http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/285 Tue, 05 May 2026 08:25:44 GMT 2026-05-05T08:25:44Z http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/1060 STRUCTURAL INTEGRITY ANALYSIS OF VVER 1200 REACTOR VESSEL WALL DUE TO PRESSURIZED THERMAL SHOCK SALMAN KARIM, MOHAMMED SARIM Reactor Pressure Vessel (RPV) is the most crucial and irreplaceable part of any Nuclear Power Plant (NPP). So, maintaining the integrity and safety of RPV during any kind of transient or accidental conditions when it can be detrimental e.g., Pressurized Thermal Shock (PTS), need to address as highest priority. Analyzing the PTS event is of utmost importance to ensure the safety and integrity of the VVER-1200, a Gen-III+ reactor model of the Russian Federation, during hypothetical transient scenarios. To analyze the PTS event, Beaver Valley-105 case which is Main Steam Line Break (MSLB) type Loss of Coolant Accident (LOCA), is taken into consideration for this study. A computational fluid dynamics (CFD) analyses was done by ANSYS Fluent considering the injection of cold-water from emergency core cooling system (ECCS) to the hot primary coolant system through cold leg. The beltline region is considered for the modeling to analyze the temperature evolution in high pressure at both coolant and inner wall surface during the transient event and found 167◦C temperature coolant at 200 sec simulations after ECCS water mixing. Sudden increase of pressure in primary system which can lead the scenario to PTS at low temperature is analyzed by ANSYS Mechanical to evaluate the stress profile. The stress analysis for the transient event, which caused PTS due to an instantaneous elevation to a primary system pressure of 16.2 MPa, has been done. It is observed that maximum stress, e.g. Equivalent (Von mises) stress was found 663 MPa and was generated at the edge of nozzle area and the stress values exceed the ultimate yield strength at 16.2 MPa i pressure. Furthermore, a comprehensive postulated crack modelling has been done considering the temperature and pressure of the selected transient case. Three different semi-elliptic crack model are analyzed at selected transient event and for elevated primary pressure of 18, 20 and 22 MPa respectively. Finally, the results of stress intensity factor (SIF) at 100 ◦C obtained for these three postulated crack cases have found, were analyzed with respect to fracture toughness value, KIC of initial state and end of service life state to assess the service life. Here, fracture toughness, KIC is calculated as per ASME and IAEA TECDOC considering the neutron fluence and Nil Ductility temperature values. SIF results (87-118 Mpa.m^1/2) were found below the limiting fracture toughness values, KIC at case II for every pressure conditions. For case I & III, SIF values exceed the KIC values limit at 22 MPa and 20 & 22 MPa respectively. A further study was done to determine the maximum allowable critical temperature for crack for these three cases. STRUCTURAL INTEGRITY ANALYSIS OF VVER-1200 REACTOR VESSEL WALL DUE TO PRESSURIZED THERMAL SHOCK Fri, 01 Mar 2024 00:00:00 GMT http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/1060 2024-03-01T00:00:00Z http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/826 ANALYSIS OF NEUTRONICS SAFETY PARAMETERS AND CORE BURNUP LIFETIME OF BAEC TRIGA MARK-II RESEARCH REACTOR USING THE DETERMINISTIC TRIGAP AND TRIGLAV CODES HASAN, MD RAKIBUL BTRR has been operating science 1986 without any form of reloading or shuffling. On September 14, 1986, 50 low enriched uranium (LEU) fuel components were put into the original core, and it reached its first initial criticality. After that, 100 fuel elements were loaded and this configuration is known as the operational core. The goals of this study are (i) to analyze the neutronics core safety parameters of Initial as well as the functioning core of BTRR and (ii) to calculate individual fuel element burn-up as well as ring wise average burnup at different burn-up conditions and core lifetime of the present low enrichment uranium (LEU) core configuration. To analyze the initial criticality experiment of BTRR, its initial critical TRIGA physical model has been developed and hence the initial effective multiplication factor, core excess reactivity has been calculated using deterministic code TRIGLAV and TRIGAP accordingly. Burnup calculations are predicated on the concept that while calculating the neutron density distribution, nuclide concentrations can be taken for granted to be constant. They are built on the neutron transportation calculation and the burnup equations, which are two fundamental equations in reactor physics. Individual fuel burnup as well as ring wise burnup calculation has been done by TRIGLAV code and it has been compared with the MVP-Burn code. Burnup has been calculated up to 1400 MWd and the 1400 MWd data has been compared with the result obtained from TRIGAP code. To estimate the core life time, core excess reactivity has been considered and the calculated results are likened with the experimental obtained values from reactor operational data log book. The reactor may be operated safely for an additional 500 MWd days in accordance with the need of burnup and excess reactivity. This study will be helpful to formulate the most economic use of the fuel rod initially overloaded in the core. Additionally, the study provides insightful information on the behavior of the reactor and will guarantee improved reactor usage and operation in the future. Additionally, utilizing the same fuel components, this can provide insight into redesigning a better core configuration. The author wishes to convey his sincere appreciation to Dr. Md. Jahirul Haque Khan, Supervisor of the thesis work and Head & CSO, RPED, for his understanding direction, passionate support, and helpful critiques of this study work. The author would also like to thank his teacher Dr. Md. Abdus Sattar Mollah, Co-Supervisor of the thesis work for his advice and assistance in keeping the author updated regarding the recent instrumentation needed for the research. The author acknowledges the generosity of the Head of the department Mr. Col Molla Md Zubaer, for financial allocation that made this study possible. The author gratefully thanks Mr. Lt Col Faisal Kader, for his help in every administrative aspect. The INST's RPED laboratory personnel were a great help to the author during the course of the study project, and they have our gratitude. The author would like to conclude by thanking his friend for their encouragement and assistance throughout the research. Wed, 01 Mar 2023 00:00:00 GMT http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/826 2023-03-01T00:00:00Z http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/808 DESIGN AND DEVELOPMENT OF A ROBOTIC SYSTEM FOR RADIATION DETECTION AND MEASUREMENT SATTAR, SABIHA Generally, nuclear radiation installations require continual radiation surveys to confirm that there is no radioactive pollution or excess radiation dose levels that might lead to radiation safety concern. The present practice for regular radiation/contamination survey works is to utilize human operators, which is both cost inefficient as well as radiation risks. As with many repetitive, regular jobs, there are enough opportunities for the radiation safety issues to be improved using autonomous commercial or in-house design robotic systems. In this study, a prototype of a portable robotic system based on wireless communication protocol has been proposed for radiation detection in nuclear environment. The main motive of the robotic system is to help the human operators from getting unavoidable excess radiation levels. The robotic system consists of a ground 4 wheels carrier, GM counter based ionizing radiation detection box, Raspberry Pi- mini-computer unit, Pi camera unit &web platform based wireless controlling and recording system. In this system, the robot is regulated from a web based server in order to move towards a desired position to locate the radioactive source along with dose measurement. Radiation dose levels from some point radioactive sources (e.g. 137Cs, 60Co, 54Mn) placed at various locations in the nuclear radiation facility, have been monitored and compared with a commercial GM dose rate counter (Gamma Scout, w/ALERT model). The percentage of error has been found varying from 0 to 14% based on the experimental data. It is concluded that in-house design portable robotic systems are worthy for ionizing radiation detection and radioactive lost source identification in lieu of occupational radiation workers in the nuclear installations. Tue, 01 Feb 2022 00:00:00 GMT http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/808 2022-02-01T00:00:00Z http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/798 MEASUREMENT OF BACKSCATTERING FACTOR FOR X RAY CALIBRATION BEAM WITH VARIOUS PHANTOMS USING IONIZATION CHAMBER AND THERMOLUMINESCENT DOSIMETER NUPUR, ROKSHANA PARVIN Radiation hazard is one of the critical issues at different medical centers, industries and nuclear facilities around the world. To protect radiation workers, public and environment, a reliable dose measurement system is required that should be complied with international recommendations. X-rays and gamma rays are a highly penetrating radiation and are widely used as a calibration source at different of medical facilities, industries and nuclear facilities. X-ray beam characterization for the calibration of radiation measuring equipment is necessary to meet the recommendations by international organizations. In the present study, characterization of X-ray beam generated from X-ray irradiator, Model: X80-225KV, Hopewell Design, Inc of Secondary Standard Dosimetry Laboratory (SSDL) has been performed in accordance with the ISO 4037-1 narrow–spectrum series. The beam characterization was done by the determination of Half-Value Layer (HVL), the Effective Energy, Homogeneity Coefficient (HC), Beam Quality Index and output air Kerma values. Experimentally measured HVL is compared with the values recommended by ISO-4037 and very small deviation is found. Filter thickness error or positioning of ionization chamber may be reason for this. The homogeneity coefficient is between 0.46 to 0.50. The homogeneity coefficient should be 1 but due to beam hardening, it is not possible. The Effective Energy, Eeff for ISO narrow spectrum series were calculated by empirical relation derived from Hubble mass attenuation coefficients. Hence, a set of conversion coefficients has been established for the effective energies of photon beam from air Kerma to dose equivalent (Sv/Gy) i.e., for ambient dose equivalent, H*(10) and personal dose equivalent, HP(10) & HP (0.07) for ISO beam code N40, N60, N80, N100, N120, N150 and N200 by empirical mathematical relationship applicable for the photons with energies between 10 keV to 10 MeV. The measured dose H*(10), HP(10) & HP(0.07) could be used for implementation of the new ICRU operational unit in radiation monitoring around the radiation facilities to protect human and environment in Bangladesh. An evaluation of backscattering factors for ISO water phantom and ICRU slab phantom has been conducted by MCNPX Code (version 2.6.0). for tube iii potential N80, N100, N120 and N140. The calculated values were compared with the values of experimental values. It is found that in personal dose monitoring, the backscattering factor contributes a considerable amount to the total absorbed dose. First of all, I am grateful to the almighty Allah for giving me the courage and enthusiasm to complete the project work. This project and the thesis paper are two of the greatest additions to my insufficient knowledge and experiences. I would like to express my deep gratitude to my research supervisor, Dr. Md. Shakiur Rahman, Chief Scientific Officer of secondary standard dosimetry laboratory (SSDL), Bangladesh Atomic Energy Commission (BAEC) and Director, Nuclear Safety Security and Safeguards Division (NSSSD) of Bangladesh Atomic Energy Commission (BAEC), Agargaon, Dhaka, for his patient guidance, enthusiastic encouragement, and useful critiques of this research work. I would also like to thank my co-supervisor, Dr. Md. Azizur Rahman, Professor, Department of Nuclear Science and Engineering, MIST, Dhaka, for his advice and assistance in keeping me updated on this project. I'd also like to thank my Department Head, Col. Molla Md. Zubaer, SPP, te., Former Department Head, Col. Salahuddin Zafor, our Program Coordinator. Lt. Col. Foysal Kadir, and Former Program Coordinator, Lt. Col. Md. Altab Hossain PhD., for their unwavering support and guidance from the very beginning and all through the seven-year eventful journey. Finally, and not least, I wish to thank all of my classmates and office staff for their support and encouragement throughout the study and thesis work. Thu, 01 Sep 2022 00:00:00 GMT http://dspace.mist.ac.bd:8080/xmlui/handle/123456789/798 2022-09-01T00:00:00Z