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In this research, we studied the performance of different vegetative and fruits electrochemical cells namely PKL, Aloe Vera, Tomato and Lemon juice electrochemical cells with load condition for 2:1 Zn/Cu based electrodes. It was also studied the variation of Load Voltage (VL), Load Current (IL), and Load Power (PL), with the variation of time for PKL, Aloe Vera, Tomato and Lemon juice electrochemical Cells. Among those cells the PKL electrochemical Cell was more efficient than the other three types of Cells regarding the load Current (IL), Load Voltage (VL), and Load Power (PL). However, we investigated the performance of different types of Cells without load condition for 1:1 Zn/Cu based electrodes. Moreover, the variation of open circuit voltage (Voc), short circuit current (Isc) and maximum power (Pmax) with the variation of time for those cells were explored. The discharge characteristic of the PKL electrochemical cell was more effective than the other three electrochemical Cells as the Open circuit voltage (Voc), Short circuit current (Isc) and Maximum Power (Pmax) are more stable and steady in comparison with others. Heat treatment temperature was a new approach by which we can enhance the performance of these electrochemical cells. Most of the results have been tabulated and graphically discussed.

The cross section of the 71 Ga(n,γ) 72 Ga reaction at 0.0334 eV was measured for the first time using monochromatic neutrons from powder diffractometer at TRIGA Mark II nuclear reactor. The 197 Au(n,γ) 198 Au reaction was used to monitor the neutron beam intensity. The HPGe γ-ray spectrometry was used to determine the radioactivity of the product radionuclides. The obtained cross section value amounted 3.42 ± 0.27 b is about 95% consistent with JENDL-4, but about 17% and 14% lower than that of the ENDF/B-VII and TENDL-2012 data libraries, respectively. The measured value at 0.0334 eV and the previous measured value at 0.0536 eV would be useful to confirm the reliability of the data evaluated by 1/ v relation in the above libraries.

The detection and measurement of radionuclides in feed coal, bottom ash and fly ash samples collected from Barapukuria 2×125 MW coal-fired thermal power plant in Dinajpur district of Bangladesh, have been performed by gamma ray spectrometry technique. The average activity concentrations of 226 Ra, 232 Th and 40 K in feed coal, bottom ash and fly ash samples were 10.46±5.24, 23.50±10.88 and 232.23±131.94 Bqkg -1 ; 56.91±2.77, 69.22±4.26 and 189.79±64.65 Bqkg -1 ; and 70.91±2.90, 115.26±5.79 and 205.53±65.56 Bqkg -1 ; respectively. These measured values were compared with other literature values. The calculated absorbed dose rates were found higher than the worldwide average values for both the bottom ash and fly ash samples. Moreover, the radium equivalent activity in all the samples was less than 370 Bqkg -1 and external hazard

Accurate determination of the radionuclide concentration in environmental sample depends directly on the accurate efficiency calibration in gamma ray spectrometry. The detection efficiency may vary due to some factors (density, sample filling height, matrix, measurement geometry, self-absorption, detection system, etc.) for voluminous samples. Some studies has been conducted in this work and presented in here. The efficiency calibration was performed for two HPGe detectors (relative efficiency of 20% and 40%), using samples (STDMS, STDMP, STDMPF, STDMRL, STDB, STDMC, & STDMG) of various densities (0.55 to 1.59 g/cm 3 ) in cylindrical containers (V=240 cm 3 ) for the gamma ray energy range of 121 keV to 2614 keV. Secondary reference material was prepared by using the standard reference liquid source 152 Eu. From the experimental result it can be seen that efficiency value is inversely proportional to the gamma ray energy. For lower gamma energies of selected nuclide, the efficiency value comparatively higher than the high energies of that nuclide. Efficiency value also varies with filling height of sample. Computer software ETNA was used for the efficiency transfer measurements of three samples (STDMP, STDMPF, & STDMRL) for the configuration of 20% relative efficiency detector. The efficiency value obtained from ETNA was compared with analytical efficiency value. Approximately 2.2% to 15.5% variation was found within the two systems. From the result it can be said that at the initial stage this kind of technical approach is not convenient for the environmental sample measuring laboratory due to the availability of input data for running the ETNA.

The activity concentrations of soil, water and fertilizer samples were determined by using high-resolution gamma spectrometry (HPGe detector: 40% relative efficiency) with a PC-based MCA system. The samples were collected from the area of a urea fertilizer factory, lagoon and Shitalakhya river in Narsingdi, Bangladesh. The activity concentrations of 226 Ra ranged from 3.16 ± 0.32 to 10.28 ± 0.55 Bq.kg-1 , 1.22 ± 0.41 to 7.36 ± 0.42 Bq.L-1 and 3.55 ± 0.33 to 90.65 ± 3.17 Bq.kg-1 for soil, water and fertilizer samples, respectively. The 232 Th activity concentrations ranged from 4.89 ± 0.45 to 15.82 ± 0.45 Bq.kg-1 , 1.21 ± 0.06 to 8.59 ± 0.37 Bq. L-1 and 4.76 ± 0.25 to 26.38 ± 1.40 Bq.kg-1 for soil, water and fertilizer samples, respectively. The 40 K activity concentrations ranged from 24.96 ± 0.23 to 60.49 ± 0.56 Bq.kg-1 , 7.48 ± 0.53 to 35.48 ± 0.24 Bq.L-1 and 3.55 ± 0.05 to 3051.71 ± 19.53 Bq.kg-1 for these samples, respectively. The radium equivalent activity, the hazard indices, the gamma activity concentration index, the indoor absorbed dose rate and the corresponding annual effective dose were estimated for the potential radiological hazard of the collected samples. The calculated values of the representative level index values (Igr) for all samples of the study area are lower than unity except the MOP sample. The activity ratios were also measured. These values are compared with reported values for other countries of the world. The results of the comparison studies show that the radioactivity concentrations and other radioactive indices of the samples of the study area are below the internationally accepted maximum permissible values. Therefore, this region is safe from any radiation hazard and no significant radiological threat was observed to the population of the study area.

The neutron capture cross section for the 139 La(n, γ) 140 La reaction at 0.0536 eV was measured for the first time using monochromatic neutrons coming from a Triple Axis Spectrometer at the TRIGA Mark-II reactor. The neutron beam intensity was determined by irradiating Au-monitor foil together La 2 O 3 target. The radioactivity of the products was determined via high resolution γ-ray spectrometry. The obtained cross section value is 6.13±0.3b, which is 1.34% and 0.21 % lower than that in the ENDF/B-VII and JENDL-3.3 data library, respectively.

The thermal neutron-induced activation cross section for the 186 W( n , γ ) 187 W reaction was measured at 0.0536 eV neutron energy using TRIGA Mark-II research reactor, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The 197 Au( n , γ ) 198 Au monitor reaction induced in a high-purity gold foil was used to determine the effective neutron beam intensity. The activities induced in sample and monitor foils were measured nondestructively by a high-resolution HPGe γ -ray detector. The present experimental cross-section value is the first one at 0.0536 eV. The obtained new cross section that amounts to 26.6±1.6 b is 2% higher than the recently reported data in ENDF/B-VII and 5% lower than that of JENDL-3.3.

The neutron capture cross-section for the 71 Ga(n, γ) 72 Ga reaction at 0.0536 eV energy was measured using activation technique based on TRIGA Mark-II research reactor. The 197 Au(n, γ) 198 Au monitor reaction was used to determine the effective neutron flux. Neutron absorption and γ-ray attenuation in gallium oxide pellet were corrected in determination of cross-section. The cross-section for the above reaction at 0.0536 eV amounts to 2.75 ± 0.14 b. As far as we know there are no experimental data available at our investigated energy. So far we are the first, who carried out experiment with 0.0536 eV neutrons for cross-section measurement. The present result is larger than that of JENDL-3.3, but consistent within the uncertainty range. The value of ENDF/B-VII is higher than this work. The result of this work will be useful to observe energy dependence of neutron capture cross-sections.

The objective of this paper is to select an inimitable nuclear data library from JEFF-3.1.2 & JENDL-4.0u which are valid for theoretical safety analysis of TRIGA MARK-II research reactor. In this work the integral parameter (such as k eff , ρ 28 , δ 25 , δ 28 , C ∗ ) of TRX and BAPL benchmark lattices of thermal reactors are compared with the experimental result by Cross Section Evaluation Working Group (CSEWG), USA. The nuclear data processing code NJOY99.0 is used to generate 69-group cross-section library from the basic evaluated nuclear data files JEFF-3.1.2 & JENDL-4.0u. TRX and BAPL benchmark lattices are modeled with optimized inputs, which are suggested in the final report of the WIMS Library Update Project Stage-I (WLUP). The integral parameters of five uranium-fuel thermal assemblies: TRX-1 and TRX-2 and BAPL-UO 2 -1, BAPL-UO 2 -2 & BAPL-UO 2 -3 are calculated with the help of reactor lattice code WIMSD-5B based on the generated 69-group cross-section library. Form the comparison of the integral parameters with the experimental values it is found that the obtained results between the two libraries is nearly alike with some uncertainties. But the degrees of uncertainties for the values of integral parameters of JEFF-3.1.2 library are comparatively less. JEFF-3.1.2 is the better library and could be selected for the neutronic calculation of TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh.

Shielding properties of polyethylene (PE), heavy concrete (HC), ordinary concrete (OC), lead (Pb) and their multilayers PE + HC, PE + OC and PE + Pb have been measured. The results of this investigation reveal the effectiveness of the multilayers over their respective composites for neutron shielding.

The aim of this analysis is to bear out the nuclear data files of JEFF-3.1.2 for theoretical safety analysis of a 3 MW TRIGA MARK-II research reactor is custom-made at AERE, Dhaka, Bangladesh through the study of integral parameters of benchmark lattice TRX and BAPL of thermal reactor. The basic evaluated nuclear data files of JEFF-3.1.2 are selected for TRIGA reactor and processed by using nuclear data processing code NJOY99.0. Different cross-sections of U-235 and U-238 are computed from the NJOY output of the evaluated nuclear data library. The 69 group cross-section library is engendered from the processed file for reactor code WIMSD-5B. From the generated 69 group cross-section library, the integral parameters of yardstick lattices TRX and BAPL are premeditated by using cell code WIMSD-5B. The calculated integral parameters are compared to the deliberated values as well as the consequences of Monte Carlo Code MCNP. From the assessment it is found that all the integral parameters are in good concurrence with some suspicions. Through benchmarking the integral parameters of TRX and BAPL lattices this analysis reflects the support to the evaluated nuclear data files of JEFF-3.1.2 for safety analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.