TRIGA Mark-II Research Reactor Research Articles

Temperature distribution in the PGAA system: Collimator, shutter, and filter in TRIGA Mark II reactor

This article examines the impact of temperature on the steel collimator cap and the primary beam shutter. These components will be used to implement the Prompt Gamma Activation Analysis (PGAA) technique in the Moroccan TRIGA Mark-II research reactor. The steel collimator plug is essential for forming the neutron beam, while the main role of the shutter is to stop the beam when the channel is inactive. This study analyzes the effect of temperature on the collimator and shutter system, particularly focusing on the variation in maximum temperature over a month of operation with 8-h cycles per day, the behavior of temperature over 24 h, the total heat flux as a function of the length of the experimental device, the temperature distribution in mild steel (E235) and 304L stainless steel materials, and the total displacement and strain gradient as a function of temperature. All calculations were performed using COMSOL Multiphysics simulation software, based on the finite element method.

Assessment of Radiation Dose Associated with the Atmospheric Release of 41Ar from the TRIGA Mark-II Research Reactor in Bangladesh

Assessment of Radiation Dose Associated with the Atmospheric Release of 41Ar from the TRIGA Mark-II Research Reactor in Bangladesh

An approach of radiological impact analysis due to a simulated accident of the TRIGA research reactor of Bangladesh

The present study focused on the radiological impact analysis by assessing radiological doses due to a postulated TRIGA Mark-II research reactor accident at AERE, Savar, Dhaka, Bangladesh. Core inventory and source term of radionuclides from different groups such as noble gas, halogen, tellurium, alkali metal, lanthanide, alkaline earth metal, cerium and transition metal were analyzed for 365 days of operation at full power (3 MWth) due to a postulated nuclear accident. The HotSpot code was used for the dose calculation, and the dose values of TED, and Thyroid were calculated for a 40 km radius from the reactor site. Local meteorological data from Bangladesh Meteorological Department were used for the dose assessment and doses were calculated in the three dominant directions such as South, Southeast, and West. The TED and Thyroid dose values are assessed and compared to the PAZs dose criteria of the U.S. Environmental Protection Agency. It is found that doses are higher in stability class F, and evacuation, sheltering, and iodine prophylaxis are needed at longer distances. Since outside of the AERE campus is densely populated, these calculated data may help to ensure radiological safety by taking required protective actions like evacuation, sheltering, and iodine tablet consumption for the same type of accidental situation.

Assessment of radiological dose and emergency planning zones of TRIGA Mark-II research reactor of Bangladesh due to a postulated severe accidental condition

Radiological dose and emergency planning zones assessment has been conducted for dry and rainy seasons in Bangladesh for the TRIGA Mark-II research reactor. A postulated severe accident of a plane crash was considered in this analysis. Due to the full collapse of the reactor building and core damage, core inventory radionuclides were released at their highest release fraction. In this instance, doses due to the release of radionuclides such as Xe, Kr, I, and Cs were estimated using two different atmospheric dispersion models utilizing meteorological data collected from the Bangladesh Meteorological Department. The source term analysis has been performed using the ORIGEN 2.1 code. The HotSpot 3.1.2 and RASCAL 4.3 codes were used for the assessment of radiological doses. Based on the results of both the RASCAL 4.3 and HotSpot 3.1.2 codes, the TEDE and thyroid CDE were found to be substantially higher than the allowable limit in a large area. The dose values were compared to the dose standards of the IAEA General Safety Requirements (GSR) Part 7 and the United States Environmental Protection Agency (U.S. EPA) to identify the evacuation, sheltering and iodine prophylactic administration distances for both dry and rainy seasons. The evacuation and sheltering distances were obtained at approximately 2.25–5 km, 4.5–15 km, and 3–6.6 km, 3.7–8.5 km according to U.S. EPA dose criteria in the dry and rainy seasons, respectively. On the other hand, according to IAEA GSR part 7 dose criteria, the evacuation and sheltering distances are obtained at 1.5–3.5 km and 2.5–5.25 km in the dry and rainy seasons, respectively. Thus required emergency protective actions must be undertaken by the local authority to lessen the health impact on the general public.

Optimizing Neutronic and Photonic Performance in Irradiation Systems of Symmetric TRIGA Cores

The BAEC TRIGA MARK-II Research Reactor (BTRR) in Bangladesh has been used for a wide range of purposes, including basic and applied nuclear research and human resource development. Therefore, its core management should be flexible to meet various objectives with different priorities and to deliver the best possible outcome. In this study, neutron and gamma photon flux variation was studied at different radial and axial irradiation systems of the current core (C-0) as well as six symmetric reconfigurations (C-1, C-2, C-3, C-4, C-5, and C-6) of the existing BTRR using the universal MCNP code. While keeping the exact core component and material density, the symmetric reconfigured cores were modeled based on core criticality calculation and excess reactivity in the critical state. Finally, it was observed that the reconfigured core C-1 has the best neutronic and photonic performance at the irradiation systems compared to other reconfigured cores, against the reference core C-0.

Dose Distribution of Radioxenon Due to a Hypothetical Accident of TRIGA Research Reactor in Bangladesh

Radiological dose distribution owing to the deposition of 131m Xe, 133m Xe, 133 Xe, 135m Xe, 135 Xe, and 138 Xe on ground and immersion considering a postulated accident of TRIGA Mark-II research reactor has been assessed. The radiological dose distribution has been carried out in various directions with the help of Gaussian Diffusion Model. Local meteorological data such as average wind speed, frequency, etc. has been collected and evaluated for various directions around the reactor site. For all the dominant directions, the maximum dose values due to 131m Xe, 133m Xe, 133 Xe, 135m Xe, 135 Xe, 138 Xe and the total ( 131m Xe + 133m Xe + 133 Xe + 135m Xe + 135 Xe + 138 Xe) were observed within the limit 3.03E-7–1.23E-4 µSv/h, 1.01E-5–4.09E-3 µSv/h, 0.0003–0.14 µSv/h, 2.29E-5–9.26E-3 µSv/h, 0.002 –1.111 µSv/h, 1.11E-5–4.55E-3 µSv/h, and 0.003–1.269 µSv/h, respectively. Dose distribution was found to be dominant due to immersion and the contribution was 87.55 %. There is shortage of data regarding the release of radioxenon in the atmosphere during nuclear accident especially in the case of TRIGA type research reactor. This paper is the first such detailed study on atmospheric release of radioxenon and its dose distribution for a full power- reactor and the consequences towards the environment and public health. The result can be applied to develop the radiological protective measures and to prepare an emergency response plan for the TRIGA reactor site.

Inter-Code Comparison of Computational VERA Depletion Benchmark Using OpenMC, OpenMC-ONIX and DRAGON

This research focuses on the comparative analysis of the PWR fuel assembly based on VERA depletion benchmark problems using community-developed open source Monte Carlo code OpenMC, python based burnup code system ONIX (a coupling interface for Monte Carlo code OpenMC), and deterministic DRAGON code. The depletion analysis was performed using OpenMC and ONIX with ENDF/B-VII.1 nuclear data library, and DRAGON with SHEM-361 based DRAGLIB format library (ENDF/B-VII.1). The code-to-code analysis on the evolution of k eff , atom number density, and power distribution as a function of burnup has been performed and the result shows a good agreement with the maximum difference within 200 pcm at EOC. However small discrepancy around 90 pcm has been observed in k eff calculated by DRAGON compared to OpenMC in the presence of integral fuel burnable absorbers (IFBA). The above-mentioned codes have been validated successfully for the first time against PWR fuel assembly based on VERA depletion benchmark problems. It can be concluded that initial implementation of these codes at the Department of Nuclear Science and Engineering under Military Institute of Science and Technology, Dhaka, was successful and that further research works are to be performed to utilize these codes for depletion/neutronics calculation of existing 3MW TRIGA Mark-II research reactor and VVER-type power reactor that is to be commissioned in Bangladesh.

Experimental analysis of step reactivity insertion effect on reactor power, fuel temperature and reactor period in BAEC TRIGA research reactor

A series of experiments has been performed in a 3 MW BAEC TRIGA Mark-II Research Reactor to analyze the effects of positive and negative step reactivity insertion. The responses of reactor power and fuel temperature were determined. The reactor period is also calculated and a comparison is shown how the initial power affects the reactor period. This research illustrates the behavior of the delayed neutron in response to step reactivity insertion. After sudden step reactivity insertion (positive or negative), a sharp change in reactor power and fuel temperature has occurred initially, then the variation slows down followed by stable reactor period. In spite of inserting same positive and negative step reactivity, the amount of subsequent increase and decrease in reactor power and fuel temperature were not same. Significant variations in the reactor power and reactor period are found due to the effect of low initial power.

Modeling and simulating the induced effect on the steel collimator plug used in the PGAA facility of the Moroccan TRIGA Mark-II reactor under different neutron irradiation levels

The main objective of this study is to assess the behavior of the steel collimator plug (steel plug) dedicated to the Prompt Gamma Neutron Activation Analysis (PGAA) facility in the neutron beam tube (NB1) of the 2 MW Moroccan TRIGA Mark-II research reactor. The main function of this steel plug is to reduce the neutron and gamma beam cross section from 15 cm to 5 cm in diameter. This steel plug plays a crucial role in reactor safety because it replaces the original neutron beam plug while also stopping the entire incoming neutron beam. Three aspects were therefore involved in this study, including: i) the released heat induced by both neutron and gamma radiation, ii) the swelling effect caused by both radiation and temperature increases in the steel and iii) the radioactivity induced by neutron radiation in the steel plug. An MCNP6.2 model for the TRIGA Mark-II reactor was used to calculate the neutron spectrum inside the NB1 beam tube at the inlet part of the steel plug. A gamma spectrum of a 900 MW PWR reactor was used as input to account for the gamma radiation's effects on the steel plug. In order to select a convenient steel, two investigations were carried out for two grades of steel, namely mild steel (E235) and 304L stainless steel (SS304L). The results were calculated using an in-house FORTRAN-based program and validated using COMSOL, SPECTER codes, and “Neutron Activation and Scattering Calculator” tools.The results revealed that both the E235 and 304L steels are convenient from a safety point of view, although the E235 steel is preferred at the decommissioning stage.

Radiological Concentration Distribution of <sup>134</sup>Cs and <sup>137</sup>Cs Due to a Hypothetical Accident of TRIGA Research Reactor

The assessment of the radiological concentration of 134Cs and 137Cs owing to hypothetical accident of TRIGA Mark-II research Reactor at AERE, Savar, Bangladesh is presented here in this work. The concentration of 134Cs and 137Cs was estimated in different pathways consisting of the ingestion of plants, milk, and meat. The highest air concentration has been determined at 65 m distance from the core of the reactor. The maximum concentration passed off without delay simply after the accident in various directions. Local meteorological information such as average wind velocity and wind frequency were analyzed. Considering all directions, the highest concentration has been observed in the “S” direction. The concentrations of 134Cs and 137Cs were determined in ground, vegetation, milk and meat. The concentration of 137Cs is investigated to be higher than the 134Cs. The concentration of 134Cs and 137Cs was found to be lower in vegetation, milk, and meat than that of ground concentration. Overall, in this study, the concentration in meat has been investigated to be lower. In case of a reactor accident, the concentration assessment due to the ingestion of vegetables, milk, and meat will be a valuable guide for insuring radiological protection across the research reactor at AERE, Savar, Bangladesh.

Reactivity Initiated Transient Response of TRIGA with the Progress of Core Burnt

Normal 0 false false false EN-US X-NONE X-NONE /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Table Normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:10.0pt; font-family:"Times New Roman","serif";} This paper illustrates the effects on safety of TRIGA Mark-II research reactor of Bangladesh at its different steps of core burnt during reactivity induced transient. The modeling and simulation were carried by coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/RR based on neutronics data calculated previously by Monte Carlo code for different burnt states of TRIGA core. Three burn steps until 150 MWD have been considered for present analysis which are regarded here as beginning of cycle (BOC); middle of cycle, MOC (75 MWD); and end of cycle, EOC (150 MWD). Initially, the results of steady state analysis obtained from EUREKA code for these three burn steps have been verified with that of COOLOD code. Based on consistency of the results from the two codes, transient simulation has been conducted considering reactor with non-function of scram. Reactivity inserted amount ranges within 0.001 to 0.02 dk/k with three durations of insertions of 0.1s, 1s, and 5s. Major parameters such as reactor core maximum power and fuel clad maximum temperature have been reported. The analysis presents transient pattern of these parameters due to change in amount and duration of inserted reactivity. The maximum imposed reactivity that causes the fuel clad to exceed its design temperature at each burn step has been evaluated in this study. It is also observed here for each amount and durations of inserted reactivity, maximum value of both power and clad temperature found to decrease with the progress of core burnt. These results obtained from this analysis will be useful for reactor operators and management team during core upgrading and modification program.

Experimental cross sections of the 55Mn(n,γ)56Mn reaction at the neutron energies of 0.0334 eV and 0.0536 eV

The thermal neutron capture cross sections of 55Mn (n,γ)56Mn reaction were measured for the first time at 0.0334 eV and 0.0536 eV energies by using a monochromatic neutron beam from TRIGA MARK-II research reactor. Neutron fluxes during the experiments were measured by using 197Au (n,γ)198Au monitor reaction. After considering all the correction factors such as neutron self-shielding and gamma-ray attenuation effects, the cross sections obtained at 0.0334 eV and 0.0536 eV are 11.07 ± 0.64 b and 9.03 ± 0.52 b, respectively. Our experimental data at 0.0334 eV and 0.0536 eV are 4.3% and 1.0% lower than the theoretical evaluated data of ENDF/B-VIII.0 and JENDL-4.0 data libraries, respectively. By following 1/v dependence in the thermal energy region, the resultant cross section values extrapolated to the average thermal neutron energy 0.0253 eV were calculated and compared with reported literature values. Our experimental data for this reaction would be helpful to ensure the reliability of the cross sectional data evaluated by 1/v relation in the nuclear data libraries.

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.

Neutronic investigation of the thorium-based mixed-oxide as an alternative fuel in the TRIGA Mark-II research reactor – Part I: A beginning of life calculations

This article presents a neutronic investigation of the effect of replacing some uranium zirconium hydride (U-ZrH1.6) fuels with thorium-based mixed-oxide ((Th,U)O2) fuels in the TRIGA Mark-II research reactor without changing the core geometry, while the only variable is the fuel material.The possibility of utilizing thorium mixed oxide fuel in the TRIGA Mark-II reactor is assessed using the famous seed-blanket configuration or also known as Radkowsky Thorium Fuel (RTF) concept. Using the RTF concept, a long core lifetime can be achieved by enhancing the internal conversion rate of fertile to fissile materials, thanks to the thermal capture cross-section of 232Th that is around three times that of 238U. Additionally, on the fuel cycle aspect of using the thorium-based fuel is that the thorium cycle produces much less plutonium and other transuranic radioactive elements than uranium fuel cycle.Actually, neutronic and kinetic calculations, at the beginning of life conditions, for nine different seed-blanket configurations have been investigated using the MCNPX2.6 code. These calculations for all proposed configurations in comparison with the conventional U-ZrH1.6 fueled core were performed to pick out an optimal core(s) design ensuring the most efficient use of (Th,U)O2 fuel in it and with minimum changes in the main safety parameters. The parameters consist of effective multiplication factor keff, delayed neutron fraction βeff, neutron flux distribution, power distribution and radial hot rod power peaking factor (FHR), and shutdown margin (SDM).The calculation results exhibit that the highest initial criticality achieved is 1.07716. The results of the other neutronic investigations of thorium-based mixed-oxide fuel elements and their positions, including their numbers in the core are summarized and discussed in the following paper.

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.

Modeling and simulation of a TRIGA MARK-II research reactor using WIMSD-5B and CITATION codes

The main objective of this study is to analyse neutronic safety parameters of the Moroccan TRIGA Mark-II research reactor using the WIMSD-5B and CITATION computer codes. New 172-group libraries of multi-group constants for the lattice code WIMSD-5B have been generated for all isotopes presented in the TRIGA reactor core by processing nuclear data from ENDFB-VII.1, JENDL-4.0 and JEFF-3.1.1 using NJOY99. The lattice code WIMSD-5B was employed to generate multi-group cross sections in the suitable format that will be used by the 3-dimensional diffusion code CITATION. This later was used to calculate various neutronic safety parameters of the TRIGA Mark-II research reactor, such as reactivity excess and neutron fluxes profiles. The results of these calculations are compared to the results of Monte Carlo calculation based on MCNP code. A good agreement is achieved and the current computation scheme will be adopted for our further coupling neutronic/thermal-hydraulic study of the Moroccan TRIGA reactor.

Estimation of control rod worth, xenon effect on reactivity and power defect of BAEC TRIGA Mark-II research reactor

In this work, control rod worth, xenon (Xe) effect on reactivity and power defect have been measured by doing experiments in the BAEC TRIGA Mark-II research reactor (BTRR) and through established theoretical analysis. Firstly, to study the xenon-135 effect on reactivity, reactor is critical at 2.4 MW for several hours. Next, experiments have been performed at very low power (50 W) to avoid temperature effects. Moreover, for the power defect experiment, different increasing power level has been tested by withdrawing the control rods. Finally, it is concluded that the total control rods worth of the BAEC TRIGA Mark-II research reactor, as determined through this study, is enough to run the reactor at full power (3 MW) considering the xenon-135 and fuel temperature effects.

Investigation of the influence of various reflector elements material types on the core nuclear parameters of the 2 MW Triga Mark-II research reactor using the Monte Carlo code

The influence of variation in the kind of material of neutron reflector elements on the core neutronic parameters of the Moroccan Triga Mark-II research reactor was examined. The comparison was made between three kinds of reflector element materials: graphite, beryllium, and beryllium oxide. MCNP5 Monte Carlo computations of the effective neutron multiplication factor (${k}_{{eff}})$ of the reactor according to the number of reflector material elements inserted in the reactor core for these three reflector elements materials families were carried out. The results found could be used in order to increase the ${k}_{{eff}}$ parameter without changing the size of the reactor.

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.

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor.