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Electrochemical Separation Of Multivalent Species On A Liquid Bismuth Cathode In Licl-Kcl Eutectic For Used Nuclear Fuel Reprocessing, Michael Woods 2019 Virginia Commonwealth University

Electrochemical Separation Of Multivalent Species On A Liquid Bismuth Cathode In Licl-Kcl Eutectic For Used Nuclear Fuel Reprocessing, Michael Woods

Theses and Dissertations

The presence of group I/II fission products (Cs-137, Sr-90, and Ba-137) within molten salt nuclear processes degrades operational efficiencies by contributing to increased radiation levels in the case of material handling processes or to loss of criticality in the case of a reactor. While methods such as zone freezing and ion exchange have been proven for the separation of these fission products in LiCl-KCl salts, they require extra equipment and processing steps. Addition of a liquid metal electrode to molten salt media, such as the electrorefiner of a pyroprocessing scheme or the salt cleaning stage of a molten salt ...


Experimental Investigation Of Liquid Contact In The Developing Post-Dryout Chf Flow Boiling Regime Using Surface Mounted Thermistors, Hiralkumar Harshadbhai Patel 2019 Missouri University of Science and Technology

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 ...


Microstructural Evolution Of Zirconium Carbide (Zrcₓ) Ceramics Under Irradiation Conditions, Raul Florez 2019 Missouri University of Science and Technology

Microstructural Evolution Of Zirconium Carbide (Zrcₓ) Ceramics Under Irradiation Conditions, Raul Florez

Doctoral Dissertations

A comprehensive understanding of the microstructural evolution of Zirconium Carbide (ZrC2) ceramics under irradiation conditions is required for their successful implementation in advanced Gen-IV gas-cooled nuclear reactors. The research presented in this dissertation focusses on elucidating the ion and electron irradiation response of ZrC2 ceramics. In the first part of the research, the microstructural evolution was characterized for ZrC2 ceramics irradiated with 10 MeV Au3+ ions up to doses of 30 displacement per atom (dpa) at 800 ºC. Coarsening of the defective microstructure, as a function of dose, was revealed by transmission electron microscopy analysis. The ...


Reactor Configurations To Support Advanced Material Research, Thaqal Mazyad Alhuzaymi 2019 Missouri University of Science and Technology

Reactor Configurations To Support Advanced Material Research, Thaqal Mazyad Alhuzaymi

Doctoral Dissertations

"The research goal is to configure a research reactor with multi-spectral capability for advance material research. This required the consideration of high and low-power levels alongside advanced fuel with high physical density and low enriched 235U. Selected fuel was U-10Mo with 19.75% 235U enrichment. The fuel and control rods system geometries were adopted from Missouri S&T Reactor (MSTR).

A high-power configuration (HPC) at 2 megawatts and low-power configuration (LPC) at 200 kilowatts were considered. Neutronic performances of the configurations were modeled using Monte Carlo N-particle (MCNP) transport code, version 6. Thermal-hydraulic analysis was performed with ANSYS ...


Development And Characterization Of Nanostructured Steels And High Entropy Alloys For Nuclear Applications, Andrew Kalevi Hoffman 2019 Missouri University of Science and Technology

Development And Characterization Of Nanostructured Steels And High Entropy Alloys For Nuclear Applications, Andrew Kalevi Hoffman

Doctoral Dissertations

"Nuclear reactor materials are subjected to a harsh environment including high temperatures and radiation fluences. In order to extend the lifetime of current light water reactors (LWRs) and realize the development of advanced Gen IV nuclear reactors new materials must be developed which can withstand such an environment. This thesis involves two approaches to solving this materials problem: advanced manufacturing of current commercial alloys using severe plastic deformation (SPD) and the development of new advanced high entropy alloys (HEAs).

Because SPD is effective at achieving grain refinement, this technique was used to obtain material having a high volume fraction of ...


Au@Tio2 Nanocomposites Synthesized By X-Ray Radiolysis As Potential Radiosensitizers, Maria C. Molina Higgins 2019 PhD Student

Au@Tio2 Nanocomposites Synthesized By X-Ray Radiolysis As Potential Radiosensitizers, Maria C. Molina Higgins

Theses and Dissertations

Radiosensitization is a novel targeted therapy strategy where chemical compounds are being explored to enhance the sensitivity of the tissue to the effects of ionizing radiation. Among the different radiosensitizers alternatives, nanomaterials have shown promising results by enhancing tumor injury through the production of free radicals and reactive oxygen species (ROS). In this work, Gold-supported titania (Au@TiO2) nanocomposites were synthesized through an innovative strategy using X-ray irradiation, and their potential as radiosensitizers was investigated. Radiosensitization of Au@TiO2 nanocomposites was assessed by monitoring the decomposition of Methylene Blue (MB) under X-ray irradiation in the presence of the ...


Compositional Analysis Of Cerium And Cesium In Rapid Setting Cement As An Immobilization Agent For Nuclear Waste, RIYADH M. MOTNY 2019 Virginia Commonwealth University

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 ...


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

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 2018 Wojciech Budzianowski Consulting Services

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

Wojciech Budzianowski

No abstract provided.


Heavy Water Concentration Measurement In Air, A. Gupta, D. V. Uduapa, A. Topkar, A. K. Mohanty 2018 Bhaba Atomic Research Center

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 2018 Vellore Institute of Technology, Vellore

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 Analysis Of Primary Feed Pump Trip For 700 Mwe Iphwr, S. Phani Krishna, S. Pahari, S. Hajela, M. Singhal 2018 Nuclear Power Corporation of India

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 ...


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

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 ...


Multi-Grid Acceleration Scheme For Neutron Transport Calculations Using Optimally Diffusive Cmfd Method, Lakshay Jain, Ramamoorthy Karthikeyan, Umasankari Kannan 2018 Homi Bhabha National Institute / Bhabha Atomic Research Centre

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 ...


Xenon Dynamics Of Ahwr, Arindam Chakraborty, Baltej Singh 2018 Bhabha Atomic Research Centre

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 ...


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 2018 Nuclear Power Corporation of India

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 ...


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

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 ...


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

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 ...


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 2018 Homi Bhabha National Institute

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 2018 Kansas State University

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 ...


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