Research on dose correction method of vehicle-borne environmental radiation measurement equipment

Vehicle-borne environmental radiation measurement technology measures the type and content of radionuclides on the ground and the air-absorbed dose rates along the road using the gamma spectrometer mounted on a car [4]. The Monte Carlo code FLUKA is used to simulate the ambient dose rate of vehicle-borne environmental radiation measurement equipment. A schematic of the FLUKA model used for simulations is illustrated in Figs. 1 and 2. The 4 L NaI (Tl) crystal is covered by a 0.5-mm-thick MgO reflector, which is surrounded by aluminium with a thickness of 2 mm, and by a layer of 2-mm-thick SiO₂ on the photomultiplier coupling side. Given that NaI (Tl) crystals are susceptible to deliquescence, an aluminium box is used to seal them. The dry MgO powder between the crystal and the aluminium box acts as a reflective layer to reflect the light emitted in the scintillator in all directions to the crystal, and SiO₂ can effectively transmit the light to the photomultiplier tube. The main component of the on-board radiation measurement equipment is a 4 L NaI (Tl) detector mounted on the roof (in horizontal placement) of a Cheetah Land Cruiser LBA6482LQ4. The point source is assumed to emit 1.332 MeV gamma rays, and the distance between the isotropic gamma-ray point source and the 10 cm × 10 cm surface of the NaI (Tl) crystal is 100 cm. The intrinsic detection efficiency of a NaI detector is defined as the ratio of the number of pulses deposited in the NaI (Tl) crystal to the incident radiation source on the surface of the detector. The simulation shows that the intrinsic detection efficiency of NaI (Tl) for 1.332 MeV gamma ray is 65.30%.

Fig. 1

Fig. 2

The car body structure is simplified to improve the efficiency of simulation calculations. Except for the chassis and tires, the car body adopts a box structure, while other scattered and lighter parts are ignored to simulate the real car to the maximum extent. Table 1 shows the detailed information, such as the structure and materials, of each component in the studied equipment.

A nuclear accident produces a complex radiation field composed of hundreds of radionuclides with varying distributions over time [10, 11]. Therefore, using the traditional ground radiation detection method is insufficient. In this study, the radiation field distribution in the radioactive source area of a nuclear accident is simulated using the Monte Carlo program by sampling the radiation rays and energy directions. According to the national standard GB/T17982-2018 [12], nuclear accidents are divided into time periods or phases: the early stage is the initial moment of decay (nuclear accident); the middle stage starts 1 day after the accident; and the later stage starts 100 days after the accident. The cylindrical source model is used to approximate the infinite radiation field after the nuclear accident, as shown in Fig. 3. In the early phase, radioactive substances are released into the atmosphere and dispersed in the air (Fig. 3a). These radionuclides are assumed to be uniformly distributed in the form of the plume. The cylinder has a radius of 1000 cm and a height of 2000 cm. The entire space is filled with air, and the radiation field volume is V = 6280 m³. In the mid-late nuclear phase of nuclear accidents, the radionuclides rapidly deposited in the soil after a nuclear accident, and the early radioactive materials are gradually deposited on the soil to form a radiation field of ground radioactive fallout (Fig. 3b). Similarly, the cylinder has a radius of 1000 cm and a height of 2000 cm. The bottom of the column is filled with a 1-cm-thick layer of soil [13], and the air above this surface is used as an air model. The soil density [14] is 1.7 g/cm³, and the soil mass is M = 4710 kg. Table 2 shows the air and soil content.

Fig. 3

The consequence assessment and environmental radiation monitoring of the Fukushima accident is a complex and arduous task, and the source item is an important basis for accident classification [15, 16]. Numerous researchers of the Fukushima accident assess the total amount of radioactive materials released into the atmosphere to evaluate the environmental impact [16, 17, 18, 19, 20]. However, in daily training, nuclear emergency equipment is used in a radiation field environment formed by a single nuclide. For this reason, this study specifically selects five representative radionuclides released after nuclear accidents (Table 3) as a single-nuclide reference to study the distribution of radiation fields in the early phases of nuclear accidents [16].

The Fukushima accident released various types of characteristic radionuclides. However, simultaneous analyses of the natural radionuclides in the environment are necessary considering the background effect. Table 4 shows the selected nuclides required for the background reference radiation field in the “ 2018 Annual Report of the National Radiation Environment” [21]. The aerosol activity is the air model for the early radiation field and the soil activity is the soil model for the mid-late stage. The complex radiation field distribution composed of hundreds of nuclides is affected by various factors, such as meteorological and geological conditions. The source term distribution of radiation field needs to be obtained from a specific time and location. Thus, this study uses the nuclide data during the Fukushima accident as the source term of the mixed reference radiation field.

Tables 5 and 6 include the main nuclides, radioactive concentration, and activity of the mixed reference radiation field in the early and middle phases of the Fukushima accident, respectively. The radioactivity calculation of the nuclides in the reference radiation fields is shown in Eqs. (1) and (2). (1) A1=C1×V {A_1} = {C_1} \times V (2) A2=C2×M {A_2} = {C_2} \times M where A is the radionuclide activity, C is the radionuclide concentration, 1 and 2 are the different reference radiation fields, V is the volume of the early air reference radiation field, and M is the mass of the mid-late soil reference radiation field.

When different measuring platforms are used to assess the ground deposition of nuclides, all measurement results must be corrected to maintain data consistency from the different material structures and placement height of the measuring instrument [22, 23]. In this study, the detector is installed on the top of the car, parts of the ambient gamma rays are shielded by the body material, and the height attenuation caused by the detection equipment affects the measurement results. Therefore, CF needs to match the dose measured by the vehicle-borne radiation ambient measurement equipment with the air dose. The Monte Carlo simulation is used to study the air absorption dose rate detected by NaI (Tl) detection crystal at two positions inside and outside the vehicle. Thus, the dose CF is established, as shown in Eq. (3). (3) CF=(D˙carD˙air) {\rm{CF}} = \left({{{{{\dot D}_{{\rm{car}}}}} \over {{{\dot D}_{{\rm{air}}}}}}} \right) where CF is the dose CF of the measured dose in the vehicle-borne radiation measurement equipment and and are the air absorption dose rates measured by the detector mounted on the car roof in the car and 1 m above the ground in the air, respectively. Under the same conditions, the simulated detector is placed on the car roof at different dose rates to verify the dose CF. In the end, the air dose rate at 1 m above the ground can be obtained by applying the dose attenuation CF to the vehicle-borne radiation dose measurement, as shown in Eq. (4). (4) D˙1m=D˙carCF {\dot D_{1m}} = {{{{\dot D}_{{\rm{car}}}}} \over {{\rm{CF}}}} where is the ambient dose rate at 1 m above the ground and is the ambient dose rate measured by the vehicle-borne radiation measurement equipment.

The Monte Carlo program is used to simulate the dose rate of ⁶⁰Co, ¹³¹I, ¹³²Te, ¹³³Xe, and ¹³⁷Cs radionuclides with activity of 1 Ci in the early reference radiation field, as shown in Fig. 4. The ambient dose rate increases exponentially with the increase in radionuclide energy in the single-nuclide reference radiation field. In determination radionuclide activity, the dose rate change is only related to nuclide energy.

Fig. 4

Table 7 shows the ambient dose rates at the calibration locations. The rates range from 2.8 × 10⁻⁵ μGy/s to 1.7 × 10⁻² μGy/s as measured at the height of 1 m above the ground, and the rates range from 2.5 × 10⁻⁵ μGy/s to 1.5 × 10⁻² μGy/s as measured by the detector at the car roof. Under the same conditions, the trends of environmental dose rates detected by the NaI (Tl) detector inside and outside the vehicle are roughly the same. The dose rate measured outside is larger than that measured by the vehicle-borne detector because of the shielding effect of the vehicle and its contents.

Tables 5 and 6 show various radionuclides and activity in the mixed-nuclide reference radiation fields. For the early phase of nuclear accidents, this study considers 2 h as the step length. For the mid-late stage, 100 days as a time node is used to calculate 12 different periods (1 day, 3, 5, 7, 9, 11, 13, 20, 40, 60, 80, and 100 days) to obtain the change curve of atomic number and activity of the radioactive decay progeny, as shown in Fig. 5. The radionuclide activity in two different reference fields decreases exponentially over time. The results show that these decreases are not related to the radiation field model but only related to the decay constant.

Fig. 5

Figures 6a and 6b show the trend of changes in the environmental dose rate measured by the NaI (Tl) detector inside and outside of the vehicle, respectively. In this study, the early reference radiation field is the air body source, and the air absorption dose rate is evaluated to only consider the shielding effect of the car body. Thus, the dose rates inside and outside the vehicle slightly differ. By comparison, the mid-late reference radiation field is the soil body source, and the NaI (Tl) detector is placed on the roof, which essentially increases the source range. Thus, the NaI (Tl) detector receives a large difference in the ambient dose rate inside and outside the vehicle. The dose rate changes less and tends to flatten due to the long half-life and slow decay of certain radionuclides, such as ¹³⁷Cs.

Fig. 6

By convention, the ICRP 1996 [9] stipulates measurement of ambient dose rate at 1 m height to ensure consistent results when using different vehicle radiation measurement systems. In this study, only the shielding effect of the car body and the influence of the detector distance from the ground are considered to evaluate the air absorption dose rate. The CF of the vehicle-borne radiation environment measurement equipment under different radiation fields can be obtained using Eq. (3).

The ambient dose rate of the single-nuclide reference radiation field is simulated using the Monte Carlo program. Table 8 shows the results of the dose CF obtained by Eq. (3). The mean dose CF is 0.8786, and the error is within 2%. The results show that energy change negligibly affects the dose CF.

The dose CF of the mixed-nuclide reference radiation field using the vehicle-borne environmental radiation measurement equipment is listed in Tables 9 and 10. The early reference field CF is regarded as a fixed value of 0.8830, and the mid-late mixed-nuclide reference field dose CF fluctuates at approximately 0.6711. The dose rate measured by the vehicle-borne equipment is applied to 1 m above the ground in air using Eq. (4).

The dose rate measurements are repeated at several locations with different ambient dose levels and detector installation locations to verify the validity of the dose CF. Table 11 shows the average results of the detector placed at different positions. The data are divided into dose rate and CF according to Eq. (3).