The radioactive product released from a nuclear reactor is one of the most vital safety and health physics worries in nuclear research and power reactors. Usually, nuclear reactor workers are exposed during the operation of a nuclear reactor. The product of the dose rate and the time spent in radiation-controlled area is the measurement of the external doses forms the occupational exposures in nuclear reactor. The dose rate is calculated by the amount of radioactive material in reactor, the quality of the core water, the condition of top surface and the range of decontamination used. The time is ascertained by the amount of work to be done, the continuous operation time of reactor and the range of remote technology used. As the presence of personnel is unavoidable inside reactor side during operation time, the release must be constantly monitored through the passage of time. The accidental release of radiative material above the safety level is very much hazardous to the people both inside and outside of reactor site.

BAEC TRIGA Research Reactor (BTRR) is the first and only research reactor in Bangladesh. It was commissioned in 1986 with the purpose of expanding nuclear research as well as radioisotope production and academic purpose. It contains facilities for various nuclear experiments, sample activation and manpower training (SAR, 2006; Haque et al., 2025). Like every reactor around the world, BTRR is also licensed on the ground that there will be no unjustifiable hazard or serious radiation concerns on public health and security. It is necessary to assess the radiation doses at various important points at reactor site during high power operation period to make sure that there is no undue hazard for reactor operating personnel, radiation workers and the most general public. In the previous study, an experimental study was performed to determine the radiation dose at different strategic points of BTRR (Haque et al., 2017). The present study is performed inside reactor building as well as outside the reactor building. An important part of the present study is that the number of points of dose measurement point has been increased to insure maximum undue hazard points so that no points can remain unmeasured which will cause health concerns to reactor personnel or radiological workers even general public. The dose rate (radiation) surveys were performed to investigate the radiological hazard level triggered by alpha, beta and gamma radiation. Alpha (α) particles have a few cm ranges in air, and are not considered hazardous for external radiation because they have not enough energy to penetrate the outer skin layer; therefore there was no need to investigate it. But neutrons and gamma rays are highly penetrating. They penetrate the skin and get scattered in various body tissues. Neutron is considered the most important external radiation hazard (Rahman et al., 2014). 

For maintaining a safe and sound reactor facility in addition with the security of the individuals from radiation, radiological control is one the most signi-ficant practices. Taking into account this significance, present study was carried out to identify the radiation type and determination of radiation doses to reactor personnel and general population. This paper summarizes the measurement of dose rates for beta (β), gamma (γ) and neutron (n) at different power level of BTRR at various important points. Dose rates were measured at top surface of the reactor pool with different power levels. The reactor top surface is generally used for numerous experiments performed by radiation workers and reactor operating individuals. The water coolant continuously flowing through the reactor core during operation is exposed to a strong neutron flux which results in production of different radioisotopes. 16N, 41Ar, 19O, 17N, 3H and 14C decayed by beta, gamma and neutron emission are considered hazardous of all the radionuclides formed in the water coolant. 16N is the source of gamma radiation of high energy about 6.13 MeV (haque et al.,). The investigation work for this study has been carried out using different beta, gamma and neutron survey meters. The determined dose rates have been studied, observed and estimated to minimize threat on the personnel involved with reactor facility including reactor operators, workers, user groups, visitors and also environment according to the recommendations set by IAEA safety series (BSS-115) and International Commission on Radiological Protection (ICRP).

Brief Description of BTRR

The BTRR is a light water cooled, tank type research reactor. The core uses a solid, homogeneous fuel-moderator element in which the zirconium hydride (ZrH1.6) is homogeneously combined with 20% uranium. The reactor core is located at the bottom of an aluminum tank and the tank is surrounded by a concrete shield structure. 

The tank has an outside diameter of 78 in, a depth of 27 ft, and a minimum thickness of ¼ in. The tank is filled with demineralized water to a depth of about 26 ft 7 in., providing approximately 20 ft 6 in. of shielding water above the top of the core. The reactor core is designed for 100 fuel elements, 18 graphite elements, 1 Dry Central Thimble (DCT), 1 Pneumatic Transfer System Irradiation Terminus and 1 Am-Be Neutron Source. 100 fuel elements include 93 standard fuel elements (FE), 5 fuel follower control rods (FFCR) and 2 instrumented fuel elements (IFE) (Rahman et al.,). All these elements are placed and supported in between two 55.25 cm diameter grid plates and arranged in a hexagonal lattice. Two instrumented fuel elements (IFE) in the core measure the fuel temperature during operation.  Three operation mode are possible in the design of BTRR which are steady state mode, square wave mode and pulse mode. 

Fig. 1: Different points of BTRR Facility.

The steady state operation mode can be managed under two cooling modes namely Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). The NCCM can be used for a power level up to 500 kW. For operation at higher levels up to 3 MW, FCCM is functioned. The BTRR has four water systems: (1) Primary water system, (2) Secondary water system (3) Emergency core cooling system (ECCS), (4) On-line purification system [SAR]. There are four beam ports available for providing neutron beams for the purpose of various experiments. Six control rods basically control the reactivity of the reactor and the neutron absorber Boron Carbide (B4C). The different measurement points of the BAEC TRIGA Research Reactor for this study is shown in Fig. 1.

Fig. 2: Different measurement points at the reactor/pool top of BTRR Facility.

Measurement points: 1. Public gallery (Glass wall surface), 2. Reactor control room (Glass wall surface), 3. Reactor operator (RO) Sitting chair, 4. Rail surface of reactor top, 5. Grating surface of reactor top, 6. Radial beam port-I (on leakage surface), 7. Radial beam port-I (on shielding surface), 8. Radial beam port-II 9. Tangential beam port, 10. Piercing beam port, 11. Thermal column, 12. Primary cooling valve (MOV-1), 13. Staff sitting room, 14. Main entry door (Ground floor), 15. Rabbit room, 16. Secondary pump room, 17. Heat exchangers surface, 18. Surface of ion changer resin column, 19. Shielding surface of ion exchanger resin column, 20. Primary pump room, 21. Outside the decay tank wall surface, 22. Outside reactor building (decay tank side).

Measurement of dose rate and experimental analysis

Radiation dose rates were noted at different points of BTRR during reactor operation at different power level with the help of calibrated digital survey meters. Dose rates were recorded three times at each point in step of every three minutes with same operational condition of reactor. The average of recorded data and average of standard deviations was noted as the total dose rate. Same methods were repeated for each operational condition for each data recording points. During dose rate measurement at reactor pool top surface with NCCM, survey meters were placed on the pool surface at a fixed position 20 cm apart from the pool center because of experimental limitation and previous method was followed for recording data. The exposure time of three minutes was firmly kept up for each measurement.

Dose rate measurement and detection was performed by using a set of measuring instruments. Name of survey meters used for detecting gamma radiation and measuring dose rate was Mirion RDS-31 which is a handheld seamless radiation monitoring equipment along with GMP-25 probe. Name of survey meters used for detecting neutron and measuring dose rate was VF PNM-01. These devices are well calibrated from the manufacturers end and the errors were always below 5% at calibration time and recommendations from IAEA safety series were strictly followed for the process. 

With the help of the set of instruments, an overall assessment of radiation exposure for reactor personnel, worker, and general public was achieved.

Experimental Data and Analysis

As Radiation workers work in the reactor facility for different purpose, it is very important to check for any radiation leakage or harmful points for the safety of personnel. The measured background gamma and beta dose rates range at different selected points of the facility was found at about 0.11-0.31 µSv/hr and 0.25- 0.90 µSv/hr respectively, except the surface of ion exchange resin column. At the surface of ion exchange resin column, the measured background dose rate for gamma and beta radiation was about 18.04 µSv/hr and 14.2 µSv/hr, respectively. No neutron dose rate was found at any point at the background condition. The gamma, beta and neutron dose rate at different reactor power levels at the selected points was measured and the change of radiation dose rate level was observed. Fig. 3 shows the gamma, neutron and beta dose rate at reactor power level 0 MW (background) at the selected points of the facility.

Fig. 3: Radiation Monitoring and Measurement devices uses for the study (a. Mirion RDS-31, b. VF PNM-01, c. RDS-31 along with GMP-25 Probe).

Fig. 4 shows the investigated dose rates at different strategic points around the reactor facility at 0.0 MW (background level). During background measurement time, there were no remarkable dose were found around the reactor facility except gamma radiation. The highest gamma dose rate was found at the surface of the Ion-exchange resin column. This is because of the resin filter and the dose rate was found at 18.04 μSv/h. The second highest dose rate was found at shielding surface of the Ion-exchange resin column at 0.31 μSv/h as well as the lowest dose rate was found at 0.10 at Reactor Operator (RO) Sitting area.  However, at shutdown condition, all the areas are in less concern except at the surface of the Ion-exchange resin column.

Fig. 4: Dose rates at selected points of facility at reactor power level 0 MW.

Fig. 5 shows the investigated dose rates at different strategic points around the reactor facility at 0.5 MW. After background measurement the reactor is operated at 500 KW. Radiation level is measured at different strategic points and recorded. The gamma, beta, and neutron radiation are plotted against different measurement points and shown in Fig. 4. Among the 3 types of radiation measurements, the gamma radiation shows the highest value whereas the neutron radiation shows the lowest. It is observed that the gamma radiation level was highest (16 μSv/h) at the point of radial beam Port 1 and the lowest radiation level (0.1 μSv/h) was found at Reactor operators sitting area as well as outside of reactor hall. 

Fig. 5: Dose rates at selected points of facility at reactor power level 0.5 MW.

Fig. 6 shows the investigated dose rates at different strategic points around the reactor facility at 1.0 MW. Completing the survey measurement at 500 KW, the reactor power is increased to 1 megawatt and necessary measurements are done very carefully. All the measurements are recorded and plotted the results as shown in above figure. It is found that the gamma radiation level was highest (109 μSv/h) at the point of radial beam Port 1.

Fig. 6: Dose rates at selected points of facility at reactor power level 1 MW.

Fig. 7 shows the investigated dose rates at different strategic points around the reactor facility at 1.5 MW.  When the reactor is operated at 1.5 MW, the radiation level increases than the previous measurements. The major gamma radiation levels at Radial beam port 1, Grating surface of the reactor top and Surface of the Ion-exchange resin column were 146, 119 and 101 in μSv/h, respectively. Among them the highest radiation level was at Radial beam port 1.

Fig. 7: Dose rates at selected points of facility at reactor power level 1.5 MW.

The radiation level increases than the above measurements when the reactor is operated at 2 MW. The major gamma radiation levels at Radial beam port 1, Piercing beam port, Grating surface of the reactor top and Surface of the Ion-exchange resin column were 187, 140, 106 and 72 in μSv/h, respectively. The highest radiation level was measured at Radial beam port 1.  

Fig. 8: Dose rates at selected points of facility at reactor power level 2 MW.

Fig. 9 shows that maximum dose rate was found at near radial beam port-I (on leakage surface). The maximum dose rate caused by gamma, neutron and beta radiation was found at Radial Beam port-I (on leakage surface). Radial Beam port-I (on leakage surface) and piercing beam port are the points where very high dose rate was found for gamma, neutron and beta radiation. So, we investigate detailed for the above-mentioned Radial Beam port-I (on leakage surface) and piercing beam port Points. The maximum dose rates (gamma, neutron, beta) for Radial Beam port-I (on leakage surface), Radial Beam Port-I (on shielding surface) and piercing beam port are (207.00, 18.17, 232.00) and (113, 17.30, 178.00) respectively. At Radial Beam port-I point at shielding surface the dose rate level was lower, the point Radial Beam port-I (on leakage surface) is of less concern. Although the dose rate was very high at Radial beam port-I (on leakage surface) but it was lower at Radial beam port-I (on shielding surface).

Fig. 9: Dose rates at selected points of facility at reactor power level 2.4 MW.

By providing necessary shielding at the both beam ports, this high radiation level can be minimized and the area will be suitable for working. 

Fig. 10: Dose rate at radial beam port-1 (on leakage surface) at different reactor power level.

Due to the leakage at these two points, the dose rates at other beam ports were measured in detail.

This study deals with the gamma, neutron and beta dose rate at Radial beam port-1 and Piercing beam port positions at different power level which is shown graphically in fig-10 and 11which shows that radiation dose rate increases with power level respectively. The beta radiation increased with power level. The ration of change is not constant following changes in the reactor power level. As the gamma rays are from the decay of fission product with half-lives ranging from less than a second to several years. 

Fig. 11: Dose rate at piercing beam port at different reactor power level.

At main entry (on ground floor), neutron dose rates were found higher than beta and gamma dose rates. This could be due to the leakage at radial beam port-I, or from any point between tangential beam port to radial beam port-I. More study will be done on this to investigate the real point of neutron leakage. The neutron dose rate was 4.43 μSv/hr for 2.4MW reactor power level at main entry ground. It is a matter of concern for working personnel around the facility.

Fig. 12: Dose rate at reactor pool top (Rail Surface) from low power to high power level for NCCM (0 to 500 kW) to FCCM (1000 to 2400 kW).

The whole experimental data were taken in the presence of a radiation control officer. The activities were performed according to NSRC rules and regulations. The idea about radiation is safety first. This work will provide vital information for the individuals working in the BTRR reactor facility. The important thing about this study is that it was found background level radiation dose rate outside the reactor facility. So, there is no leakage for radiation to affect outside environment.

 Fig. 13: Dose rate at reactor pool top (Grating Surface) from low power to high power level for NCCM (0 to 500 kW) to FCCM (1000 to 2400 kW).

This present study is the reflection of a routine study of the reactor facility radiation dose rate to check how the radiation dose rates are present at different key points. The maximum dose rates were found at Radial Beam Port-I (on leakage surface). The piercing beam port dose rate is a matter of concern and immediate action must be taken to ensure the safety of individuals of reactor facility. The maximum dose rates (gamma, neutron, beta) for Radial Beam port-I (on leakage surface), Radial Beam Port-I (on shielding surface) and piercing beam port are (207.00, 18.17, 232.00) and (113, 17.30, 178.00) respectively. The increasing rate of gamma and beta dose rate is a matter to look into. Special care and action must be taken to avoid any accidental situation due to leakage. More protective shielding must be taken at radial beam port-I (on leakage surface). Protective and effective shielding is must now at piercing beam port for the safety of individuals working in the facility. The present study is another step towards keeping the reactor facility safe for everyone working and doing research work utilizing the facility.

A research reactor facility is very important facility for different nuclear research related works as well as different experiments and other purposes. It cannot be avoided having a research reactor at the time of nuclear era. But it is important to ensure the quality control of the facility. This study is a motivation towards avoiding excessive radiation doses and assessing the present condition of shielding around the reactor facility for all around the reactor facility. From this study, higher dose rates were found at radial beam port-I (on leakage surface) and piercing beam port. The maximum dose rates (gamma, neutron, beta) for Radial Beam port-I (on leakage surface), Radial Beam Port-I (on shielding surface) and piercing beam port are (207.00, 18.17, 232.00) and (113, 17.30, 178.00) respectively. Neutron dose was found at main entry door during the operation of reactor.  The authority should design appropriate shielding to prevent leakage at piercing beam port. Additional shielding must be added for safety at radial beam port-I to shield the doses. The cause of neutron does at main entry door need to be investigated to avoid unwanted circumstances. These study outcomes will be effective for the prevention of any radiological damagers to the individual related to reactor facility. This kind of experimental study must be performed on a scheduled basis for the betterment of research and other works around any reactor facility.

The authors are very much thankful to honorable Director General, Atomic Energy Research Establishment for his constant inspiration to solve radiological issues of the reactor. The authors are also very thankful to the BTRR facility staff for their cooperation in the experimental works.

The authors sincerely admitted no conflicts of interest to declare.