Advisor: Dr. Inam ur Rehman, PIEAS (former Centre for Nuclear Studies, CNS), P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza, PIEAS (former CNS), P.O. Nilore, Islamabad
Co-advisor: Dr. Sikander M. Mirza, PIEAS (former CNS), P.O. Nilore, Islamabad
Thesis Title: Design of Training & Research Reactor (100 - 500 KW range)
Date of Completion: April, 1991, CNS REPORT No. CNS-NE 91/163
This project aimed at studying the mechanical design of a typical training and research reactor in the range of 100 - 500 KW power. The design studies were done using 20% enriched fuel. The reactor is supposed to be installed in the existing pool of PARR-II. Three possible approaches were studied for different types of fuel. It included standard plate type fuel, an irradiated fuel of PARR-I and pellet type fuel of KANUPP dimensions. For the plate type fuel, various thermal hydraulic parameters were calculated using NATCON computer code. The maximum power with natural convection coolling is about 623.8 kW for the given size of the core. Utilization of irridiated fuel poses problems of activity and corrosion. For the pellet type fuel with KANUPP dimensions having 20% enrichment, the study shows thermal hydraulic problems. Reduction in enrichment upto 3% in this case requires forced cooling systems. Drawings for the Ball-nut drive and rack & pinion derives for the control rods along with their working have been included in the report.
Advisor: Dr. Inam ur Rehman Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza, Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Sikander M. Mirza, Centre for Nuclear Studies, P.O. Nilore, Islamabad
In this work design calculations for a Training & Research Reactor in the power range of 100 - 500 KW were done using constraints of fuel dimensions, fuel type and available space at Nilore. The neutronic studies were performed using LEOPARD, WIMSD-IV, EXTERMINATOR and CITATION computer codes. The pitch was selected after parameteric studies using zero dimensional LEOPARD code. The effect of temperature was also studied using the Cell of the reactor core. Then group constants were generated using LEOPARD and WIMSD-IV for various designs. Using these group constants the two dimensional analysis was performed using state of the art EXTERMINATOR code. The size of the core with a graphite reflector was determined using this analysis. Neutron and gamma ray shielding studies for the design were also done in this work to ensure the safety of the design.
Advisor : Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
A computer program INGRE (Indoor Gamma-Ray Exposure) has been developed in this work to calculate the exposure rate due to gamma rays emitted from the construction materials of a room, at any point in the room for variety of energies of gamma rays. The code takes the voids ( i.e. doors and windows etc) in the room walls under consideration. An option has been made available in the code to select Berger or Taylor's form of Buildup factor in the exposure calculations. The code can optimize the mesh size for the permissible error and CPU time to perform volume integration. The study has been done for three different materials and results have been compared with analytical results for simple cases. The computational results are in good agreements with analytical results for simple geometry.
Advisor: Dr. Sikander M. Mirza Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Date of Completion: May, 1992, CNS REPORT No. CNS-NE 92-
A computer program has been developed in this work to calculate the fluxes due to neutrons and gamma rays inside and outside the concrete shield of a typical power reactor. Using the program, the flux and doses due to gamma rays and neutrons at various positions around the reactor have been computed. The gamma rays were distributed into six energy groups from sero to 10 MeV range. This work shows that doses outside a typical concrete shield of a PWR have contributions due to gamma rays only. The dose rate decreases as the shield thickness is increased. The results for variety of shield designs are discussed in the thesis.
Advisor: Dr. Nasir M. Mirza Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Visualization of Avalanches in Proportional Gas Detector Using Monte Carlo Method
Date of Completion: September, 1995, CNS REPORT No. CNS-NE 95 - 282
Stochastic modeling of the electron avalanches in a proportional counter has been done by developing a three-dimensional Monte Carlo simulation program MCPC. The effect of variations in applied voltage, gas pressure, and detector dimensions, on the amplification factor of the detector have been studied. A good agreement between the values of various integral parameters as predicted by Monte Carlo simulations and analytical expressions have been found and deviations remain less than 7.4%.
Advisor: Dr. Nasir M. Mirza, Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Sikander M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Nuclear Instruments Simulator Development on Personal Computers
ABSTRACT
In this work a computer program Nuclear Instruments Module lABoratory (NIMLAB) has been developed to visualize and simulate all the useful functions of a typical NaI(Tl) detector based gamma ray spectroscopy system. The program has been written in C-language and has been made functional for IBM compatible personal computers. The NIMLAB code performs all primary functions of a power supply, preamplifier, spectroscopy amplifier, single channel analyzer, counter timer, multi-channel analzer and oscilloscope. In this work parametric studies for gamma ray spectroscopy has been done using NIMLAB code and results were compared with experimental values. Comparisons show good agreements. The NIMLAB code development and its application are discussed in detail in this thesis.
Advisor: Dr. Mushtaq Ahmad Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Development of Simple Computational Procedure for Reactor Analysis
The linear reactivity model based simple codes BRIC, BRACC and RPM were converted from BASIC to FORTRAN language on VAX/VMS and PC. Some additional features were included in the programs. The programs in FORTRAN were then applied to test problems given in literature. The results of BASIC and FORTRAN were compared and found almost same. The input requirements of the program RPM for CHASNUPP were obtained by using LEOPARD and WIMS/D4 codes. The RPM program was then applied on CHASNUPP core to find the burnup and power distribution at different burnup steps. The burnup distribution and normalized power distribution obtained by using RPM code were compared with the reported data. The results were found in good agreement with the reported results.
Advisor: Dr. Nasir M. Mirza Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Sikander M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. M. Afzal Bhatti, Prof., Computer Science, Quaid-e-AzamUniversity.
Date of Completion: March, 1996
ABSTRACT
A dual trace medium frequency oscilloscope has been developed in this work with an intent to visualize and simulate all the basic functions of the actual physical instrument. The actual instrument responses have been simulated using three basic types of pulses viz. The sinusoidal, linear and the logic pulses. The pulse generator has also been simulated with options, square buttons and various icons. These are employed in the emulation of the medium frequency dual trace oscilloscope. The effect of some basic phenomenon such as intensity, focus and horizontal and vertical scaling have successfully accomplished. The effects and results are close to the results produced by actual instrument.
Advisor: Dr. Nasir M. Mirza Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Study of Fuel reloading Strategies in a Typical Pressurized Water Reactor
Date of Completion: September, 1996, CNS REPORT No.
ABSTRACT
This work aims at analysis of different reload configurations and fuel management strategies employed in a typical PWR. The study includes effect of fuel management parameters on steady state and transition cycles. These parameters include fuel enrichment, fuel to moderator ratio and discharge burnup. Linear reactivity based codes BRICC, BRACC and RPM were employed in this work. Fuel management strategies for equal power sharing and unequal power sharing systems were analyzed. Results are discussed in this thesis for out-in, in-out and low-leakage fuel reloading options.
Advisor: Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Determination of Corrosion Product Activity in Pressurized water reactors Under Power & Flow Transients
Date of Completion: September, 1996, CNS REPORT No. CNS - NE -
Simulation of coolant activation due to corrosion products and impurities in a typical Pressurized Water reactor (PWR) has been carried out using MATLAB-SIMULINK program. Calculations have been done for both steady state and transient conditions. The problem of determination of corrosion product activities under steady state condition is formulated into coupled differential equations which are then patched using SIMULINK. In first part of thesis, calculations for steady state conditions were made. The results for 56Mn, 24Na, 60Co, and 59Fe show that the equilibrium state of their specific activities in primary loop water under full power condition, is approached fairly rapidly. Then in second part, specific activity of corrosion products in water of the coolant system is determined under flow and power transients. Under flow transient conditions, effects of ramp flow transient, linear decrease in flow rate and primary pump coast down conditions are simulated in this work. For power transients, conditions of ramp transients, quasi-static condition, and power peak transients were simulated and effect on specific activity were calculated.
Advisor: Dr. Nasir M. Mirza, Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Study of A Prototype Actinide Burner Reactor
Date of Completion: September, 1997, CNS REPORT No. CNS-NE-91/67
This work aimed at designing a core configuration for a prototype actinide burner reactor that burns the actinides and fission products efficiently. An initial design of a typical fast reactor was choosen and an optimum core configuration was obtained. A parametric study was performed to find an optimum core configuration to transmute long lived fission products and minor actinides in a fast reactor with MOX fuel. The transmutation rates of the reactor were computed using ORIGEN-2 computer code and were compared with values from a typical PWR. The comparison shows that the fast burner reactor is more efficient in transmuting actinide and fission products as compared with a typical PWR.
Advisor: Dr. Mushtaq Ahmed Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Nuclear Reactor Core Calculations
Date of Completion: September, 1997, CNS REPORT No. CNS-NE
ABSTRACT
This work aims at the development of a code to solve steady state multi-group neutron diffusion equation in one-dimension. A computer code has been developed to solve the steady state multi-group neutron diffusion equations in one-dimension for slab, cylindrical and spherical geometry. The code solves the diffusion equation discretized into one hundred mesh points in twenty different material regions. Power iteration method has been applied to solve the eigen-value problem having multiplication factor as the eigen value and group fluxes as the eigen functions. The LU-decomposition method has been applied to solve system of linear equations. The output of the code includes the converged value of k-effective and the corresponding spatial distribution of each group flux. The results of the code have been compared with the existing three group diffusion theory based code ODMUG. These results are in good agreement within the errors of two percent.
Advisor: Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Analysis of Reactivity Transients in a Typical Pool-Type Research Reactor Using PARET Code
Date of Completion: September, 1997,CNS REPORT No. CNS - NE 97 -
The objective of the work is to analyze transient behavior of High Enriched Uranium (HEU) and Low Enriched Uranium (LEU) cores of a typical pool Type research reactor using PARET Code. The transient problem was forced through specification of externally inserted reactivity as a function of time. Reactivity insertions are idealized by ramps and steps. Quasi-static transients, super delayed-critical transients and super prompt-critical transients are selected for the analysis. The effect of initial power on transient behavior has also been investigated. The low enriched uransium core is also analyzed for transients with out scram. First order perturbation theory was applied to estimate the reactivity increment by accidental insertion of a standard fuel element in to flux trap of both cores. The calculated values of low enriched and highly enriched uranium cores are $1.451 and $0.53 respectively. The magnitude of maximum reactivity insertions are chosen to be in the range of $0.05 to $2.0 for various insertion times. Transient analysis with scram reveals that the response of both LEU and HEU cores is similar for selected ramps and steps. The difference is observed in the peak values of power and coolant, clad and fuel temperatures. Trip level is achieved earlier in the case of LEU core. The peak clad temperatures reached in both LEU and HEU cores remain below the melting point of aluminum for the selected reactivity insertions. Transient behavior of LEU core at very low power is found to be different from the transients at full power. For reactivity transients at low power, power rises sharply to a higher peak value whereas in transients at full power the peak power barely exceeds the trip level. The oscillations in power after main peak are observed for transients without scram. These oscillations are stabilized after certain time due to contributions of feed back coefficients.
Advisor: Dr. Nasir M. Mirza PIEAS, P.O. Nilore, Islamabad
Co-advisor: Dr. Sikander M. Mirza , PIEAS, P.O. Nilore, Islamabad
Thesis Title: Measurements of Effective Neutron Diffusion Parameters in Water with Voids
Date of Completion: September, 1997
CNS REPORT No. CNS-NE 97 -377
ABSTRACT
Preliminary experimental studies of diffusion parameters for Am-be neutrons have been done in water for different void fractions. A 10 Ci Am-Be neutron source has been employed in PIEAS Neutron Transport Facility. The fast and thermal neutron spatial distributions were measured in water for two different void fractions. These void fractions were simulated by perspex tubes of different diameters arranged in regular assembly attached to the water tank. The simulated void fractions were 0.03 and 0.05 respectively. The neutron measurements were done with BF3 - proportional detector. Then employing one-dimensional one group diffusion model the Fermi-age, diffusion area and migration area were determined. Results show that the Fermi age changes from 38.8 cm2 to 48.4 cm2 when the void fraction is increased from zero to 0.05. For same void fraction range the diffusion area varies from 39.9 to 49.1 cm2. The effect of voids on relaxation length were also studied.
Advisor: Dr. Nasir M. Mirza PIEAS, P.O. Nilore, Islamabad
Co-advisor: Dr. Sikander M. Mirza , PIEAS, P.O. Nilore, Islamabad
Date of Completion: September, 1998, CNS REPORT No. CNS-NE
During past two years experiments at Joint European Torus and Japanese Tokamak designs have demonstrated the energy break-even and high Q-values. This has made the study of the dynamics of burning plasma important so as to develop appropriate control mechanisms. In this work a simulation model has been developed to study the global dynamics for DT-fueled TOKAMAKS. This model is based on set of non-linear differential equations describing the evolution of the plasma density and temperature. The change of plasma power with an external heating, its approach to ignition temperatures and final equilibrium conditions were studied. Several gain and loss mechanisms in the confinement were also explored.
Advisor: Dr. Nasir M. Mirza PIEAS, P.O. Nilore, Islamabad
Thesis Title: Reactivity Insertion Limits in a Typical Pool Type Research Reactor
Date of Completion: September, 1998, CNS REPORT No. CNS-NE
Reactivity Insertion limits imposed by clad melting temperatures and their sensitivity with respect to safety parameters have been investigated in a typical pool type research reactor. Simulations were conducted using modified PARET code for High Enriched Uranium (HEU) UAlx-Al fuel and Low enriched uranium cores having U3O8-Al and U3Si2-Al fuels respectively. Super-prompt critical transients of more than one dollar reactivity were investigated. Reactor was kept critical at low power. Scram disabled and enabled transients were studied for ramp reactivity insertions. The clad meltdown was observed at $2.0, $1.9 and $2.2 per 0.5 sec in above mentioned three cores. In LEU (U3Si2-Al) core the clad failure is predicted for step reactivity insertions of $2.1. The behavior of the core with scram enabled at 12 MW power has also been investigated. In this case the clad melts at reactivities higher than $6.5 per 0..5 sec. Oscillations in the power, fuel and clad temperatures have been observed. The secondary power peak within 0.1 sec of the primary peak causes a non-linear rise in temperatures within the fuel, which leads to the clad meltdown. The sensitivity analysis with respect to beff, Doppler Coefficient and void coefficient of reactivity were also done in this work. Part of the thesis has been published in International two papers.
Advisor: Dr. S.A. Bhatti, Chairman Computer Science Department, Quaid-e-Azam University, Isamabad
Co-advisor: Dr. Nasir M. Mirza, PIEAS, P.O. Nilore, Islamabad
Thesis Title: Modeling of Human Population Dynamics
Date of Completion: September, 1998, CNS REPORT No. CNS-NE
A simple matrix model for the dynamics of human population has been developed. The basis is a set of mathematical relationships as a mixture of coupled differential equations and algebraic equations. In this model human population, pollution, capital investment and natural resources are used as variables. The project aimed at first model development and then providing continuous simulation by solving coupled differential equations. Then a simulating methodology was developed to show the effect of various parameters on human population. Results are discussed in the thesis, JAVA environment has been employed as working environment for the project.
Advisor: Dr. Sikander M. Mirza Centre for Nuclear Studies, P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza , Centre for Nuclear Studies, P.O. Nilore, Islamabad
Thesis Title: Stochastic Optimization Technique for Fuel Reload Pattern
Date of Completion: September, 1999
CNS REPORT No. CNS-NE 99 -
Simulated annealing based stochastic optimization methodology has been developed for the fuel reload pattern optimization in PWRs. This methodology has been implemented as a computer code TPMSA which automatically generated acceptable load pattern starting from random or user specified configuration employing flat power profile as objective function. The fixed enrichment, fixed number of assemblies and limit on local power peaking factor have been used as constraint conditions. This methodology uses PSY-LEOPARD generated depletion dependent group constants. The TPMSA code uses MCRAC code as a subroutine to obtain assembly - wise normalized power distribution. Acceptable loading patterns have been generated using TPMSA computer program starting from both standard as well as random loading patterns. Slow and fast linear as well as exponential annealing schedules have been tested. Slow annealing have been found to yield acceptable loading pattern in smallest number of annealing cycles. The transient probability matrix based simulated annealing adaptive methodology has been found effective for slow annealing at low temperatures. Multiple shuffling per annealing cycle has been found to accelerate the convergence rate of the simulated annealing procedure.
Advisor : Dr. Sikander M. Mirza, PIEAS P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza , PIEAS, P.O. Nilore, Islamabad
Date of Completion: September, 1999, CNS REPORT No. CNS - NE -
ABSTRACT
Inclusion of thermalizing regions in compact Liquid Metal Fast Reactors have led to transitory spectra near interfaces. Such spectra can not be studied using space independent group constants. In this work WIMSD-IV code has been employed to generate space dependent group constants in a typical LMR. Then three-dimensional model of a typical LMR with a BeO reflector has been studied in detail. The effect of global parameters such as reflector size, leakage currents and flux profiles on local spectral index, have been investigated. Detailed results are reported in the thesis.
Advisor : Dr. Nasir M. Mirza, PIEAS, P.O. Nilore, Islamabad
Co-advisor : Mrs. Rubina Nasir, PIEAS, P.O. Nilore, Islamabad
Thesis Title: Modeling and Simulation of Loss of Flow Transients in a Typical Research Reactor
Date of Completion: September, 1999, CNS REPORT No. CNS-NE 99 -
When heat transport system in a typical research reactor fails to remove the heat from the reactor core the reactor power starts increasing the temperature and feedback starts playing their part in the reactor kinetics. Even if the reactor is shutdown, the decay heat from the fission products can cause problems. This project first aimed at modeling a loss of flow transients in a typical pool type research reactor. Then simulation were carried out by using PARET code for Highly Enriched Uranium and Low Enriched Uranium cores of a typical pool type research reactor. Three types of perturbations were studied. These included flow rate perturbations only, transients with reactivity perturbations only and transients with both flow rate and reactivity perturbations. Calculations were done by varying the speed of primary pump at different time constants. The selected transients were studied for HEU and LEU cores. Also effect of reactor scram conditions and kinetic parameters were simulated.
Advisor: Dr. Nasir M. Mirza , PIEAS, P.O. Nilore, Islamabad
Advisor: Dr. Sikander M. Mirza , PIEAS, P.O. Nilore, Islamabad
Thesis Title: Simulation of Impurity Concentration Effects on Power Multiplication in TOKAMAK
Date of Completion: September, 1999, CNS REPORT No. CNS - NE 99 -
Heavy ions such as tungsten emit intense line radiation even in hot plasma core, so only small fractions are tolerable in a fusion reactor where one requires energy loss to be small as compared to the alpha heating rate. Similarly the light impurities such as carbon and oxygen enhance Bremsstrahlung so relatively large concentrations may be allowed in critical reactors. This work aimed at studying the effect of these high-Z and low-Z impurities on power multiplication of a typical toroidal fusion reactor. Also various methods to control such impurities in TOKAMAKS were explored.
Supervisors: Dr. Nasir M. Mirza
Examiners: Dr. Asad Majeed and Dr. Sikander M. Mirza
Thesis Title: Energy Cycle Modeling for a Typical Toridial Fusion Reactor
Synposis: This project aims at modeling a typical energy cycle of a Toridial fusion reactor. The energy flow rate in the reactor and out of the reactor will be considered. The total energy flow mechanisms and fractions going into various components will be established. The Q-value, efficiencies, required stability criteria and burnup fractions as a function of reactor parameters will be studied. The pulsed DT and steady state DT Tokamaks will be used in the model and comparisons will be done. A FORTRAN - 77 based program will be developed to simulate the energy cycle.
Advisor : Dr. Nasir M. Mirza PIEAS, P.O. Nilore, Islamabad
Co-Advisor : Dr. Sikander M. Mirza PIEAS, P.O. Nilore, Islamabad
Thesis Title: Computer Modeling of Geiger Discharge in a Typical Cylindrical Gas Detector
This project is concerned with detailed three dimensional modeling of typical Geiger discharge, its initiation and cascades of avalanches in a typical Cylindrical Gas Detector. This work aims at identifying geometry configurations material parameters, operating voltages and driving / quenching mechanism for cascades of avalanches in a typical GM Tube. In this project, first the development of a realistic Monte Carlo model for the electron history in a gas, its multiplication, and initiation of avalanches will be studied. Then triggering of Geiger discharge and quenching process will be simulated. Finally study will be done to find integral parameters of the detector as a function of the detector operating conditions.
Advisor: Dr. Sikander M. Mirza PIEAS, P.O. Nilore, Islamabad
Co-advisor: Dr. Nasir M. Mirza PIEAS, P.O. Nilore, Islamabad
Thesis Title: Determination of Neutron Albedo for Complex Geometries Using Monte Carlo Method
The values of neutron albedo for flux, current and the absorbed dose are typically available for monolithic regions. Also, the measured and the calculated values of neutron albedo are generally restricted to mono-energetic neutron beams in the high energy range. The aim of this project is to carry-out detailed Monte Carlo calculations for the neutron albedo for complex geometric involving layers / regions with different materials.
Parametric studies were performed on the energy dependence of neutron albedo and on the dimensions of the regions involved. The behavior of the neutron albedo in the low energy range was also be studied. The results have been compared with the available published data.
Advisor: Dr. Nasir M. Mirza
Co-advisor: Dr. Sikander M. Mirza
Thesis Title: Development of a Point Kinetic Three Species Model for DT- Fusion in Tokamak Fusion Reactors
Synopsis: This project aims at extending a single species simple point kinetic model of a typical TOKAMAK Fusion Reactor to a three species model for DT-plasma. The plasma will be now characterized by electrons, ions and alpha particles as compared to an earlier model developed by Riaz Khan for ions only. Balance equations for electrons, ions and alpha particles will be developed for a typical torus. Employing theoretical and semi-experimental models for MHD instabilities and confinement times. The energy and population balance will be developed. These equations will be numerically solved to find electrons, ions and alpha particle densities and temperatures as a function of time. Finally break -even conditions will be studied.
Thesis Title: Development of a visual work bench for the study of nonlinear dynamics of systems
Non linear dynamical systems have been found to exhibit quite complex behavior. The theory of such systems has many common features independent of system details and is found to cross many disciplinary boundaries. This project aims at development of a visual workbench for the study of non-linear dynamics of systems. The program will be designed at carryout standard non-linear dynamics analysis task with graphical / text output. This program will be validated by comparing its results with published results.
