked many years as a public health nurse. Maria married Dr. Alfred Holt and the couple moved to Bath when Dr. Holt transferred his practice there. Maria served two terms as a State Representative and ultimately became an environmental activist, fighting against construction of Maine Yankee Nuclear Power Plant in Wiscasset, Maine. In 2017, Maria co-authored the book, The Death of Maine Yankee: Antinuclear Activists' Adventures, 1969-1996, with Elisabeth King.

Records include correspondence, publications, flyers, notes, and other materials documenting the efforts to ...


Call For Abstracts - Resrb 2019, July 8-9, Wrocław, Poland, Wojciech M. Budzianowski Dec 2018

Call For Abstracts - Resrb 2019, July 8-9, Wrocław, Poland, Wojciech M. Budzianowski

Wojciech Budzianowski

No abstract provided.


Xenon Dynamics Of Ahwr, Arindam Chakraborty, Baltej Singh Dec 2018

Xenon Dynamics Of Ahwr, Arindam Chakraborty, Baltej Singh

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Large core reactors where the core dimension is significantly large compared to the migration length of neutron are more susceptible to xenon instability due to local perturbations. Advanced Heavy Water Reactor (AHWR) is being designed for on-power refueling. Therefore, refueling or movement of control devices in AHWR causes local perturbation. Preliminary modal analysis of AHWR equilibrium core also showed that the eigenvalue separation between fundamental mode and 1st azimuthal mode is small indicating its susceptibility to xenon oscillation in azimuthal plane. Therefore, xenon dynamic studies for AHWR with explicit xenon calculations were carried out using diffusion theory based computer code ...


Transient Analysis Of Primary Feed Pump Trip For 700 Mwe Iphwr, S. Phani Krishna, S. Pahari, S. Hajela, M. Singhal Dec 2018

Transient Analysis Of Primary Feed Pump Trip For 700 Mwe Iphwr, S. Phani Krishna, S. Pahari, S. Hajela, M. Singhal

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is a horizontal channel type reactor with two loops of Primary Heat Transport (PHTS) system. Three (two operating and one stand by) main boiler feed water pumps (BFP) supply feed water to Steam Generators (SGs). In the event of one of the running BFP trip, standby comes on line on auto. Transient analysis for this event is performed using in- house computer code ATMIKA.T .The transient has been initiated by tripping one of the pumps.

Two cases are postulated:

1: BFP Trip and Standby BFP available on auto
2: BFP Trip ...


Multi-Grid Acceleration Scheme For Neutron Transport Calculations Using Optimally Diffusive Cmfd Method, Lakshay Jain, Ramamoorthy Karthikeyan, Umasankari Kannan Dec 2018

Multi-Grid Acceleration Scheme For Neutron Transport Calculations Using Optimally Diffusive Cmfd Method, Lakshay Jain, Ramamoorthy Karthikeyan, Umasankari Kannan

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Method of characteristics (MOC) is one of the most efficient deterministic techniques for high fidelity neutronic analysis of complex and heterogeneous reactor problems. However, the conventional MOC inner-outer iteration scheme suffers from poor convergence speeds for problems with large scattering to transport cross-section ratio and/or large dominance ratio. This creates a serious hindrance for its effective application to realistic reactor problems. A High Order – Low Order (HO-LO) multi-grid scheme using optimally diffusive coarse mesh finite difference (odCMFD) method has been introduced for improving the performance of code DIAMOND, an assembly level neutronic analysis code based on MOC and unstructured ...


Review Of Fuel Management Practices At Various Stages Of Nuclear Fuel Cycle In Phwrs In View Of Environmental Effects, Ravi Kumar Bansal, H. S. Sharma Dr, R. K. Singh Dr, P. N. Prasad Dec 2018

Review Of Fuel Management Practices At Various Stages Of Nuclear Fuel Cycle In Phwrs In View Of Environmental Effects, Ravi Kumar Bansal, H. S. Sharma Dr, R. K. Singh Dr, P. N. Prasad

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Nuclear Power is emerging as a promising source of environmentally benign energy source alternate from both pollution free environment as well as solution to global warming because of minimal carbon footprint. However, release of radiation and radioactive contamination during fuel cycle operations comprising the optimum fuel utilization in Nuclear Reactors, still remains a challenge to contain the sources of radiation and contamination away from public domain. This review article envisages qualitatively the environmental effects w.r.t. radiation during flow of Natural Uranium fuel used in Indian Pressurized Heavy Water Reactors (IPHWRs) at various stages of mining, fabrication, transportation, operation ...


Heavy Water Concentration Measurement In Air, A. Gupta, D. V. Uduapa, A. Topkar, A. K. Mohanty Dec 2018

Heavy Water Concentration Measurement In Air, A. Gupta, D. V. Uduapa, A. Topkar, A. K. Mohanty

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

The heavy water in PHWRs flows at high temperature and pressure, hence leaks in the heat transport system are not uncommon. The loss of heavy water due to such leaks can lead to spreading of radioactivity and it also contributes to operating cost of the nuclear reactor. It is advantageous to detect small leaks, because if remains undetected, they may develop into a severe leak, which may lead to reactor shutdown. None of the sensors which are currently in use can meet all the requirement of high sensitivity, and real time measurement which is free from interference from other gamma ...


Flow And Thermal Effects Of Blockages In A Nano-Fluid Cooled Nuclear Fuel Subassembly, Shubham Mandot, N. Govindha Rasu Dec 2018

Flow And Thermal Effects Of Blockages In A Nano-Fluid Cooled Nuclear Fuel Subassembly, Shubham Mandot, N. Govindha Rasu

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Nanofluids have a great impact on heat transfer characteristics due to increased thermal conductivity and heat transfer coefficient. In this study, Titanium nanoparticles mixed in liquid sodium has been chosen for analyzing the effect of Nanofluid coolant for a Nuclear Sub- assembly. This study is conducted to observe the effect of nanoparticles on the flow properties and heat transfer characteristics such as velocity, heat transfer coefficient, clad temperature, coolant temperature etc. These effects have been observed for varying nanoparticle concentration and different flow blockage sizes. For this study, 7-pin fuel bundle with and without blockage has been modeled and analyzed ...


Transient Simulation Of Lbe Cooled Chtr Under Natural Circulation With 3d Multi-Physics Code Arch-Th, D. K. Dwivedi, Anurag Gupta, Umasankari Kannan Dec 2018

Transient Simulation Of Lbe Cooled Chtr Under Natural Circulation With 3d Multi-Physics Code Arch-Th, D. K. Dwivedi, Anurag Gupta, Umasankari Kannan

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

India is developing a 100kWth Compact High Temperature Reactor (CHTR) to facilitate demonstration of technologies for high temperature process heat applications. CHTR is being designed as thorium based TRISO fueled and beryllium oxide moderated prismatic block type vertical core cooled with lead-bismuth eutectic (LBE) under natural circulation for 1000°C outlet. The new concept of high temperature core requires multi-physics multi-scale modeling based tools for investigating the normal operational behavior as well as anticipated transients of CHTR. In view of that, 3D multi-physics code ARCH-TH is being indigenously developed and validated for coupled neutronics-thermal hydraulic benchmarks. The multi-group diffusion based ...


Cfd Simulation Of Hydrodynamics And Scrubbing Behaviour Of Iodine Vapors In A Self-Priming Venturi Scrubber, Paridhi Goel, A. K. Nayak Dec 2018

Cfd Simulation Of Hydrodynamics And Scrubbing Behaviour Of Iodine Vapors In A Self-Priming Venturi Scrubber, Paridhi Goel, A. K. Nayak

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

In a severe accident scenario, the inadequate heat removal in a nuclear reactor can lead to over pressurization of the containment thus challenging its integrity. If not controlled, this can lead to release of radionuclides and high pressure steam in the environment. To ensure that the containment building remains intact and the reactor depressurizes, the vent line from the reactor is directed to a scrubber tank consisting of multiple venturi scrubbers, metal fiber filter and demister pad (known as Filtered Containment Venting System (FCVS)). This is a passive safety measure suggested for installation in advanced and existing nuclear reactors post ...


Non-Optical Imaging Of Flow, Boiling, And Salt Deposition In A Simulated Debris Bed, Molly Ross, Alan Cebula, Steven Eckels, D. S. Mcgregor, Hitesh Bindra Dec 2018

Non-Optical Imaging Of Flow, Boiling, And Salt Deposition In A Simulated Debris Bed, Molly Ross, Alan Cebula, Steven Eckels, D. S. Mcgregor, Hitesh Bindra

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Determining flow and heat transfer characteristics in a debris bed or a packed bed is difficult due to the lack of optical access. Non-optical imaging methods, such as x-ray or neutron imaging, can be used to observe flow characteristics and particle deposition, as well as boiling in a packed bed. An amorphous Silicon detector based digital radiography camera can be used to image with either x-rays or neutrons at up to 100 frames per second. The digital radiography camera, coupled with digital image analysis techniques was used to characterize fluid fraction and flow rates in a simulated debris bed. A ...


Experimental Evaluation Of Critical Heat Flux In Downward-Facing Boiling On Flat Plate Relevant To In-Vessel Retention In Indian Phwrs, Sumit V. Prasad, A. K. Nayak Dec 2018

Experimental Evaluation Of Critical Heat Flux In Downward-Facing Boiling On Flat Plate Relevant To In-Vessel Retention In Indian Phwrs, Sumit V. Prasad, A. K. Nayak

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Retention of corium inside the CV and cool it by calandria vault water is essential to mitigate severe accidents in PHWRs. The thermal failure of CV can be prevented by effective decay heat removal on the outer surface of CV using vault water, which depends on the heat transfer behaviour from the outer surface of CV to the vault water. Determination of limiting heat removal capability of vault water through outer surface of calandria vessel is very important. Since, the calandria vessel has a very large diameter and length, the bottom most part of the calandria vessel almost behaves as ...


Enhancements To The Discrete Generalized Multigroup Method, R. L. Reed, J. A. Roberts Dec 2018

Enhancements To The Discrete Generalized Multigroup Method, R. L. Reed, J. A. Roberts

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

This work seeks to improve the practicality of the discrete generalized multigroup (DGM) method. The DGM method divides a fine-group energy domain into a set of coarse groups. Fine-group fluxes within each coarse group are expanded in an orthogonal basis, and cross section moments are defined to preserve the reaction rates of the fine-group solution. Previous implementations of DGM suffered from large memory requirements, so this work work explores options to reduce the memory footprint by (a) homogenizing cross-section moments over coarse regions and (b) representing discrete-angle dependence through truncated Legendre expansions. Tests were performed using a 1-D, discrete ordinates ...


A Review Of Core Catchers For Advanced Reactors, Samyak Munot, A. K. Nayak Dec 2018

A Review Of Core Catchers For Advanced Reactors, Samyak Munot, A. K. Nayak

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

In order to address the challenges of severe accident and to ensure safety of people and environment, a core retention device called as “core catcher” has been incorporated in the present and future reactor designs. The concept of core catcher came into existence as early as in early nineties. It is the system which is placed inside the reactor in such a manner that even in the severe accidental scenario, it will retain the corium, quench it and then sustain the coolability of the debris formed due to corium water interactions. From then various approaches to development of core catchers ...


Critical Heat Flux And Power Transients At Low-Pressure Low-Flow Conditions In Vertical Flow Boiling, S. R.G. Vadlamudi, A. K. Nayak Dec 2018

Critical Heat Flux And Power Transients At Low-Pressure Low-Flow Conditions In Vertical Flow Boiling, S. R.G. Vadlamudi, A. K. Nayak

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

In the advanced boiling water reactor concepts such as AHWR and ESBWR, the recirculation pumps are eliminated in order to simplify the design. During the start-up of such reactors, to establish the natural circulation, the primary coolant has to be heated up slowly in steady steps; it is quite important to determine the characteristics of boiling and critical heat flux (CHF) values in order to establish operational limits. Flow boiling experiments were conducted in an annular channel at very low flow rates and atmospheric pressure varying the inlet subcooling from 20 to 400 C. CHF occurred during the oscillatory flow ...


Progress In Micro-Layered Fast Neutron Detectors, Priyarshini Ghosh, W. Fu, Mark J. Harrison, Patrick K. Doyle, Jeremy A. Roberts, D. S. Mcgregor Dec 2018

Progress In Micro-Layered Fast Neutron Detectors, Priyarshini Ghosh, W. Fu, Mark J. Harrison, Patrick K. Doyle, Jeremy A. Roberts, D. S. Mcgregor

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

The study of accident tolerant fuels is ongoing at the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory. The TREAT Facility provides quick, high-energy neutron pulses that simulate various accident conditions. These neutron pulses are presently detected using an array of fast-neutron detectors called Hornyak buttons. Hornyak buttons suffer from poor detection efficiency and significant Ĉerenkov radiation contamination in the signal. A new enabling technology, the micro-layered fast-neutron detector (MLFD), is presented to monitor neutron flux changes during mild-to-severe reactor accidents. The MLFD was designed to overcome the shortcomings of the Hornyak buttons and to improve detection efficiency. The ...


Matryoshka Inspired Statistical Surrogate For Turbulent Mixing, Abhinav Gairola, Hitesh Bindra Dec 2018

Matryoshka Inspired Statistical Surrogate For Turbulent Mixing, Abhinav Gairola, Hitesh Bindra

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Thermalhydraulics of reactor plena is often the most challenging problem due to turbulent mixing, role of jets and stratified flows. Due to which, the safety analysis of the overall reactor vessel or reactor system become complex. Inability of the system level analysis codes to accurately model the role of tur- bulent mixing in scalar transport leads to a major source of uncertainties. The computational cost of accurate CFD models renders the sensitivity studies using those for reactor safety ineffective. Reduced order models are needed which can effectively capture the effects of turbulent mixing and can be used for parametric studies ...


Thermal Stratification In Liquid Metal Pools Under Cold Transients, Brendan Ward, Hitesh Bindra Dec 2018

Thermal Stratification In Liquid Metal Pools Under Cold Transients, Brendan Ward, Hitesh Bindra

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Experimental results are presented for understanding the thermal stratification or mixing in a low Prandtl number, Pr, pool due to the injection of a colder (higher density) jet at the bottom of the pool. In liquid metals (Pr<<1), the higher volumetric thermal expansion enhances buoyant forces, aiding in thermal stratification while the low Pr extends the thermal boundary layer. Rayleigh backscattering with swept wavelength interferometry is used to generate the high fidelity distributed temperature data. The high spatial and temporal resolution of the sensors are required to capture the temperature gradient and fluctuations of temperature allowing more complete understanding of ...