Compact shielding design of a portable 241Am–Be source

Author’s Accepted Manuscript Compact Shielding Design of a Portable 241Am–Be Source Feng Zhang, He Wu, Xinguang Wang, Guoli Wu, Wenbao Jia, Yongzhou Ti www.elsevier.com/locate/apradiso

PII: DOI: Reference:

S0969-8043(16)31088-0 http://dx.doi.org/10.1016/j.apradiso.2017.06.033 ARI7937

To appear in: Applied Radiation and Isotopes Received date: 15 December 2016 Revised date: 7 June 2017 Accepted date: 22 June 2017 Cite this article as: Feng Zhang, He Wu, Xinguang Wang, Guoli Wu, Wenbao Jia and Yongzhou Ti, Compact Shielding Design of a Portable 241Am–Be Source, Applied Radiation and Isotopes, http://dx.doi.org/10.1016/j.apradiso.2017.06.033 This is a PDF file of an unedited manuscript that has been accepted for publication. As a service to our customers we are providing this early version of the manuscript. The manuscript will undergo copyediting, typesetting, and review of the resulting galley proof before it is published in its final citable form. Please note that during the production process errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain.

Compact Shielding Design of a Portable 241Am–Be Source Feng Zhang1*, He Wu1, Xinguang Wang1*, Guoli Wu2, Wenbao Jia3, Yongzhou Ti4 1

School of Geosciences, China University of Petroleum (East China), Qingdao 266580, China

2

School of Science, China University of Petroleum (East China), Qingdao 266580, China

3

College of Materials Science and Engineering, Nanjing University of Aeronautics and

Astronautics, Nanjing 211106, China 4

Zhengzhou Key Laboratory of Isotope Tracing and Detecting, Isotope Research Institute of

Henan Academy of Sciences Co. Ltd, Zhengzhou 450015, China

Abstract A compact shielding tank for portable 241Am–Be neutron source, which is used in nuclear well logging, was designed according to the Monte Carlo simulation. From inner to outer, the proposed tank has two shielding layers to shield the high- and low-energy neutrons. In this study, the shielding properties of several materials were evaluated. Tungsten was selected as the neutron moderator to build the inner layer. The thermal neutron absorber in the outer layer was made of polyethylene containing 1.2% boron carbide. The volume and weight of the new tank reduced by 86% and 54%, respectively, by using this dual-layer shielding model, when compared with those of the old shielding tank. Moreover, the simulation results indicated that the total dose rate anywhere outside the tank is less than 0.025 mSv/h and that the intensity of gamma flux at the tank surface becomes lower. Keywords: Nuclear well logging;

241

Am–Be source; dose rate; composite material

shielding; Monte Carlo simulation

Introduction 241

Am–Be neutron sources in nuclear well logging tools are commonly used to measure the formation porosity and element content and for other applications. Because of the growing requirement of the oil logging industry and according to the Health, Safety and Environment management, an inevitable strict shielding and protection system of neutron sources has to be established. Several previous works have provided useful references for designing and analyzing neutron generator radiation shielding scheme though the Monte Carlo simulation method. In addition, some research focused on the nuclear physics laboratory shielding project, in particular, the radiation dose produced by the DT or DD neutron generators and spallation neutron source facilities [1-6]. It was revealed that the shielding effectiveness of neutron shielding material depends on the neutron energy. At present, paraffin or other low-density, hydrogen-rich materials are generally selected as a single shielding material for the traditional neutron shielding tank, which requires a volume and weight of the old tank to be 1.36 m3 and 1060 kg, respectively [7]. In this work, the authors aimed to design a new compact shielding tank for 1

241

Am–Be neutron sources, which are used in the well logging field and whose intensities are as high as 18 Ci, that are highly mobile and can be frequently transported across the world [8]. The shielding schemes for neutron sources are based on the moderation and absorption of the sources during transport. Therefore, the new shielding tank was divided into two layers to moderate high-energy neutrons and absorb thermal neutrons. The shielding properties of various materials were evaluated according to the Monte Carlo simulation to design a safe, compact, and efficient neutron source-shielding tank. In this simulation study, 5×107 source neutrons were assessed, and the statistical error was less than 0.5% in all calculations.

Design Calculations of Shielding Tanks a) Dose calculations from old shielding tank 18 Ci 241Am–Be source with the yields of 4.14×107 n/s neutrons that is used in the well logging tool. The average neutron energy released from the source is approximately 4.5 MeV, and the maximum energy is approximately 11 MeV [9-10]. The energy spectrum of the neutron source is shown in Fig. 1. The design of the neutron source-shielding tank focuses on the following three major aspects: – To reduce the size of the tank as much as possible – The energy distribution and intensity of the induced gamma rays should be lesser than those in the old shielding tank – According to the standards of the IBSS and GB 18871-2002 and the average working time of professional radioactive workers, the radiation dose rate at any point of the tank surface should be less than 0.025 mSv/h [11-12]. Relative Probability of Emission

0.05

0.04

0.03

0.02

0.01

0.00 0

2

4

6

8

10

12

Energy /MeV

Fig. 1 Neutron energy spectrum of 241Am–Be neutron source

First, to optimally design the shielding tank, radiation dose distribution of the conventional shielding tank was investigated by using a computational model based on the Monte Carlo N-Particle Transport (MCNP) Code [13], as shown in Fig. 2.

2

Fig. 2 MC model of the conventional shielding tank

In the MCNP model of the old shielding tank, a 241Am–Be neutron source is located in the center of the tank. The neutron source energy spectrum (Fig. 1) was input to the MCNP simulation model. Paraffin was used as the filler material in the shielding tank because of its relatively modest cost and well-known neutron shielding ability. The parameters, materials, and components of the conventional shielding tank are shown in Fig. 2. The selected counting meshes were set as 1.5 cm×1.5 cm, and the recommended neutron and gamma flux-to-dose conversion factors used in this study are listed in Tables 1 and 2, respectively [14-15]. The distribution of neutron field and the two-dimensional maps of gamma rays, neutrons, and total dose rate in the old shielding tank were obtained from MCNP5 simulations, as presented in Fig. 3 and Fig. 4, respectively. Energy (MeV) 2.5E–08 1.0E–07 1.0E–06 1.0E–05 1.0E–04 1.0E–03 1.0E–02 1.0E–01

Table 1 Neutron flux-to-dose conversion factors Dose Energy Dose (mSv/h)/(n/cm2-s) (MeV) (mSv/h)/(n/cm2-s) 3.67E–05 5.0E–01 9.26E–04 3.67E–05 1 1.32E–03 4.46E–05 2 1.43E–03 4.54E–05 2.5 1.25E–03 4.18E–05 5 1.56E–03 3.76E–05 7 1.47E–03 3.56E–05 10 1.47E–03 2.17E–04 14 2.08E–03

Energy (MeV) 0.01 0.015 0.02 0.03 0.04 0.05 0.06 0.08 0.1 0.15 0.2

Table 2 Gamma flux-to-dose conversion factors Dose Energy Dose (mSv/h)/(n/cm2-s) (MeV) (mSv/h)/(n/cm2-s) 2.78E–06 0.5 9.09E–07 1.11E–06 0.6 1.14E–06 5.88E–07 0.8 1.47E–06 2.56E–07 1 1.79E–06 1.56E–07 1.5 2.44E–06 1.20E–07 2 3.03E–06 1.11E–07 3 4.00E–06 1.20E–07 4 4.76E–06 1.47E–07 5 5.56E–06 2.38E–07 6 6.25E–06 3.45E–07 8 7.69E–06 3

0.3 0.4

5.56E–07 7.69E–07

10

9.09E–06

The total neutron cross-section of a material is related to the neutron energy and the type of material. Fig 3 illustrates the distribution of three energy groups of neutrons: thermal neutrons, 0.025 eV – 2 MeV group, and 2 – 12 MeV group. It is apparent that their distributions in shielding tank are quite different and that they require different shielding materials for the optimization of study design.

Fig. 3 Distribution of neutrons in conventional shielding tank

Fig. 4 displays the contribution of gamma rays and neutrons to the total dose rate. Clearly, the dose rate induced from gamma rays was considerable because of the low shielding effectiveness of the neutrons in the tank. Even though the total dose rate anywhere outside the tank was less than 0.025 mSv/h, the volume and weight of the conventional shielding tank were substantially too large when paraffin was used as a single shielding material. 60

60

mSv/h

40

40

9.119E-04

1.000E-04

40

9.119E-04

0.006738

0.006738

20

0

0.3679

2.718

-20

1097

-60 -60

-40

-20

0

20

X /cm

40

60

2040

0.3679 2.718

-20

20.09 148.4

0

2.718

-20

-40

0.04979

Y /cm

0.04979

Y /cm

0.3679

9.119E-04

0.006738

20 0.04979

0

mSv/h

1.000E-04

20

Y /cm

60

mSv/h

1.000E-04

20.09

-40

148.4 1097

-60 -60

-40

-20

0

20

X /cm

40

60

2040

20.09

-40

148.4 1097

-60 -60

-40

-20

0

20

40

60

2040

X /cm

(a) Gamma dose rate (b) Neutron dose rate (c) Total dose rate Fig. 4 Gamma, neutron, and total dose rate maps of conventional shielding tank

b) Design calculations of the new shielding tank Moderator material According to the physical arguments [16], composite nuclei could be formed by the inelastic collisions before the energy of neutrons drops to 12 MeV, and then inelastic gamma rays are released and high-energy fast neutrons are moderated. With the gradual decrease in energy of neutrons of 1 – 2 MeV, elastic scattering played a dominant role in slowing down the neutrons. Finally, capture gamma rays were released during the process of radiation capture when the neutron energy was declined to the thermal neutron level. Therefore, the main work of 241Am–Be neutron source shielding scheme is to moderate the high-energy neutrons (energy higher than 2 MeV in particular) because the thermalized neutrons and thermal neutrons can be easily absorbed by boron-enriched organic matters. By using such a scheme, Monte 4

Carlo-based simulation studies were conducted to optimize the moderator lining and the outer thermal neutron absorbing material. The calculation model is presented in Fig. 5.

(a) side view (b) cross-sectional view Fig. 5 MCNP model of the new dual-layer filling method

Fig. 5 is the MCNP model of the new shielding tank; particularly, the lining material in green represents the moderator, and the red area represents the thermal neutron absorber. M and L in Fig. 5a represent the thickness of the absorber and moderator materials, respectively. For the purposes of improving the shielding effectiveness and reducing the tank size by using this dual-layer shielding model, as a first stage of the study, we moderated the high-energy fast neutrons as much as possible by selecting proper core material of the shielding tank. The inelastic cross-sections of copper, iron, lead, tungsten, and other common heavy metals are higher than that of paraffin, for which the high-energy fast neutrons are more concentrated near the tank center. In the new design, inelastic scattering played a major role in slowing down the high-energy fast neutrons [17]. Therefore, to select the best suitable moderating material for the high-energy fast neutrons having energy higher than 2 MeV, the MCNP model, which is shown in Fig. 5, was established. Copper, iron, lead, tungsten, graphite, polyethylene (PE), paraffin, and polyvinyl chloride (PVC) were used as moderator materials, M was set to 5, 10, 15, 20, 25, 30, 35, 40, 45, 50, 55, 60, and 65 cm, and the outer thermal neutron absorber was absent. The flux of neutrons with energy higher than 2 MeV outside different moderator materials was calculated, and the relationship between the relative neutron flux and M is illustrated in Fig. 6.

5

Relative Neutron Flux (energy 2~12MeV)

1E-3 1E-4 1E-5 1E-6 1E-7 1E-8 1E-9 1E-10

Graphite Copper Lead Tungsten

1E-11

Paraffin PE PVC Iron

1E-12 0

10

20

30

40

50

60

70

M /cm

Fig. 6 Relationship between the relative neutron flux and M

As it can be seen from Fig. 6, the neutron flux sharply declined with the increase of moderator thickness. However, the neutron flux outside lead, graphite, PE, paraffin, and PVC were higher than that outside copper, iron, and tungsten. Among these shielding materials, tungsten moderated most of the high-energy fast neutrons that were emitted from 241Am–Be neutron source, and the corresponding neutron flux is the lowest. Therefore, tungsten was selected as the moderator material to structure the lining of the new shielding tank. Neutron absorbing material Tungsten was selected as the moderator lining for the reasons above. The energy distribution of neutrons that interacted with tungsten moderator was investigated when the M value was set at 5, 15, and 20 cm as shown in Fig. 7. The simulation results revealed that the percentage of high-energy neutrons (energy higher than 2 MeV) in energy spectrum was significantly reduced, when compared with the initial neutron source energy spectrum (see Fig. 1). Moreover, the energy of most of neutrons outside tungsten moderator was less than 2 MeV, and these neutrons can be easily slowed down by elastic scattering and capture interaction. 1E-4

Relative Neutron Flux

1E-5

1E-6

1E-7

1E-8

1E-9

1E-10

1E-11 0.01

M 5cm 15cm 20cm 0.1

1

10

Energy /MeV

Fig. 7 Neutron energy spectrum after interaction with tungsten moderator

For selecting the proper material to absorb neutrons with energy less than 2 MeV, a similar MCNP model as shown in Fig. 5 was used. In this model, a 15 cm thick tungsten moderator was used. The outer absorber layer (red colour in Fig. 5) was 6

chosen to be graphite, PE, paraffin, PVC, and water. L was set to 5, 10, 15, 20, 25, 30, 35, 40, 45, 50, 55, 60, and 65. The relationship between L and the total neutron flux outside the absorber layer of various absorbing materials was simulated, and the results are shown in Fig. 8. 1E-3

Relative Total Neutron Flux

1E-4 1E-5 1E-6 1E-7 1E-8 Graphite Water PE PVC Paraffin

1E-9 1E-10 1E-11 0

10

20

30

40

50

60

70

Thickness /cm

Fig. 8 Relationship between the total neutron flux and L

The simulation results showed that the total neutron flux decreases with increase in the thickness of the absorbing material. They indicate that the neutron absorption capability of the absorbing materials is in the following order: PE>paraffin>water>PVC>graphite. Therefore, PE was selected as the neutron absorbing material of the outer layer. Boron carbide content in absorbing material 10

B has a high capture cross-section of thermal neutrons, thereby making natural boron to be an attractive material for shielding thermal neutrons [18]. The MCNP model as shown in Fig. 5 was used to determine the optimal boron content in PE absorber. In particular, the lining of the tank was made of PE with 10, 15, 20, and 25 cm thicknesses, and the outer material was absent. This section purposes to determine boron carbide content in PE absorber that is expected to absorb low-energy neutrons. According to the neutron energy spectrum after interaction with tungsten moderator (see Fig. 7), the input neutron energy was set not higher than 2 MeV in the simulation model. The weight percentage content of boron carbide in PE was set to 0%, 0.5%, 1.0%, 1.5%, 2.0%, 2.5%, 3.0%, and 3.5%. For comparison, the relationship between normalized neutron flux and different natural boron carbide content in PE with different PE thicknesses is shown in Fig. 9.

7

1.0

Simulated, 10cm Simulated, 15cm Simulated, 20cm Simulated, 25cm Fitting, 10cm Fitting, 15cm Fitting, 20cm Fitting, 25cm

Normalized Neutron Flux

0.8

0.6

0.4

0.2

(Selected point) 0.0 0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

The Mass Percentage of Boron Carbide /％

Fig. 9 Relationship between the neutron flux and boron carbide content

As expected, the total neutron flux sharply decreased with the increase in boron carbide content, as shown in Fig. 9. However, it changed gradually when the weight percentage content of boron carbide in PE exceeded 1.2%. For this reason, a boron carbide content of 1.2% in PE was determined to be the optimal value.

Results and Discussion a) Parameters of the new shielding tank Tungsten and PE containing 1.2% boron carbide (B-PE) were selected as the moderator and absorber, respectively, for the new shielding tank according to the Monte Carlo simulation. Additionally, to determine the optimal specifications of the inner tungsten and outer B-PE, by using the calculation model shown in Fig. 5, the thickness of tungsten (M) was set to 5, 10, 15, 20, and 25 cm, and the thickness of B-PE (L) was set to 5, 10, 15, 20, 25, 30, 35, 40 45, 50, 55, and 60 cm. The total dose rate at the tank surface for every combination of M and L shown in Fig. 5 was calculated, and the results obtained from the simulations are summarized in Fig. 10. 60

50

Total dose≤0.025 mSv/h

L /cm

40

0.0

25

30

mS v/h

Bo u

nd

ary

(Selected point)

20

10

Total dose＞0.025 mSv/h

0 0

5

10

15

20

25

M /cm

Fig. 10 Dose rate of various combinations of M and L

Obviously, the optimized thicknesses of tungsten and B-PE should be determined by the 0.025 mSv/h dose rate boundary as shown in Fig. 10. In the present model, the volume, weight, and cost were considered; finally, 13-cm-thick tungsten and 18-cm-thick B-PE were chosen as the moderator and absorbent, respectively. For the 8

new model, the optimized specific parameter values and the maximum surface dose rate of the 241Am–Be neutron source shielding tank were found and compared with those of the old design, as shown in Table 3. Table 3 Parameters of the old and new shielding tank Parameter

Moderator

Absorbent

B4C content in absorbent

Volume (m3)

Weight (kg)

Dose rate (mSv/h)

Old design

Paraffin

Paraffin

0

1.36

1060

0.012

New design

Tungsten

B-PE

1.2%

0.194

485

0.021

b) Radiation dose calculation In this section, gamma, neutron, and total radiation dose maps of the new shielding tank were separately evaluated. The simulation model was established as shown in Fig. 5, and the optimum parameters that are shown in Table 3 were input. Calculated results are shown in Fig. 11. Obviously, it should be noted that the dose rate induced from gamma rays (Fig. 11a) is very limited when comparing with that induced from neutron dose rate (Fig. 11b), and the total dose rate (Fig. 11c) anywhere outside the new shielding tank was less than 0.025 mSv/h. mSv/h

mSv/h

1.000E-04

30

Tank Border

Tank Border

20

Tank Border

20

20 0.006738

10

2.718

-10

0

0.3679 2.718

-10

20.09

-20

-30 -10

0

X /cm

10

20

30

2040

2.718 20.09

148.4

148.4

-30

1097

-20

0.3679

-10

-20

148.4

-30

0

20.09

-20

-30

0.04979

Y /cm

0.04979

Y /cm

0.04979 0.3679

9.119E-04

0.006738

10

0

1.000E-04

30

9.119E-04

0.006738

10

Y /cm

mSv/h

1.000E-04

30

9.119E-04

1097

-30

-20

-10

0

10

20

30

X /cm

2040

1097

-30

-20

-10

0

10

20

30

2040

X /cm

(a) Gamma dose rate (b) Neutron dose rate (c) Total dose rate Fig. 11 Gamma, neutron, and total dose rate maps of the new shielding tank

c) Induced gamma rays outside the new shielding tank The number of induced gamma rays outside the new shielding tank was calculated and compared with that of the old tank, as shown in Fig. 12. In the new neutron source shielding tank, gamma rays could be from 241Am–Be source itself (can be neglected) and from neutron interactions such as 10B (n, γ) 11B, 56Fe (n, γ) 57Fe, 184W (n, n′ γ) 184W, 1H (n, γ) 2H, and 12C (n, n′ γ) 12C. However, Fig. 12 shows that the intensity of surface gamma rays becomes lower because most of the high-energy neutrons are efficiently moderated (Fig. 7); in addition, the core matter of tungsten has a strong attenuating capacity for gamma rays.

9

1E-5

Old design New design

Relative Gamma Flux

1E-6

1E-7

1E-8

1E-9

1E-10 0

1

2

3

4

5

Energy /MeV

Fig. 12 Comparison of the surface gamma rays

Conclusion (1) A Monte Carlo study was conducted to optimally design a compact 241Am–Be neutron source shielding tank. Gamma, neutron, and total radiation dose maps of the new and old shielding tank were separately evaluated. (2) The dual-layer shielding model improved the shielding effectiveness of the new source tank. Tungsten with a thickness of 13 cm was selected as the inner moderator material, and 18-cm-thick PE containing 1.2% boron carbide was used to construct the outer neutron absorber of the new shielding tank. (3) The volume and weight of the new tank were as low as 0.194 m3 and 485 kg, respectively. In addition, the total dose rate outside the new shielding tank was less than 0.025 mSv/h, and the surface intensity of gamma rays became lower, which satisfactorily meet the established principles and the relevant standards.

Acknowledgments The authors would like to acknowledge the assistance of the National Natural Science Foundation of China (41574119 and 41504099), the National Major Oil and Gas Special Fund of China (2017ZX05019005-004), the Natural Science Foundation of Shandong Province (ZR2015DQ003), and the constructive contribution of the anonymous reviewers and language editor.

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[6] Nemati M J, Habibi M, Amrollahi R (2013) Analysis of concrete labyrinth shielding and radiation dose for APF plasma focus neutron source by FLUKA Monte Carlo code. J Radioanal Nucl Chem 295(1):221-226 [7] Wraight P D, Robinson E, de Fleurieu R (1991) Opportunities to reduce risks associated with nuclear logging techniques. SPE Health, Safety and Environment in Oil and Gas Exploration and Production Conference. Society of Petroleum Engineers [8] Badruzzaman A, Barnes S, et al. (2009) Radioactive sources in petroleum industry: applications, concerns and alternatives. Asia Pacific Health, Safety, Security and Environment Conference. Society of Petroleum Engineers [9] Abdelmonem M, Naqvi A, Al-Ghamdi H, et al (2007) Performance comparison of 2.8 MeV and 241Am-Be neutrons based moisture measurement setups. J Radioanal Nucl Chem 274(1):131-137 [10] Kluge H, Weise K (1982) The neutron energy spectrum of a 241Am-Be (Alpha, n) source and resulting mean fluence to dose equivalent conversion factors. Radiat Prot Dosim 2(2):85-93 [11] World Health Organization (1996) International basic safety standards for protecting against ionizing radiation and for the safety of radiation sources. [12] Ziqiang P, Changqing Q, et al (2003) GB 18871-2002 Basic standards for the safety of ionizing radiation protection and radiation sources. [13] X-5 Monte Carlo team (2003) MCNP-a general Monte Carlo N-particle transport code, version 5”, LA-CP-03-0245, LANL. [14] ANS-6.1.1 Working Group, M. E. Battat (1977) American National Standard Neutron and Gamma-Ray Flux-to-Dose Rate Factors, ANSI/ANS-6.1.1-1977 (N666), American Nuclear Society, LaGrange Park, Illinois [15] ICRP Committee 3 Task Group, P. Grande and M. C. O’Riordan, chairmen (1971) “Data for Protection Against Ionizing Radiation from External Sources: Supplement to ICRP Publication 15,” ICRP-21, International Commission on Radiological Protection, Pergamon Press [16] Kane G L. Modern (1993) elementary particle physics: the fundamental particles and forces? Westview Press. [17] Karimi-Shahri K, Rafat-Motavalli L, Miri-Hakimabad H (2013) Finding a suitable shield for mixed neutron and photon fields based on an Am–Be source. J Radioanal Nucl Chem 298(1):33-39 [18] Korkut T, Karabulut A, et al (2010) Investigation of fast neutron shielding characteristics depending on boron percentages of MgB2, NaBH4 and KBH4. J Radioanal Nucl Chem 286(1):61-65 Highlights 

Compact shielding design of a portable 18 Ci 241Am–Be source.



Tungsten and polyethylene containing 1.2% boron carbide were selected as the composite shielding material.



We achieved 86% and 54% reduction in the volume and weight of the shielding tank, respectively.



A reduced dose rate less than 0.025 mSv/h was observed outside the shielding tank.

11