TRIGA Mark-II Research Articles

Quality Study of Automated Machine Made Environmentally Friendly Brick (KAB) Sample Using Film Neutron Radiography Technique

Neutron radiography (NR) technique has been adopted to study the internal structure and quality of the KAB bricks made by Hoffman kiln method. Thermal neutron radiography facility installed at the tangential beam port of 3 MW TRIGA Mark-II Research Reactor, AERE, Savar, Dhaka, Bangladesh is used in the present study. Measurements were made to determine the internal structure and quality of the automated machine made environmentally friendly brick sample. In this case, optical density/gray values of the neutron radiographic images of the sample have been measured. From these measurements, the porosity, water penetrating height, water penetrating behavior, initial rapid absorption of water (IRA), elemental distribution/homogeneity and incremental water intrusion area in the sample have been found. From the observation of different properties, it is seen that, homogeneity of the Hoffman kiln brick KAB is not perfectly homogeneous and contains small internal porosity; the incremental water intrusion area is very poor, and the water penetrating height through the two edges is higher than the middle part; the initial rapid absorption (IRA) rate is also very poor and the water penetrating behavior of the samples is different as like as stair, capillary, wave and zigzag shape. From these points of view, it is concluded that the quality of the environmentally friendly brick KAB is better. The results obtained and conclusion made in this study can only be compared to the properties of bricks produced under similar conditions with similar raw materials.

Operation and Control of TRIGA Nuclear Research Reactor with PLC

Abstract — Nuclear reactor operation and control is a sophisticated system involving modern equipments and techniques. This paper proposes an operation and control feature of TRIGA Mark-II research reactor and discusses the structure and implementation of smooth automatic operation and control using programmable logic controller (PLC). The automation of the reactor and accurate power generation by demand is carried out by inserting or withdrawing control rods. Four trip conditions and one external faults (generalized for any external mechanical or electrical fault) are taken into account to shut down the reactor suddenly (SCRAM). This paper proposes better flexibility in operation & control and low cost & time for trouble shooting with PLC compared to the instrumentation currently offered by General Atomics for TRIGA reactor. Index Terms— I. INTRODUCTION A nuclear reactor is a system in which a nuclear reaction named fission (splitting the atom) chain reaction is initiated, sustained and controlled. Reactors are used for generating electricity, moving aircraft carriers and submarines, producing medical isotopes for imaging and cancer treatment and for conducting research. The energy in form of heat produced by fission reaction is carried out by coolant generally water that produce steam. The thermal power generation is manipulated by control rods. When inserted (moved down) in the reactor core, control rods absorb neutron and reduce generation of power. On the other hand when control rods are withdrawn (moved up) the number of fission and power level increases.

Validation study of SRAC2006 code system based on evaluated nuclear data libraries for TRIGA calculations by benchmarking integral parameters of TRX and BAPL lattices of thermal reactors

The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20°C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh.

Computational analysis of thermo-hydraulic behavior of TRIGA research reactor

Key thermal hydraulic parameters of the 3MW TRIGA Mark-II research reactor operating under steady-state conditions were investigated using the thermal hydraulic codes NCTRIGA, PARET and COOLOD-N2. Results of the neutronic analysis performed by 3-D Monte Carlo code MCNP4C were used in NCTRIGA and coupled output of neutronic analysis carried out by using 3-D diffusion code CITATION and 3-D Monte Carlo code MCNP4B2 were used in the PARET to study the steady-state thermal hydraulic behavior of the reactor. To benchmark the NCTRIGA, PARET and COOLOD-N2 models, data were obtained from different measurements executed by thermocouples in the instrumented fuel elements (C1 and D2) and the hottest fuel element (C4) during the steady-state operation both under forced and natural convection mode and compared with the calculation found to be quite consistent. The mass flow rates needed for input to PARET and COOLOD-N2 were taken from final safety analysis report (FSAR) for a downward forced coolant flow equivalent to 3500gpm. For natural convection cooling of reactor, mass flow rate was generated using NCTRIGA code. The testing of the NCTRIGA, PARET and COOLOD-N2 model calculations through benchmarking the available TRIGA experimental and operational data showed that NCTRIGA, PARET and COOLOD-N2 codes can successfully be used to analyze the thermal hydraulic behavior of the reactor for the steady-state operation under both natural and forced convection mode of coolant flow to predict the safety margins. The result obtained in this investigation can be used for upgrading the current core configuration of the TRIGA reactor.

Dose distribution of 131I, 132I, 133I, 134I, and 135I due to a hypothetical accident of TRIGA Mark-II research reactor.

An attempt has been made in the work to assess radiological dose due to the deposition of 131 I, 132 I, 133 I, 134 I, and 135 I on ground, vegetation, milk and meat considering a hypothetical accident of TRIGA Mark-II research reactor at AERE, Savar. Contribution of dose from ingestion of vegetation due to 131 I has been found to be dominant (62.96%). Also found that contribution of 132 I from immersion is higher (53.67%) and contribution of 133 I in meat is higher (45.64%). The contribution due to 134 I in both from immersion (46.13%) and from ground deposition (47.19%) was dominant while the contribution of 135 I from ground deposition (30.11%), inhalation (26.77%) and immersion (25.92%) were higher. Final result says that the vegetation is the dominant contributor (66.91%) for iodine.

Radiological Concentration Distribution of <sup>131</sup>I, <sup>132</sup>I, <sup>133</sup>I, <sup>134</sup>I, and <sup>135</sup>I Due to a Hypothetical Accident of TRIGA Mark-II Research Reactor

The present work gives a methodology for assessing radiological concentration of 131I, 132I, 133I, 134I, and 135I due to a hypothetical accident of TRIGA Mark-II research Reactor at AERE, Savar, Bangladesh. The concentrations were estimated through different pathways like ingestion of vegetation, milk, and meat from air and ground deposition. The maximum air concentrations for all 16 directions were found at 110 m distance from the core of the reactor and it was found to be highest in the southern (S) direction. The maximum ground concentration occurred immediately just after the accident in different directions. In all pathways, the most concentration was found to be in S-direction. The concentrations in vegetation of 131I, 133I, 135I were significant, while no concentrations of 132I and 134I were observed. The concentration in vegetation for 131I was found to be highest than all other isotopes of iodine. The concentrations of 133I were found to be higher and concentrations of 134I were observed to be lower in both milk and meat compared to other radio isotopes of iodine. In the case of a radiological accident, the results of the present study will be a valuable guide for adopting radiological safety measures for radiation protection against the ingestion of vegetables, milk and meat from around the research reactor at AERE, Savar, Bangladesh.

Thermal hydraulic transient study of 3 MW TRIGA Mark-II research reactor of Bangladesh using the EUREKA-2/RR code

EUREKA-2/RR code has been used for the analyses of reactivity insertion accident (RIA) and loss of flow accident (LOFA) of 3 MW TRIGA Mark-II research reactor of Bangladesh. Transient characteristics of different parameters such as core power, fuel temperature, clad temperature, departure from nucleate boiling ratio (DNBR) due to the different form and magnitude of reactivity insertion has been focused. It is found from the analysis that the magnitude of insertion reactivity and the reactor operating power during this insertion impose a total effect on the core safety. Also, transient effects on reactor were studied for 15% loss of flow of the primary coolant. Provided the scram system is available, the reactor is found to shutdown safely in both cases. From these two studies in series, it is seen that EUREKA-2/RR is well suited for the analyses of reactor safety parameters with good approximations.

GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×1011 ncm2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent–divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line (http://www.info.cern.ch/asd/geant4/geant4.html [4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al2O3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al2O3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to define the design of the moderator in the inlet of the radiation channel. A graphite block of 22 cm thickness seems to be the optimal neutron moderator. The results showed that the combination of 5 cm of bismuth with 5 cm of sapphire permits the filtration of gamma-rays, epithermal neutrons as well as fast neutrons in a considerable way without affecting the neutron thermal flux.

Soft computing technique for power control of Triga Mark-II reactor

In this study, a trajectory tracking fuzzy genetic controller for Istanbul Technical University Triga Mark-II nuclear research reactor design approach is given. Power output of reactor is controlled along the predefined trajectory by fuzzy logic controller. Designed zero order Sugeno type fuzzy logic controller membership boundary value and rule weights are found by genetic algorithm. Non-chattering control with smooth control surface is also achieved using constrained fitness functions. Simulation results shows that reactor power successfully tracks the given trajectory under various working conditions and reaches the desired power level within the determined period within small tracking error.

Measurement of Natural and Artificial Radioactivity in Soil at Some Selected Thanas around the TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka

The activity concentration of natural and fallout radionuclides in the soil at some selected Thanas around the TRIGA Mark-II Research Reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka were measured by using a high purity germanium detector (HPGe). The study revealed that only natural radionuclides were present in the samples and no trace of any artificial radionuclide was found. The average activity concentration of 238U, 232Th and 40K were found to be 37.8 ± 5.6 Bq.kg-1, 58.2 ± 11.0 Bq.kg-1 and 790.8 ± 153.4 Bq.kg-1 respectively. The radium equivalent activity (Req), absorbed dose rate (D), external radiation hazard index (Hex) and internal radiation hazard index (Hin) were also calculated to find out the probable radiological hazard of the natural radioactivity.

On Line Measurement of Reactivity Worth of TRIGA Mark-II Research Reactor Control Rods

The reactivity worth measurement system for control rods of the TRIGA MARK-II research reactor of Bangladesh has been design and developed. The theory of the kinetic technique of measuring reactivity has been used by this measurement system. The system comprises of indigenous hardware and software for online acquisition of neutron flux signals from reactor console and then computes the reactivity worth accordingly. Here for the TRIGA MARK-II research reactor, the reactivity measurement system was implemented with a dedicated circuit assembly and a conventional personal computer. A high-level Visual Basic real-time programming has been developed for data acquisition, reactivity calculation, online display (numerically as well as graphically), saving data, etc. To measure reactivity worth of TRIGA reactor control rods the rod drop experimental technique has been adopted. The results of tests experiments, carried out with the rod drop method for measuring various reactivity worth of control rods have been presented in the paper. A comparison between this results with the results using period method and that of computation method, demonstrated that the response of this reactivity measurement system is fast enough to monitor and measure the safety-related reactivity and power excursions in the reactor.

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k eff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal–hydraulics study of the TRIGA core.

Experimental cross section for the 139La(n, γ)140La reaction at 0.0536 eV

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.

Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors

The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2-1, BAPL-UO 2-2 and BAPL-UO 2-3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.

Thermal neutron capture cross sections for the 152Sm(n,γ) 153Sm and 154Sm(n,γ) 155Sm reactions at 0.0536 eV energy

The neutron capture cross sections for the 152Sm(n,γ) 153Sm and 154Sm(n,γ) 155Sm reactions at 0.0536 eV neutron energy were measured using an activation technique based on the TRIGA Mark-II research reactor, relative to the reference reaction 197Au(n,γ) 198Au. The activity was measured nondestructively using gamma-ray spectroscopy. Our measured values at this neutron energy are the first ones and are compared with 1/ v based evaluated cross sections reported in the ENDF/B-VII and JENDL-3.3 libraries. The measured value for the 152Sm(n,γ) 153Sm reaction is 0.28% lower than JENDL-3.3 and 0.48% higher than ENDF/B-VII. Our value for the production of 155Sm is about 3% and 2.3% higher than the evaluated value with ENDF/B-VII and JENDL-3.3 at 0.0536 eV, respectively.

A qualitative identification and analysis of hazards, risks and operating procedures for a decommissioning safety assessment of a nuclear research reactor

Korea Research Reactor-1(KRR-1, TRIGA Mark-II type reactor), the first nuclear research reactor in Korea, is being prepared for a decommissioning. The decommissioning methods and procedures of KRR-1 ought to be based on its structural conditions and radiological characteristics. Also, a systematic approach to the decommissioning tasks must be followed by reviews and assessments of the decommissioning workers’ safety. In this paper, the characterizations of the KRR-1 facility and its installations are the focus of attention. The expected radiation exposure data for the main decommissioning objects are measured and the appropriate health and physics plans are established. Based on the hazards and operability (HAZOP) criteria of expertise and the experienced data from the KRR-2 decommissioning project, the first decommissioning project completed in Korea, the hazards and precautions to avoid harm to the workers are identified and the radiological accidents are also analyzed. In the end, the operating procedures for the decommissioning tasks of KRR-1 are established.

A comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors using the cross-section library generated from JENDL and JEF evaluated data files

The objective of this paper is to present the results of comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors. The nuclear data processing code NJOY'99 was deployed for the generation of the 69-group cross-section library from the basic evaluated nuclear data files JENDL-3.2 and JEF-2.2. TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project Stage-I. The inputs were the results of a detailed parametric study of the WIMS input options and also optimized for accuracy. The integral parameters (such as k eff, ρ 28, δ 25, δ 28, C ∗) of five uranium-fuel thermal assemblies: TRX-1 and TRX-2 and BAPL-1, BAPL-2, and BAPL-3 were calculated with the help of WIMSD-5B code based on the generated 69-group cross-section library. The calculated results are compared with those of experiments and it is found that the obtained results between the two libraries are in good agreement with each other. Besides, the calculated integral parameters are also well consistent with the measured values, which reflect the validation of the generated 69-group cross-section library and this library thus obtained is necessary to meet up the nuclear data for neutronics calculation of TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh.

Measurement of neutron capture cross-section of the 71Ga(n, γ) 72Ga reaction at 0.0536 eV energy

The neutron capture cross-section for the 71Ga(n, γ) 72Ga reaction at 0.0536 eV energy was measured using activation technique based on TRIGA Mark-II research reactor. The 197Au(n, γ) 198Au 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.

Measurement of the thermal neutron capture cross section and the resonance integral of radioactiveHf182

The neutron capture cross sections of the radioactive isotope {sup 182}Hf (t{sub 1/2}=8.9x10{sup 6} yr) in the thermal and epithermal energy regions have been measured by activation at the TRIGA Mark-II reactor of the Atomic Institute of the Austrian Universities in Vienna, Austria, and subsequent {gamma}-ray spectroscopy of {sup 183}Hf. High values for the thermal (kT=25 meV) cross section {sigma}{sub 0}=133{+-}10 b and for the resonance integral I{sub 0}=5850{+-}660 b were found. Additionally, the absolute intensities of the main {gamma}-ray transitions in the decay of {sup 182}Hf have been considerably improved.

Neutron capture cross-section measurement for the 186W( n, γ) 187W reaction at 0.0536 eV energy

The thermal neutron-induced activation cross section for the 186W( n, γ) 187W reaction was measured at 0.0536 eV neutron energy using TRIGA Mark-II research reactor, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The 197Au( n, γ) 198Au 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.