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An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was 7.40×105Bq/cm3 and at the bay region was 3.39×10−5Bq/cm3. At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was 4.14×10−1Bq/cm3.

• Source term analysis of several radionuclides using ORIGEN 2.1 code. • Dose calculation using the HotSpot 3.1.2 and RASCAL 4.3 codes for dry and rainy seasons. • Observation of maximum TEDE and Thyroid CDE dose values. • Assessment of Evacuation and Sheltering distances for both dry and rainy seasons. 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. [ABSTRACT FROM AUTHOR]

Highlights • The control rod worth, core excess reactivity and shutdown margin have been measured. • Reactivity change is low with each control rod movement when the CT is filled with water. • The core excess reactivity decreases with increasing water height in the central tube. • BTRR reactivity worth with WCT is large enough to make the reactor shutdown at any condition. Abstract Nuclear safety and controlling system is a prime issue for the operation of a nuclear reactor. Control rod worth is one of the important neutronic safety parameters to ensure the safe operation of a nuclear reactor and experimental research. Control rod reactivity depends on neutron flux of the core and core configuration of a reactor. In this study, the control rod reactivity, core excess reactivity and shutdown margin with Wet Central Tube (WCT) of the BAEC TRIGA Research Reactor (BTRR) have been measured. The change of reactivity with the change of water height in the central tube has also been investigated in this study. The measured control rod reactivity, core excess reactivity and shutdown margin with WCT are then compared to those with the Dry Central Tube (DCT). The total worth of control rod of BTRR with DCT and WCT have been found to be 16.96 $ and 15.99 $ respectively at 775 MWds burn up. The values of the core excess reactivity and shutdown margin are found as 6.62 $ and 6.29 $ respectively for WCT, and 7.34 $ and 6.12 $ for DCT at 100 W reactor power level. It has been observed that the core excess reactivity decreases with increasing water height in the central tube of the reactor core. On the other hand, the shutdown margin increases with increasing water level. The values of the control rod worth, core excess reactivity and shutdown margin have been found to be within the safety limit as recommended in the General Atomic (GA) Safety Analysis Report (SAR) of BTRR. [ABSTRACT FROM AUTHOR]

The objective of the study has been to verify the calculated residual activity in the decommissioning waste of the TRIGA Mark II-type research reactor FiR 1 in Finland. Knowledge of the radioactive inventory of irradiated materials is important in the planning of decommissioning activities and is essential for predicting the radiological impact to personnel and the environment. Measurements are performed for low active material samples from outer parts of the reactor. Methods include gamma spectrometric measurements, composition measurements with mass spectrometry, oxidation measurements of especially 14C in graphite, and full combustion measurements of lithium-enriched shielding materials. Results are compared to estimates calculated with a combined Monte Carlo model of the reactor and a point-depletion code modeling the irradiation history. Decommissioning waste consists mainly of ordinary concrete, aluminum, steel, and graphite parts. Only preliminary measurements of low active samples are reported so far, but the same methods will be used later for characterizing and classifying dismantling waste. Some discussion of characterization requirements and future sampling is also included. [ABSTRACT FROM AUTHOR]

Within the bilateral project between the CEA Cadarache and the Jožef Stefan Institute (JSI) a wide variety of measurements using multiple fission chambers simultaneously inside the reactor core were performed. The fission rate axial profiles were measured at different positions in the reactor core and at different control rod configurations. A relative comparison of the calculated fission rates using the MCNP code and the measured fission rates was performed. In general the agreement between the measurements and calculations is good, with the deviations within the uncertainties. For better observation and understanding of the neutron flux redistribution due to the control rod movement, the neutron flux and fission rate had been calculated through the entire reactor core for different control rod configurations. The detector position with minimum signal variations due to the regulating and compensating control rod movement during normal operation was determined. The minimum variation is optimal in case we want to reliably determine the reactor power without influence of the regulating and compensating control rod positions. [ABSTRACT FROM AUTHOR]

The TRIGA research reactor ex-core nuclear instrumentation provides information on the reactor thermal power level, which is essential for the safe operation of the facility. In addition, exact knowledge of the reactor power level is necessary for the analysis of experiments and the normalization of reactor calculations. At the Jožef Stefan Institute TRIGA Mark II reactor the instrumentation is periodically calibrated using the calorimetric method. Though relatively simple and reliable it can be burdened with up to 30% uncertainty. In the paper a new calibration procedure using electrical heaters is presented, with which the heat capacity constant of the reactor pool is calculated to be C  = 19.6  kW h / K   ±  0.3  kW h / K and the uncertainty of the thermal power value is significantly reduced to approximately 2%. [ABSTRACT FROM AUTHOR]

The 3 MW 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 2.4 MW 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. [ABSTRACT FROM AUTHOR]

Air samples were taken at various locations around The University of Texas at Austin's TRIGA Mark II research reactor and analyzed to determine the concentrations of 37 Ar, 41 Ar, and 133 Xe present. The measured ratio of 37 Ar/ 41 Ar and historical records of 41 Ar releases were then utilized to estimate an annual average release rate of 37 Ar from the reactor facility. Using the calculated release rate, atmospheric transport modeling was performed in order to determine the potential impact of research reactor operations on nearby treaty verification activities. Results suggest that small research reactors (∼1 MWt) do not release 37 Ar in concentrations measurable by currently proposed OSI detection equipment. [ABSTRACT FROM AUTHOR]

The Bangladesh Atomic Energy Commission (BAEC) TRIGA Research Reactor is a unique nuclear installation of the country generally implemented for a wide variety of research applications and serves as an excellent source of neutron. During reactor operation it is necessary to measure and control the activity concentration of the pool water for fuel element failure detection and for the determination of contamination. The present study deals with the estimation of activity concentration for Na-24 present in water coolant produced as a result of 23 Na (n, γ ) 24 Na reaction. Several governing equations have been employed to estimate the Na-24 activity concentrations theoretically at different reactor power levels including maximum reactor power of 2.4 MW. From the obtained result it is ensured that the estimated Na-24 activity of 8.83 × 10 −3 μ Ci /cm 3 is not significant enough for any radiological hazard. Thus for ensuring radiological safety issues of the research reactor the assessment performed under the present study has an implication. [ABSTRACT FROM AUTHOR]

As decommissioning of a research reactor and a nuclear installation requires a long period of time from the decommissioning preparation work to the site remediation, the management of the data generated during the entire period of decommissioning is one of the most important tasks. In particular, the data obtained from research reactor decontamination and decommissioning activities can be important resources securing the safety and economic feasibility for other research reactor decommissioning. The owner of the research reactor and nuclear power plant need to submit decommissioning plan to the regulatory body at the starting stage of the research reactor and nuclear installation decommissioning project. The cost plan for decommissioning and the method for assessing the amount of exposure to protect workers must be stated in the decommissioning plan. This paper introduces the DES (Decommissioning Engineering System) that can be able to manage the data generated in the process of decommissioning of the TRIGA research reactor, to calculate an amount of waste, to evaluate decommissioning cost after deriving unit work productivity factor, and to predict the decommissioning process in advance. To verify the usability of this system and data integrity through connections among the unit systems, it describes the process to calculate the decommissioning cost using the data generated in dismantling an activated bio-shielding concrete in the TRIGA research reactor. As a result of the experiment to calculate the decommissioning cost with the TRIGA research reactor structure, it was found that the calculations were done precisely without flaw as the purpose of the experiment. Therefore, the DES can not only be used for other research reactors decommissioning, but also it is expected to be applied to other research reactors in the future. As a decommissioning cost between an activated concrete and a non-activated concrete according to the method of the dismantling procedure was significantly different, a study regarding the dismantling procedure needs more research. [ABSTRACT FROM AUTHOR]

The BAEC TRIGA research reactor (BTRR) is the only nuclear reactor in Bangladesh. Bangladesh Atomic Energy Regulatory Authority (BAERA) regulations require that nuclear reactor licensees undertake all reasonable precautions to protect the environment and the health and safety of persons, including identifying, controlling and monitoring the release of nuclear substances to the environment. The primary activation product of interest in terms of airborne release from the reactor is 41 Ar. 41 Ar is a noble gas readily released from the reactor stacks and most has not decayed by the time it moves offsite with normal wind speed. Initially 41 Ar is produced from irradiation of dissolved air in the primary water which eventually transfers into the air in the reactor bay. In this study, the airborne radioisotope 41 Ar generation concentration, ground level concentration and release rate from the BTRR bay region are evaluated theoretically during the normal reactor operation condition by several governing equations. This theoretical calculation eventually minimizes the doubt about radiological safety to determine the radiation level for 41 Ar activity whether it is below the permissible limit or not. Results show that the estimated activity for 41 Ar is well below the maximum permissible concentration limit set by the regulatory body, which is an assurance for the reactor operating personnel and general public. Thus the analysis performed within this paper is so much effective in the sense of ensuring radiological safety for working personnel and the environment. [ABSTRACT FROM AUTHOR]

This work presents the preparation of radiolabelled cis-dichlorodiammineplatinum (II), CDDP*, from its direct irradiation into a cadmium capsule, using the TRIGA MARK I IPR R-1 research reactor of the CDTN. The ability to detect CDDP* in Ehrlich tumour-bearing mice after administration via an intravenous route was evaluated. After 24 hours, blood and some organs were collected to determine the incorporated activity. The CDDP* showed a great chemical purity and high specific activity that resulted in an optimum in vivo detection. The CDDP* was taken up principally by liver, spleen and kidney. The CDDP* obtained from this condition was shown to be a good tool for biodistribution studies.O presente trabalho refere-se à obtenção de cis-diclorodiaminoplatina (II) radiomarcada, CDDP*, a partir de sua irradiação direta numa cápsula de cádmio. A irradiação foi realizada no reator de pesquisa TRIGA MARK I IPR R-1. Em seguida, foi avaliada a capacidade de detecção da radioatividade emitida por CDDP* após sua administração intravenosa em camundongos acometidos por tumor de Ehrlich. Após 24 horas, fez-se a coleta de amostras de sangue e de alguns órgãos, a fim de se avaliar o nível de radioatividade presente nos mesmos. A CDDP* obtida apresentou uma ótima pureza química e elevada atividade específica que permitiu um excelente nível de detecção. A CDDP* foi captada principalmente pelo fígado, baço e rins. Portanto, a CDDP* obtida na condição de bombardeamento supracitada mostrou ser uma ferramenta útil para a realização de estudos de biodistribuição.

The main aim of this paper is the development and validation of a 3D computational model of TRIGA research reactor using Serpent 2 code. The calculated parameters were compared to the experimental results and to calculations performed with the MCNP code. The results show that the calculated normalized reaction rates and flux distribution within the core are in good agreement with MCNP and experiment, while in the reflector the flux distribution differ up to 3% from the measurements. [ABSTRACT FROM AUTHOR]

Abstract: 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. [Copyright &y& Elsevier]

Highlights: [•] Measurement of reactor safety parameters is essential for operational safety. [•] Control rod reactivity, shut down margin and core excess reactivity were measured. [•] Measured values were found within the safety limit of BAEC research reactor. [•] Results are in good agreement with the digital and analogue I&C system. [Copyright &y& Elsevier]

Abstract: Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. [Copyright &y& Elsevier]

Abstract: Considering that the power of the IPR-R1 TRIGA reactor, located at the Nuclear Technology Development Center, Brazil, will be increased from 100kW to 250kW, some experiments were done in order to evaluate the magnitude of the reactivity effects associated with the reactor operation. The core excess of reactivity obtained was 1.99$, and the shutdown margin was 1.33$. The reactivity needed to operate the IPR-R1 reactor at 100kW was 0.72$, mainly due to the prompt negative temperature coefficient. A significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels. The loss of reactivity due to xenon poisoning after 8h of operation at 100kW was around 0.20$, and the highest reactivity loss value caused by a void inserted in the central thimble was 0.22$. From the results obtained, it was possible to balance all the determined reactivity losses with the reactivity excess available in the reactor, considering the present and the future reactor power operation. [Copyright &y& Elsevier]

Abstract: The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5¢/°C, and the reactivity measurements were performed at 10W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63¢/kW, and the temperature reactivity coefficient of the reactor was −0.8¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250kW), was 1.6$. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state. [Copyright &y& Elsevier]

Abstract: The on-line monitoring of several new process variables of the IPR-R1 TRIGA Reactor of the Nuclear Technology Development Center – CDTN became possible after the data acquisition and processing system implementation and the installation of one instrumented fuel rod in the reactor core. Several neutronics and thermo-hydraulics parameters are now registered, such as the operation power, the reactivity insertion in the core, the control rod position, the fuel and the water temperatures, and so on. Since the inherently safe operation of a reactor is dependent on the reactivity control, it is essential to have information on this parameter over many different temperature ranges. The fuel elements have been designed to provide a significant negative prompt temperature coefficient that allows safe reactor operation. The developed monitoring system gives the reactivity worth of the control rods, when the rod considered is inserted into the core or withdrawn from it, and also the loss of reactivity during the reactor operation. This paper describes the methodology and the results found with the on-line monitoring of the reactivity behavior of the IPR-R1 TRIGA Reactor. [Copyright &y& Elsevier]

Abstract: This work presents the preparation of radiolabelled cis-dichlorodiammineplatinum (II), CDDP*, sealed in a cadmium capsule. The irradiation of CDDP covered by cadmium, employing exposure times longer than 2h, demonstrated good chemical purity and high specific activity. This finding allowed a better detection of in vivo CDDP* and suggests that it may be a good tool for studies of long-term biodistribution of pharmaceutical formulations containing this drug. [Copyright &y& Elsevier]

The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×1013 n/cm2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future. [Copyright &y& Elsevier]