MW TRIGA Mark-II Research Reactor Research Articles

Neutronic study of the 2-MW TRIGA MARK-II research reactor by the deterministic codes DRAGON5 & DONJON5

The main objective of this work is to perform a neutronic study of the 2 MW TRIGA MARK-II research reactor of the National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN), Rabat, Morocco and then validate the results by comparing the experimental values and those published for an ordinary 2 MW TRIGA MARK II research reactor. The core diffusion code DONJON5 and the lattice code DRAGON5 were coupled to perform a full model of the TRIGA core and their consistency and accuracy were established by benchmarking the TRIGA experiments. In this study, the nuclear data libraries ENDF/B-VII.1 and JEFF3.1 based on 172 energy groups were used. The group constants of all the reactor components were generated using DRAGON5 code and the collision probability method. These group constants were used then in the DONJON5 core code to calculate the multiplication factor, core excess reactivity, total and integral control rods worth as well as power peaking factors. Good agreement found between the calculated and measured results.

Modeling of the 2 MW TRIGA Mark-II Research Reactor Using the Deterministic Codes System DRAGON5 and DONJON5

This paper investigates the applicability of the DRAGON5 and DONJON5 code system to develop an accurate full-core model of the Moroccan 2 MW TRIGA Mark-II research reactor. The evaluated cross-section data library ENDFB.VII.1 with SHEM-295 group structure was used. A validation approach is applied to verify the consistency of the developed model. The results of the deterministic calculations based on ENDFB.VII.1/DRAGON5/DONJON5 scheme are compared with the experimental values available in the safety analysis report (SAR) as well as the Monte Carlo results. It was found that the DRAGON5 and DONJON5 code system results for core excess reactivity, critical size, reactivity worth of the five control rods, the power distribution and hot rod power peaking factor are in good agreement with measurements and the MCNP5 simulation values. Therefore, this study proves the capability of the DRAGON5 and DONJON5 code system to reliably simulate the TRIGA Type research reactors.

Validation Study of the Reactor Physics Lattice Transport Code DRAGON5 Based on DRAGLIB Libraries by TRX and BAPL Critical Experiments of Light Water Reactors for Neutronic Analysis of TRIGA MARK-II Research Reactor

The objective of this analysis is to present the validation study of the latest release of the reactor physics lattice transport code DRAGON5 by TRX and BAPL critical experiments of light water reactors for neutrons calculations of Moroccan 2 MW TRIGA MARK-II research reactor. This study is accomplished through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2, and BAPL-UO2-3 based on the open-source Draglibs nuclear data libraries ENDF/B-VII.1 and JEFF-3.1.2. These experiments are being considered as standard benchmarks lattices for testing and validating the reactor physics lattice transport code DRAGON5 as well as nuclear data libraries. The integral parameters—keffρ28, δ25, δ28 and C* —of the said lattices are calculated using the reactor physics lattice transport code DRAGON5 at a room temperature and compared to measured values as well as the Monte Carlo values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that the results calculated show good agreement with the experiment and the MCNP results. Therefore, this study reflects the validation, the adequacy, and the accuracy of the reactor physics lattice transport code DRAGON5 based on the open-source Draglibs nuclear data libraries ENDF/B-VII.0.1 and JEFF 3.1.2 evaluation with SHEM-295 group-structure and can also be important to execute neutronic analysis of thermal reactor such as the Moroccan 2 MW TRIGA MARK-II research reactor at CENM, achieved initial criticality on May 2, 2007.

Substantiation of Data Files of JEFF-3.1.2 for Safety Analysis of TRIGA Mark-II Reactor through the Scrutiny of Integral Parameter of Benchmark Lattices TRX and BAPL

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.

Measurement of neutronic safety parameters of the 3 MW TRIGA Mark-II research reactor

Measurement of neutronic safety parameters of the nuclear reactor is very important for reactor safety, efficient operation and experimental research. In this study, some neutronic safety parameters such as control rod worth, core excess reactivity, loss of reactivity with power increases, power defect, reactivity coefficients, cooling effect on fuel temperature and xenon poisoning were measured of the 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC). All of these safety parameters have significant effect on the reactor control system. Many of the reactor parameters which determine the reactivity of reactor are dependent on the temperature of the fuel, moderator and coolant. The TRIGA fuel elements have been designed to provide a significant prompt negative temperature coefficient of reactivity that allows safe reactor operation. The neutronic safety parameters were measured with digital instrumentation & control (I&C) system of the 3 MW TRIGA research reactor. The experimental results of neutronic parameters were found within the safety limit as mentioned in the Safety Analysis Report (SAR) of the BAEC 3 MW TRIGA research reactor. The present study will contribute to analyze the behavior of neutronic safety parameters that occur in the reactor core and to provide an experimental database for validation and improvement of the models of the reactor.

Measurement of control rod reactivity and shut down margin of 3 MW TRIGA Mark-II research reactor using analogue and digital I&C system

Measurement of reactor safety parameters is essential for reactor safety, operation and experimental research. Reactor control rod reactivity has been measured by the positive period method using analogue and digital instrumentation & control (I&C) system of the 3MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC). The BAEC research reactor had been operating with the analogue I&C system since September 1986–July 2011. The analogue I&C system has been replaced by a new digital I&C system on June 2012 and the reactor is now operating with the PC based digital I&C system. In the newly installed digital I&C system several modifications, upgradations and replacement tasks were carried out including control rod drive systems. After the installation of the digital I&C system, some nuclear safety parameters (e.g., control rod worth, core excess reactivity and shut down margin) were measured to ensure the safe operation of the reactor. In this study, the measured control rod reactivity, core excess reactivity and shut down margin were also compared with the previous measured (analogue system) values for the validation of the digital I&C system of the reactor. The measured safety parameters were found within the safety limit as mentioned in the safety analysis report (SAR) of the BAEC research reactor.

Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor

The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) heat flux and DNB ratio, axial fuel centreline temperature and compared. The comparison indicates that the calculated values are in satisfactory agreement between the codes. The data obtained in this investigation are largely far to compromise safety of the reactor. The results can also be used to upgrade the current core configuration of the TRIGA reactor.

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.

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.

Neutronics analysis of the 3 MW TRIGA Mark-II research reactor by using SRAC code system

This study deals with the neutronics analysis of the current core configuration of a 3 MW TRIGA Mark-II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Safety Analysis Report (SAR) values. The comprehensive neutronics code system SRAC was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Cross-section data library generated from JENDL-3.2 were used. The validation of the model against benchmark experimental results is presented. The SRAC predictions and the experimentally determined values are found to be in very good agreement, which indicates that the SRAC model is correctly simulating the TRIGA reactor.

Comparative Study of Water Absorption Behavior of Wood and Wood Plastic Composites of Simul Using Film Neutron Radiography

The film neutron radiography method was adopted for a comparative study of water absorption behavior in wood and wood plastic composites (WPC) of Simul using the thermal neutron radiography facility of 3 MW TRIGA MARK-II research reactor at Atomic Energy Research Establishment, Savar Dhaka, Bangladesh. Variations of optical density values due to water absorption of these samples were measured from neutron radiographic images. The WPC samples were prepared with impregnating monomer methylmethacrylate (MMA) under gamma radiation in order to study their water absorption behavior. The water absorption nature of Simul wood (Salmalia mamabrarica) and its composites, with and without urea as additive, was monitored by optical density measurements. The absorption of water in wood increases continuously with time. Where as, the water absorption by wood plastic composites is less compared to virgin wood; the composites with additive urea showed less water absorption than did those of virgin wood and the composites without urea.