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Full-Text Articles in Nuclear Engineering

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 ...


Numerical Evaluation Of Micro-Pocket Fission Detectors, Wenkai Fu, Daniel M. Nichols, Douglas S. Mcgregor, Jeremy A. Roberts Dec 2018

Numerical Evaluation Of Micro-Pocket Fission Detectors, Wenkai Fu, Daniel M. Nichols, Douglas S. Mcgregor, Jeremy A. Roberts

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Micro-pocket fission detectors (MPFDs) are miniature fission chambers suitable for in-core neutron measurement that have been under development at Kansas State University for over one decade. Current-generation devices have been used at a number of university reactors (Kansas State, Wisconsin, and MIT) and as part of the first experiments performed during the recent restart of TREAT. Ongoing research aims to improve understanding of the existing MPFDs and to optimize designs for future deployment. To aid in this development, the dynamic response of a prototypic MPFD was evaluated using Garfield++, Elmer, Gmsh, and Stopping and Range of Ions in Matter (SRIM ...


Preliminary Assessment Of Steady-State And Transient Reaction-Rate Measurements At The University Of Wisconsin Nuclear Reactor, J. A. Roberts, T. R. Ochs, D. M. Nichols, W. Fu, Y. Cheng, J. C. Boyington, D. S. Mcgregor, P. P.H. Wilson, R. J. Agasie, C. S. Edwards, Y-H. Park Dec 2018

Preliminary Assessment Of Steady-State And Transient Reaction-Rate Measurements At The University Of Wisconsin Nuclear Reactor, J. A. Roberts, T. R. Ochs, D. M. Nichols, W. Fu, Y. Cheng, J. C. Boyington, D. S. Mcgregor, P. P.H. Wilson, R. J. Agasie, C. S. Edwards, Y-H. Park

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Recently, a number of reactor-physics experiments were conducted at the University of Wisconsin Nuclear Reactor (UWNR) using a set of 7 micro-pocket fission detector (MPFD) probes and 3 resistance temperature detector (RTD) probes. The UWNR core is a TRIGA-fueled, MTR conversion using 2x2 fuel bundles separated by coolant channels. Each MPFD probe contained 4 detectors, and each RTD probe contained 6 detectors, all arranged uniformly along the active fuel height. These probes were placed in four different configurations to measure fluxes and temperatures in every accessible coolant channel for a variety of steady-state and transient operations. Relative fluxes can be ...


Micro Structured Sensors For Neutron Detection, D. S. Mcgregor, S. L. Bellinger, J. C. Boyington, Y. Cheng, R. G. Fronk, W. Fu, L. C. Henson, J. D. Hewitt, C. W. Hilger, R. M. Hutchins, K. E. Kellogg, J. A. Medina, D. M. Nichols, T. R. Ochs, M. A. Reichenberger, J. A. Roberts, S. R. Stevenson, T. M. Swope, T. C. Unruh Dec 2018

Micro Structured Sensors For Neutron Detection, D. S. Mcgregor, S. L. Bellinger, J. C. Boyington, Y. Cheng, R. G. Fronk, W. Fu, L. C. Henson, J. D. Hewitt, C. W. Hilger, R. M. Hutchins, K. E. Kellogg, J. A. Medina, D. M. Nichols, T. R. Ochs, M. A. Reichenberger, J. A. Roberts, S. R. Stevenson, T. M. Swope, T. C. Unruh

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

The shortage of 3He gas, identified as a problem several years ago, initiated research into alternative neutron detectors for various applications. One such technology is the microstructured semiconductor neutron detector (MSND). These compact detectors have microstructures etched deeply into the substrates that are subsequently backfilled with neutron reactive material. Single sided devices typically have thermal neutron detection efficiencies exceeding 30%, while double sided microstructured semiconductor neutron detectors (DS-MSND) have yielded >69% thermal neutron detection efficiency. Both MSNDs and DS-MSNDs have been integrated into compact low-noise and low-power electronics modules. Dosimetry calculations indicate that these detectors can be used as ...


Stratification And Mixing In The Hot Plena Of Liquid Metal-Cooled Reactors, Hitesh Bindra, Brendan Ward, Graham Wilson, Abhinav Gairola Dec 2018

Stratification And Mixing In The Hot Plena Of Liquid Metal-Cooled Reactors, Hitesh Bindra, Brendan Ward, Graham Wilson, Abhinav Gairola

Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Understanding or modeling the role of stratification and mixing in the plena or containments of nuclear reactors is of prime significance to their safety analysis. Particularly, in case of liquid metal-cooled reactors, thermal stratification in the hot pools under off-normal transients is one of the least understood problems that have multi-physics effects on thermo- mechanics and reactor physics. This is primarily due to lack of high fidelity experimental data for validating CFD or system scale models, which are essential for improved understanding. A scaled liquid metal thermal-hydraulic facility with a scaled hot plenum has been developed at Kansas State University ...