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ISSN : 1738-1894(Print)
ISSN : 2288-5471(Online)
Journal of Nuclear Fuel Cycle and Waste Technology Vol.21 No.4 pp.481-487

Development of a Methodology for Estimating Radioactivity Concentration of NORM Scale in Scrap Pipes Based on MCNP Simulation

Wanook Ji1, Yoomi Choi1, Zu-Hee Woo2, Young-Yong Ji1*
1Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon 34057, Republic of Korea
2Korea Institute of Nuclear Safety, 62, Gwahak-ro, Yuseong-gu, Daejeon 34142, Republic of Korea
* Corresponding Author. Young-Yong Ji, Korea Atomic Energy Research Institute, E-mail:, Tel: +82-42-868-4958

November 16, 2023 ; November 30, 2023 ; December 18, 2023


Concerning the apprehensions about naturally occurring radioactive materials (NORM) residues, the International Atomic Energy Agency (IAEA) and its member nations have acknowledged the imperative to ensure the radiation safety of NORM industries. Residues with elevated radioactivity concentrations are predominantly produced during NORM processing, in the form of scale and sludge, referred to as technically enhanced NORM (TENORM). Substantial quantities of TENORM residues have been released externally due to the dismantling of NORM processing factories. These residues become concentrated and fixed in scale inside scrap pipes. To assess the radioactivity of scales in pipes of various shapes, a Monte Carlo simulation was employed to determine dose rates corresponding to the action level in TENORM regulations for different pipe diameters and thicknesses. Onsite gamma spectrometry was conducted on a scrap iron pipe from the titanium dioxide manufacturing factory. The measured dose rate on the pipe enabled the estimation of NORM concentration in the pipe scale onsite. The derived action level in dose rate can be applied in the NORM regulation procedure for on-site judgments.


    1. Introduction

    Some raw materials contain naturally occurring radioactive materials (NORM) and they are processed to manufacture industrial material. The 238U and 232Th decay series and 40K are the main interest in the scope of NORM [1-6]. While concentration of the natural radionuclides in raw materials is usually low, but the concentration can be enhanced during manufacture processing [6-8] in some by-products. It is called TENORM (technically enhanced NORM) and the elevated radioactivity induces the increase of radiation dose. To protect people from the exposure, Korean government enforced the ‘The Act on Protective Action Guidelines against Radiation in the Natural Environment’ in 2012 and TENORM are under the regulation boundary to protect people from radiation exposure. One of the main concerns of the TENORM exposure is due to scrap iron pipes discharged in large quantities from the TENORM industries. Due to the diverse sizes of the scrap pipes, it is not easy to measure the radioactivity of the scales (Fig. 1). Titanium dioxide (TiO2) processing factory is one of the NORM related industries in Korea. The factory treats titanium minerals and various feedstocks containing natural radionuclides. While the radioactivity in the raw materials is usually very low in the range of few Bq per gram, scales in the scrap pipes have elevated radioactivity during manufacturing processes [6]. In this study, we conducted in situ gamma spectrometry to the discharged scrap iron pipes from the factory. Action level (AL) is derived the in terms of dose rate by Monte Carlo simulation to estimate the radioactivity of the pipe scale. The derived AL can be applied to the inspection and assessment of discharged scrap iron from the NORM industries.

    Fig. 1

    The scales in the discharged scrap iron pipes during maintenance period of the TiO2 processing factory.


    2. Materials and Methods

    2.1 MCNP Simulation for Pipe Residues

    MCNP is a Monte Carlo simulation code to conduct radiation transport. Before actual measurement, we conducted a simulation to estimate dose rate of the pipe by the detector as shown in Fig. 2. Internal diameter and thickness are considered as two shape parameters of the pipe which are able to affect the dose rate. In each shape, the thickness of scale was varied from 1 to 10 mm with uniform concentration. The length of pipe was fixed at 100 cm. Gamma radiations were included from isotopes such as 214Pb, 214Bi in the 238U decay series and 228Ac, 212Pb, 212Bi, and 208Tl in the 232Th decay series except below less than 1% emission probability. The density of the scale was assumed as 2.5 g·cm−3. The LaBr3(Ce) detector was placed either at the central position in the pipe or at the outside of the pipe with 10 cm away (Fig. 2). When the detector was placed inside, thickness of the pipe was fixed as 10 mm because the effect of the pipe thickness has little effect on the situation. Likewise, the diameter of pipe was determined as 30 cm because pipe diameter can be neglected when detector is positioned at outside. Table 1 shows the MCNP input conditions including detector location and other simulation conditions. At each condition, MCNP simulation was conducted to derive dose rate and energy spectra with F4 and F8 tally.

    Fig. 2

    Simulations of the scale at the inner surface of the pipe. MARKB2 is located in (a) central position, (b) outside of the pipe.

    Table 1

    Input parameters of the MCNP simulation for scale measurements

    Input parameter Content

    Pipe   Material SUS (density: 7.82 g·cm–3)
      Inner diameter 30, 50, 70, 100 cm
      Thickness 1, 2, 4, 6, 8, 100 mm
      length 100 cm

    Scale   Material CaSO4 and MgSO4 (density: 2.5 g·cm–3)
      Thickness   1, 2, 4, 6, 8, 10 mm
      Radionuclides 214Pb, 214Bi, 228Ac, 212Pb, 212Bi, 208Tl

      Detector 2”φ × 2” LaBr3(Ce) scintillator

    2.2 The LaBr3(Ce) Detector System

    The radiation detector system contains 2”φ × 2” LaBr3 (Ce) scintillation detector (51s51_B380, Saint Gobain, France) with a digital signal processing unit (SI Detection Company, Ltd., HAMPack MCA 527, Korea), controller, and GPS. The signal processing unit includes a high voltage supply, preamplifier, amplifier, and multichannel analyzer (MCA) with 1,024 channels. The LaBr3(Ce) scintillation detector has high performance for gamma spectrometry with good energy resolution. But this type of scintillator has intrinsic background radiations from 138La and 227Ac, which exist in natural lanthanum element about 0.09% abundance. We applied a simple algorithm to subtract the intrinsic background radiation from the measured energy spectrum using the LaBr3(Ce) detector [9-11]. The ambient dose rate can be calculated from the measured energy spectrum by applying dose rate-to-spectrum method (DRS method) [12]. The Gfactor (nGy/hr/cps) which is the conversion factor between the count rate and the dose rate was derived from Monte Carlo calculation from the measured energy spectrum [9- 12]. The equation (1) shows the assessment of the field gamma dose rate by subtracting the intrinsic background radiation of the detector.

    X ˙ = n ( E ) G ( E ) d E X ˙ i B K G

    Fig. 3

    The composition of the MARK-B2 detector system.


    ( = dose rate [nGy∙hr−1], iBKG = intrinsic background radiation)

    2.3 Radiation Measurement of the Residual Scale of the Pipe

    In situ radiation measurement was conducted to an iron scrap pipe discharged from the factory. The pipe was transported to a background location to prevent from the other radioactive residues. Fig. 4 shows the in-situ measurement conducted inside the pipeline. To position the detector deeply inside the pipe, a slider was inserted up to a depth of 10 cm, and measurements were taken. Repeating measurements at both the front and rear ends can enhance the accuracy of residual radioactivity concentration evaluations. The internal and external diameters of the pipeline were averaged at approximately 600 mm and 610 mm, respectively through repeated measurements. The length of the pipe was averaged to be around 1,000 mm and the pipe thickness was set at 3 mm, and the thickness of the scale was averaged as 2 mm.

    Fig. 4

    In situ measurement of the scrap pipe with the MARK-B2 detector at (a) the central position and (b) the outside of the pipe.


    3. Results and Discussion

    3.1 MCNP Simulation Results

    Figs. 5 and 6 show the dose rate from the MCNP simulation at each detector location. When the detector is positioned at the central position (Fig. 4(a)), it exhibited a precisely symmetric structure. As the internal diameter is increased, the dose rate exhibits an exponential decrease due to longer source to detector (STD) distance. For example, in the case of the pipe of internal diameter of 30 cm, the distance from the detector surface to the scale is approximately 10 cm. With the internal diameter of 100 cm, the distance from the detector surface to the scale is 45 cm. The increased diameter means the increased source volume, but the STD is more effective parameters in the central position. And the scale gets thicker, the dose rates go higher for increase of source amount. When the detector is positioned at the outside (Fig. 4(b)), as the pipe thickness is increased, the calculated dose rate decreases due to attenuation by the pipe. Considering these factors, to specify the effect of pipe thickness and diameter is necessary to estimate the amount of scale from outside of the pipe.

    Fig. 5

    Calculated dose rate for pipe diameter and sludge thickness in the central position of the pipe.

    Fig. 6

    Calculated dose rates for pipe and sludge thickness in the outside position of the pipe.


    3.2 Deriving Equivalent Dose Rate for Action Level

    The action level is specified at 1 Bq·g−1 which is referred as registration criteria in ‘The Act on Protective Action Guidelines against Radiation in the Natural Environment’. The equivalent dose rate corresponding to concentration of 1 Bq·g−1 was calculated (Table 2). When the inner diameter of the pipe is 30 cm, and the scale thickness is determined as 4 mm, the dose rate corresponding to the action level of 1 Bq·g−1 is 170.6 nGy·h−1. Therefore, if the measured dose rate exceeds the sum of the background dose rate and action level, it is highly likely to exceed the action level. When the detector is located at outside, there is an overall decrease due to attenuation by the pipeline thickness. The simulation result for the scrap pipe measurements serves as a conceptual approach as one of the wide range of sizes and shapes of materials and equipment (M&E). In other words, developing conversion factors for radioactivity concentration for different measurement conditions is necessary. However, deriving conversion factors for such a diverse range of sizes and shapes of M&E involves extensive simulations and has a significant drawback in that it requires repeated efforts whenever there are changes in size and shape. Therefore, it is necessary to create representative geometries for various types of industrial byproducts and to tabulate conversion factors for specific variables. Utilizing these results, it is desirable to determine dose rate corresponding to action levels for the target M&E, using a conservative approach.

    Table 2

    Dose rate corresponding to action level which is referred as 1 Bq·g−1 for pipe scale

    Dose rate corresponding to action level (nGy·hr−1)

    Scale thickness (mm) Central position Outside position

    Pipe diameter (cm) Pipe thickness (mm)

    30 50 70 100 1 2 4 6 8 10

    1 43 37 32 26 22 20.4 17.7 15.62 13.88 12.42
    2 85.6 73.8 64 52 43.4 40.4 35.2 31 27.6 24.8
    4 170.6 147.4 127.6 104 85.4 79.4 69.6 61.4 54.6 49
    6 254 220 191.2 155.8 125.8 117.4 102.8 91 81 72.6
    8 338 294 254 208 165 154 135.4 119.8 106.8 95.8
    10 422 366 318 258 202 189.6 166.8 147.8 131.8 118.2

    3.3 In Situ Measurement of Pipe Scale

    First, the background radiation was measured with the LaBr3(Ce) detector at 1 m height from the ground. The ambient gamma dose rate was measured to be 65.9 ± 5.3 nGy·h−1. Then, the detector was positioned inside the scrap iron pipe to measure dose rate. Fig. 7 compares the results of internal static measurements with external measurements. The average dose rate value at the central position was evaluated to be 3,011 ± 227 nGy·h−1 and it was estimated as approximately 92 Bq·g−1 in the scale.

    Fig. 7

    Gamma spectrum measured at each point.


    4. Conclusion

    To establish a design for radiation assessment for industrial by-products, we conducted an investigation into the current status of industrial by-products and regulatory frameworks. Based on this, we derived dose rate in unit activity in various size of pipe and derived action levels for on-site investigations of M&E for actual occurrences of industrial by-products in the facility. For the applicability assessment, we measured a scrap iron pipe occurred in a domestic TiO2 processing facility. This application enabled the derivation of criteria for determining dose rate standards, which were subsequently applied to actual scrap iron pipes.

    Conflict of Interest

    No potential conflict of interest relevant to this article was reported.




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