The aim of this study is to measure the feedback reactivity effects of the Bangladesh Atomic Energy Commission (BAEC) TRIGA Research Reactor (BTRR). Using digital instrumentation and control (I&C) system, feedback parameters such as reactivity coefficients of fuel, moderator and power effects are measured. At 300 kW reactor power, the obtained results for fuel temperature coefficient are 0.664 ꬿ/0c  and 0.765 ꬿ/0c  at C1 and D3 core positions respectively. The results for 100 kW to 400 kW reactor power show its increasing nature as the reactor power increases. The measured value of the moderator temperature coefficient is found to be 0.419 ꬿ/0c  at 500 kW reactor power. The power coefficient of reactivity is measured for several power ranges from 50 kW to 2.5 MW and the average value is found to be 0.0757 ꬿ/kw. The measured values of the reactivity coefficients are found to be in a good agreement with the GA Safety Analysis Report (SAR) provided for BTRR. Journal of Bangladesh Academy of Sciences, Vol. 42, No. 2, 183-190, 2018

To design a Reactor Pressure Vessel (RPV), material property like crack must be considered as it is an unavoidable property of materials. Presence of crack in materials must be kept within limit to prevent material's failure. So, crack propagation must be analyzed and observed. In this paper, crack propagation due to stress and materials fracture toughness of reactor pressure vessel cladding has been observed to estimate cumulative probability of crack failure using Probabilistic Fracture Mechanics (PFM). Average crack size is guessed as 3 mm and geometry factor is considered as 1.12 to analyze edge crack. Final crack analysis range has been found to be 1.8 mm with crack propagation rate of  30% of its average size. Variation of critical crack size and crack initiation point for several design stresses and fracture toughness has been investigated with probabilistic fracture mechanics technique. The observed crack propagation by calculating final crack size and the cumulative crack failure probability of the reactor pressure vessel materials are presented in this work.