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Submitted to ‘Chinese Physics C PGNAA source moderation setup optimization

Zhang Jinzhao1（张金钊）Tuo Xianguo1（庹先国） (1.Chengdu University of Technology Applied Nuclear Techniques in Geoscience Key Laboratory of Sichuan Province，Chengdu 610059，China)

Abstract: Monte Carlo simulations were carried out to design a prompt γ-ray analysis (PGNAA) thermal neutron output setup using MCNP5 computer code. In these simulations the moderator materials, reflective materials and structure of the PGNAA 252Cf of thermal neutron output setup were optimized. Results of the calcuations revealed that the thin layer paraffin and the thick layer of heavy moderated effect is best for 252Cf neutrons spectrum. The new design compared with the conventional design, the thermal neutron and rate were increased by 3.02 times and 3.27 times. Results indicate that the use of this design should increase the of prompt gamma-ray neutron activation analysis significantly. Key word: PGNAA; neutron source; thermal neutron; moderation; reflection

1. Introduction study, Monte Carlo calculation was carried out for the Prompt neutron activation analysis design of a 252Cf neutron source moderation setup for the (PGNAA) is a rapid, nondestructive, powerful analysis cement samples[7]. The model of Monte Carlo multielemental analysis technique, large samples of some simulation was verified by experiment[8, 9].We improve minor, trace light elements and is used in industrial the thermal neutron source yield rate of 252Cf neutron by control[1-5]. In a PGNAA analysis, the sample nuclear the PGNAA neutron source structure to the design. The composition is determined from prompt gamma rays calculation results for the new design were compared which produced through neutron inelastic scattering and with the previous, example: themal neutron flux rate, fast thermal . Since the inelastic scattering neutron flux rate, gamma rays yeild, results of these cross section is small, the PGNAA design based on studies are presented. thermal neutron capture. Low- neutrons were 2. The model of Monte Carlo simulations absorbed by the target nucleus forming compound Geometry of the neutrons moderation setup, used in nucleus, which has energy equal to the the study, has two forms to compare. Both of the two and the neutron capture neutron . forms geometry consist of a NaI detector of gamma rays, Compound nucleus emits γ photon to go background a cylindrical polyethylene sample compartment and a state in the 10-14s of time. The gamma yield was neutron. Inside diameter of the cylindrical polyethylene determined by the flux of themal neutron, and sample compartment is 20cm, external diameter is 30cm. determined the analysis results. So performing a PGNAA Height is 20cm. a neutron source of the previous is only setup depends on thermal neutron flux available at the wrapped by a moderator. The new one is different. It has sample, how to improve the thermal neutron produced a moderation and a reflector. Reflector places below the ratio of neutron source has been the research focus of neutron source and the a hemispheric moderator are PGNAA. taken on the upper of the neutron source. Our goal: the Accelerator neutron, source reactor neutron source, neutron is reflected back toward the upper portion of the Am-Be neutron source and 252Cf neutron source can be lower part and the utilization ratio of neutrons is used as PGNAA neutron source[6]. 252Cf neutron source improved, because of the different materials for different which has a high neutron fluxs density and low cost used neutron reflectivity. Fig.1 is a schematic of the previous in PGNAA setup commonly. However, 252Cf source is PGNAA setup source-moderator from paper[7]. Fig.2 is a isotropic neutron source, to sample less thermal neutron. schematic of new design of this study. We determined the Neutron source utilization rate is low. In the present best geometry by calculating.

Supported by National Natural Science Foundation of China(41274109, 41025015) E-mail:[email protected]

Submitted to ‘Chinese Physics C

3.5x10-6

-6 3.0x10 Counts

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-5.0x10-7 0 2 4 6 8 10 12 Energy (MeV) Fig. 1 Schematic of the previous PGNAA setup Fig. 3 252Cf Fission neutron spectrum source-moderator This study refers to the literature[7, 10] of Portland cement composition as a sample for simulation calculation. 2.1 Moderator material Monte Carlo simulation PGNAA analysis accuracy depends on irradiation neutron fluence rate and neutron capture cross section of to be analyzed nuclides . Radiative capture cross section is the main section for the heavy nuclei and low energy neutron action, the entire cross-section is almost equal with the radiative capture cross section, showing 1/ v . Fig. 2 Schematic of the new PGNAA setup So try to increase the sample thermal neutron fluence source-moderator of this study rate and reduce the fast neutron fluence rate in the 252 The simulation uses Cf PGNAA analysis. Therefore, we must moderat neutrons, neutron source which neutron energy spectrum is a to improve the thermal neutron source neutrons in the fission spectrum and can be described watt distribution. proportion, but also pay attention to the material used in Formula: moderation to decrease γ background. We chose five E  materials H O, D O, C H , C, Paraffin to calculation. f( E ) cea sinh bE a 2 2 2 2   The resluts will be analysised, and the best material will E is the neutron energy, a 1.025 , b 1.25, be found out. Fig.4, Fig.5, Fig.6, Fig.7 is a Monte Carlo c  0.365 . As shown of Figure1, it is a 252Cf fission calculation reslut of H2O, D2O, C2H2, C, Paraffin. neutron energy spectrum by MCNP5 simulation.

1.0 1.0 H O 2 D O 0.8 2 nC H 0.8 2 2 C 0.6 H O Paraffin 2 0.6 D O 2 0.4 Counts nC H 2 2 0.4 C Counts C H 0.2 30 60 0.2

0.0 0.0 0 20 40 60 80 Thickness/cm 0 20 40 60 80 Fig. 4 Thermal neutron flux of five moderator materials increase Thickness/cm with the thickness Fig. 5 Fast neutron flux of five moderate materials increase with

Submitted to ‘Chinese Physics C stronger the reflection is. Fig.8, Fig.9 is a Monte Carlo 1.0

calculation result of H2O, D2O, C2H2, C, Be, paraffin 0.8

0.9 0.6 H O 2 D O 2 0.8 0.4 nC H Counts 2 2 C 0.7 nC H C H 2 2 0.2 30 60 0.6 C H 30 64 H O 0.0 0.5 2 Be Counts 0.4 0 20 40 60 80 C 2 Thickness/cm D O the thickness 0.3 Fig. 6 Neutron flux of five moderator materials increase with the 0.2 0.1 thickness 0 10 20 30 40 50 60

1.0 H O Reflector Thickness/cm 2 D O 2 nC H Fig. 8 The reflective neutron flux in different thickness of five 0.8 2 2 C C H 30 60 kinds of reflective material 0.6

0.14 nC H 0.4 2 2 Counts C H 0.12 30 64 H O 0.2 2 0.10 Be

0.0 C 0.08 D O 2 0 20 40 60 80 0.06

Thickness/cm Counts 0.04 Fig. 7 Gamma ray flux of five moderator materials increase with 0.02 the thickness 0.00

As shown of Figure2, Figure3, Figure4, Figure5: -0.02 0 10 20 30 40 50 60 Thermal neutrons: Thermal neutrons increased Reflector Thickness/cm rapidly with increasing light water, paraffin, Fig. 9 The γ ray flux in different material and thickness polyethylene thickness in 9cm-12cm growth to the Fig.8, Fig.9 shows the reflective neutron flux, γ ray maximum. But it with the moderator increasing the in different thickness of five kinds of reflective material. thickness starts to decrease, because a large number of is the reflective effect best material on a thin the thermal neutron are absorbed. When and layer, followed by paraffin. Thick layer of heavy water 48cm, the thermal neutron flux is maximum. and graphite is reflective of the best materials.The result Thermal neutrons loss is less with thickness increase in showed that Thick layer of heavy water and graphite of the moderator. reflection effect was the best. Because it contains Fast neutrons: The five substances fast neutron flux nuclides, light water, polyethylene, paraffin is reduced with the increase in thickness of the will absorb neutrons and γ-rays be generated. Beryllium, moderator substance, sequentially: Paraffin, light water, heavy water and graphite which do not contain neutron polyethylene, heavy water, graphite. absorbing nuclides produces γ-ray flux is extremely low γ-rays: In a processing, paraffin, as the reflective materials. light water, polyethylene will produce gamma rays. 3. Neutron source setup optimization Yields get maximums and reduce subsequent with the Considering various substances above reflection and thickness increase in at a thickness of 22cm. Graphite moderating effect, design a new neutron source. As thin and heavy water is not substantially generated γ-rays. beryllium reflector best, we use a hemispherical wrapped 252 2.2 Reflective material Monte Carlo simulation Cf fission neutron source from the lower part. Thick The reflection is determined by the scattering cross layer of heavy water and graphite having a similar section of nuclide, the absorption cross section, surface reflectivity, and they γ-ray yield lower. But heavy water is density of atoms and the thickness of the reflector. The liquid difficult to perform in practical applications. So we higher the scattering cross section is, the smaller use graphite as a thick layer of the design device reflector. absorption cross section is, the higher the surface atom Thick layer of graphite external wrap hemispherical radius density is, the greater the thickness of the reflector is, the is 20cm, so the reflection can be a total reflection.

Submitted to ‘Chinese Physics C The upper part of the neutron source was wrapped by neutrons scattering cross section greater than the Paraffin. Because neutron source is isotropic, so we absorption cross section. Then subsequent decrease in determined the hemispherical as the geometry of thermal neutron yield with further increase in the moderation. Paraffin was surrounded by graphite. This moderator radius may be due to the absorption cross design compared with the previous design which used the section than scattering cross section greater due to parallel plate institutions, not only ensured the moderator increasing thermal neutrons flux rate with increasing of the neutron source fully moderated, but also reduced the moderator radius.The fast neutrons flux rate reduces moderator material for thermal neutron absorption. linearly with the increase of the paraffin radius. As the The thin beryllium reflector size, which was shown of Fig.11, we can determine, when the paraffin optimized through thermal neutrons flux rate of the sample. radius of 7cm, reflector of the graphite, the neutron The calculated flux rate of the sample is plotted in Fig.10. source slowing-down device moderating effect is best.

-4 Thermal neutron flux rate 7.5x10 Tab. 1 The resluts of two designs compares with no sample Fast neutron flux rate 7.0x10-4 Reflection Conventional 6.5x10-4 Times structure structure 6.0x10-4

5.5x10-4 Thermal 5.49555E-4 1.81875E-04 5.0x10-4 Counts flux rate 3.02E+00 4.5x10-4 Fast 4.0x10-4

3.5x10-4 neutron flux 4.12318E-4 1.26135E-04

3.0x10-4 0 1 2 3 4 5 6 7 8 9 rate 3.27E+00 Beryllium Radius(cm) Fig. 10 Thermal Gamma ray 1.03028E-04 3.18849E-05 neutrons and fast neutrons flux rate of differernt radius beryllium flux rate 3.23E+00

0.0008 The ratio of

Thermal neutron flux ratio thermal, 0.0007 Fast neutron flux ratio fast neutron 0.0006 output flux 1.33E+00 1.44E+00 9.24E-01

0.0005 As shown of Tab.1, in this paper, design of neutron

0.0004

Counts source moderated device compared with Saudi device, 0.0003 the amount of thermal output and fast all have large 0.0002 increase has increased significantly. Thermal neutron 0.0001 output grows by 3.02 times. Fast neutron output grows 0 2 4 6 8 10 12 Paraffin radius(cm) by 3.27 times. Fast neutron ratio percentage has Fig. 11 Thermal neutrons and fast neutrons flux rate of differernt increased smaller. Purpose is to reduce the proportion of radius of moderation fast neutrons, reducing neutron inelastic collisions The results of these calculations are shown in Fig. nuclides produced by γ-rays, γ to minimize the 10. It is clear that when beryllium thickness increases, background. Increase the proportion of thermal neutrons, the thermal neutron flux increases, fast neutron flux to improve the thermal neutron capture spectra of the decreases, but the increase and decline are smaller. Due sample measured spectral contribution of the γ, to the price of beryllium is higher, and the effect is not simplified lines, improve the analysis accuracy.the noticeable, so the reflector is all made of calculated yield of the gamma rays from prompt γ-ray graphite.Thermal and fast neutron calculations yield is neutron activation is plotted in Fig.12 . As shown of plotted in Fig.11 as a function of front moderator paraffin Fig.12, the new structure of the count rate was higher radius. When paraffin radius increases, the thermal than the previous structure. The energy spectral shape is neutron flux increases, fast neutron flux decreases. The basically the same. Fig.12 illustrates the new structure to thermal neutron flux rate initially increases with the improve the prompt γ count rate which did not make the moderator radius and have a maximum value for 6-7cm excitation spectrum become complicated simultaneously. radius moderator, then began to decrease. The initial The same neutron source improves the utilization rate of increase in the thermal neutrons yield may be due to the neutron with the new structure.

Submitted to ‘Chinese Physics C

-6 Old 1.20x10 1E-6 New

-7 9.60x10 1E-7 -7 7.20x10

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20 15 0.00-2 0 10 1E-10 -1 5 5 -10 Radius-5 (cm) 0 0 -5 5 -1 0 Radius (cm) 10 -1 5 1E-11 15 -2 0 20 0 2 4 6 8 Fig. 14 Previous design thermal neutron flux rate of second layer Energy(Mev) -7 Fig. 12 New and previous design gamma rays yield from prompt 7.00x10

-7 γ-ray neutron activation 5.60x10

-7 Fig.13-Fig.22 is the thermal neutron distributions of 4.20x10 the two designs in the sample. The samples were divided -7

Counts 2.80x10 into 4 × 4 × 4cm grids. The results of these calculations -7 1.40x10 are shown in Fig. 13-Fig.22. It is clear that new design 20 15 0.00-2 0 10 -1 5 5 -10 not only on the thermal neutron flux structure has Radius-5 (cm) 0 0 -5 5 -1 0 Radius (cm) 10 -1 5 significantly improved, and thermal neutrons in the 15 -2 0 20 sample distribution more uniform. It means that the Fig. 15 Previous design thermal neutron flux rate of third layer

-7 activation of the sample thermal neutrons are more 4.5x10 uniform, which is more conducive to large samples and -7 3.6x10 heterogeneous samples for analysis. -7 2.7x10

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-6 1.68x10 -8 9.0x10 -6 1.26x10 20 15 0.0-2 0 10 -1 5 -7 5 -10 0 Counts 8.40x10 Radius-5 (cm) 0 -5 5 -1 0 Radius (cm) -7 10 -1 5 4.20x10 15 -2 0 20 20 15 0.00-2 0 10 Fig. 16 Previous design thermal neutron flux rate of forth layer -1 5 5 -10 Radius-5 (cm) 0 0 -5 5 -1 0 Radius (cm) 10 -1 5 15 -2 0 20 Fig. 13 Previous design thermal neutron flux rate of fist layer

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20 20 15 15 0.00-2 0 10 0.00-2 0 10 -1 5 5 -1 5 5 -10 0 -10 Radius-5 (cm) Radius-5 (cm) 0 0 -5 0 -5 5 -1 0 Radius (cm) 5 -1 0 Radius (cm) 10 -1 5 10 -1 5 15 -2 0 15 -2 0 20 20 Fig. 17 Previous design thermal neutron flux rate of fifth layer Fig. 18 New design thermal neutron flux rate of first layer

Submitted to ‘Chinese Physics C

-6 3.75x10 Techniques in Geoscience Key and State Key Laboratory

-6 of Geohazard Prevention & Geoenvironmental Protection. 3.00x10

-6 2.25x10 参考文献

-6 Counts 1.50x10 [1] Gardner R P, Guo P, Wang Y Y, et al. Feasibility of

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-6 1.60x10 determination of sulphur in coal seams and overburden rock

-6 1.28x10 by PGNAA[J]. Nuclear Instruments and Methods in Physics

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Counts 6.40x10 [3] Naqvi A A, Nagadi M M, Al-Amoudi O S B. -7 3.20x10 20 Interferences in Prompt γ Analysis of corrosive 15 0.00-2 0 10 -1 5 5 -10 Radius-5 (cm) 0 0 -5 contaminants in concrete[J]. Nuclear Instruments and 5 -1 0 Radius (cm) 10 -1 5 15 -2 0 20 Methods in Physics Research Section A: Accelerators, Fig. 20 New design thermal neutron flux rate of third layer Spectrometers, Detectors and Associated Equipment. 2006, -6 1.05x10 569(3): 803-809.

-7 8.40x10 [4] Naqvi A A, Garwan M A, Nagadi M M, et al.

-7 6.30x10 Non-destructive analysis of in fly ash cement

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20 15 0.00-2 0 10 and Associated Equipment. 2009, 607(2): 446-450. -1 5 5 -10 Radius-5 (cm) 0 0 -5 5 -1 0 Radius (cm) [5] Oliveira C, Salgado J, Gon C C Alves I F, et al. A Monte 10 -1 5 15 -2 0 20 Carlo study of the influence of the geometry arrangements and Fig. 21 New design thermal neutron flux rate of forth layer structural materials on a PGNAA system performance for -7 7.00x10 cement raw material analysis[J]. Applied radiation and -7 5.60x10 isotopes. 1997, 48(10): 1349-1354.

-7 4.20x10 [6] Ghassoun J, Mostacci D. A compact neutron beam

-7 Counts 2.80x10 generator system designed for prompt gamma nuclear

-7 1.40x10 activation analysis[J]. Applied Radiation and Isotopes. 2011, 20 15 0.00-2 0 10 -1 5 5 -10 69(8): 1138-1142. Radius-5 (cm) 0 0 -5 5 -1 0 Radius (cm) 10 -1 5 [7] Naqvi A A, Abdelmonem M S, Al-Misned G, et al. 15 -2 0 20 ` Fig. 22 New design thermal neutron flux rate of fifth layer Performance improvement of keV Neutrons-based PGNAA 4. Conclusion setups[J]. Applied Radiation and Isotopes. 2006, 64(12): Monte Carlo simulation was carried out to design a 1631-1636. PGNAA thermal neutron output setup for increase the [8] Oliveira C, Salgado J. Calibration curves of a PGNAA thermal neutron output efficiency. The new design system for cement raw material analysis using the MCNP PGNAA moderation neutron source device is 3.02 times code[J]. Applied radiation and isotopes. 1998, 49(12): flux of thermal neutron and 3.27 times flux rate 1685-1689. compared with previous device. It improves the [9] Naqvi A A. A Monte Carlo comparison of PGNAA utilization of the fission neutron source. system performance using 252Cf neutrons, 2.8-MeV neutrons 5. Acknowledgements and 14-MeV neutrons[J]. Nuclear Instruments and Methods in The authors wish to acknowledge the support of the Physics Research Section A: Accelerators, Spectrometers, Chengdu University of technology Applied Nuclear Detectors and Associated Equipment. 2003, 511(3): 400-407.

Submitted to ‘Chinese Physics C [10] Al-Jarallah M I, Naqvi A A, Fazal-Ur-Rehman, et al. Fast Physics Research Section B: Beam Interactions with Materials and thermal neutron intensity measurements at the KFUPM and Atoms. 2002, 195(3–4): 435-441. PGNAA setup[J]. Nuclear Instruments and Methods in