Source substitution and optimization of CBXTM PGNAA analyzer in Kangan cement plant

Source substitution and optimization of CBXTM PGNAA analyzer in Kangan cement plant

Progress in Nuclear Energy 90 (2016) 204e211 Contents lists available at ScienceDirect Progress in Nuclear Energy journal homepage:

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Progress in Nuclear Energy 90 (2016) 204e211

Contents lists available at ScienceDirect

Progress in Nuclear Energy journal homepage:

Source substitution and optimization of CBXTM PGNAA analyzer in Kangan cement plant Kamal Hadad a, b, *, Hoda Sadeghpour a, b, M.R. Nematollahi a, b a b

Department of Nuclear Engineering, Shiraz University, Molla-Sadra St., 71348-51154, Shiraz, Iran Nuclear Safety Research Center, Shiraz University, Molla-Sadra St., 71348-51154, Shiraz, Iran

a r t i c l e i n f o

a b s t r a c t

Article history: Received 18 September 2015 Received in revised form 7 March 2016 Accepted 31 March 2016

Prompt Gamma Neutron Activation Analysis (PGNAA) is a powerful nondestructive, elemental analysis tool that make use of the time-based characteristics of a nuclear reaction to enhance the capability for multi-elemental analyses. In the present study, improving the moderator and gamma radiation shield of a CBX™ PGNAA analyzer system in Kangan cement factory, as well as substituting a longer half-life neutron source are investigated by utilizing Monte Carlo simulations. The results of the Monte Carlo simulations indicate that high density polyethylene (HDPE) with a thickness of 5 cm as the moderator and lead with a thickness of 4 cm as the gamma radiation shield are the best alternatives for the intended purposes. The long half-life (~433 years) of AmeBe would provide an almost constant level of neutron source over the lifetime of the equipment (~20 years) and could be substituted for 252Cf (T1/2 ¼ 2.65 years) as neutron source. Meanwhile AmeBe has higher prompt gamma yield than original source which makes it more suitable for PGNAA. Without much complications, the improved system could be implemented in the original CBX™ PGNAA system. © 2016 Elsevier Ltd. All rights reserved.

Keywords: PGNAA Cement Cf-252 AmeBe Shielding Moderation

1. Introduction Neutron activation analysis (NAA) is a sensitive analytical technique to determine the components of materials (often with an accuracy of one ppb). This method is implemented with either delay gamma neutron activation analysis (DGNAA) or prompt gamma neutron activation analysis (PGNAA). PGNAA is a fast, online and the non-destructive elemental analysis method which is implemented in wide range of applications including environmental (Naqvi et al., 2012, 2013), health sciences (Panjeh and Izadi-Najafabadi, 2011) and industrial (IAEA, 2000; Naqvi et al., 2011). In this method, the concentration and percentage of the elements in a sample are determined by the characteristic gamma intensity, which is generated by the collision of thermal neutrons with the sample through (n, n0 g) and (nth,g) interactions (Ali-Abdallah and Mansour, 2011; YANG et al., 2012). The main capability of PGNAA method is the determination of light elements (Ca and below). These elements form the major components of rocks and minerals, cement and concrete etc. The capture

* Corresponding author. Department of Nuclear Engineering, Shiraz University, Molla-Sadra St., 71348-51154, Shiraz, Iran. E-mail address: [email protected] (K. Hadad). 0149-1970/© 2016 Elsevier Ltd. All rights reserved.

cross sections are characteristically a few tenths of a barn for these elements, which allows their assay even in weaker neutron beams. PGNAA could be implemented with 3 radionuclide sources. Two of them are a -emitter embedded in Be, using the 9Be(a, n)12C reaction, and the a-emitting source being either 241Am or 238,239Pu. The third radionuclide source is the 252Cf spontaneous fission source. The most common source used in PGNAA is 252Cf which is a spontaneous fission source with neutron mean energy of 2.1 MeV and neutron most probable energy of 0.7 MeV. Given the high average energy of the neutrons, they need to be moderated to the limit of thermal energy to improve the efficiency of the device through increasing prompt gamma flux for (nth,g) reactions. Materials used as the neutron moderator should contain light elements with high scattering cross sections and low absorption. Also, 252 Cf source emits a continuous spectrum of gamma rays with the energy range 0e10 MeV, which should be shielded as much as possible to minimize their destructive effects and potential interaction with the main spectrum of prompt gamma. In this study, Monte Carlo calculations were performed for the gamma shield design and the suitable moderator for a252Cf neutron source used in the elemental analysis instrument by PGNAA in CBX™ Analyzer. Because of efficiency and exposure time constraints required in

K. Hadad et al. / Progress in Nuclear Energy 90 (2016) 204e211


AmeBe source advantages include,

  Scfg Eg ¼

 greater half-life,  higher average neutron energy which provides the possibility of detecting a greater number of elements,  gamma ray energies are almost discrete and make no interference with the prompt gamma spectrum obtained from the sample material. Consequently, the source needs not to be shielded from the detector The main purpose of this research is to investigate the feasibility and design parameters of the present CBX™ PGNAA system and a new system when 252Cf is replaced with AmeBe source. Furthermore, the partial objectives are: 1 Maximizing neutron moderation to enhance thermal neutron flux which results in an increase in prompt gamma ray flux emanating from an irradiated sample material. It should also be noted that the moderation enhancement should not significantly increase the background gamma. 2 Minimizing the gamma-ray radiated from the source and any delayed (n,g) reaction. 3 Modeling of system based on californium source and cement activation calculation. 4 Modeling of new proposed system based on the earlier configuration except that an AmeBe replaces the 252Cf source and optimizing the system that can be implemented in CBX TM analyzer system. 5 Sensitivity Analysis of Important parameters to improve the analyzer system with the alternative source. The detailed work flow of this study is drawn in Fig. 1. To sum up, we first optimize shielding and moderation of the present PGNAA system to achieve the highest yield of prompt gamma spectra from the Kangan cement samples. Subsequently, we substitute the source with AmeBe and reevaluate the moderation and shielding for the new source to optimize prompt gamma spectra from samples. The prompt gamma spectra for both sources are compared and conclusions are drawn in the final section. 2. Material & method 2.1. Geometrical modeling of CBX™ analyzer consists of



Using the Monte Carlo code, MCNPX2.7 (Pelowitz et al., 2008), the geometry of CBX™ online analyzer is modeled. This model (Fig. 2) consists of 252Cf source container, moderator, conveyor belt,


Sinhð2:926En Þ




375Eg2 eEg 0:109 þ 0:468eEg 1:457

Eg  1:5MeV

eEg 0:851

Eg > 1:5MeV



En 1:025


1 þ 42 a/12 6 C þ 0n þ g

 Scfn ðEn Þ ¼ 0:30033e


9 4 Be

þ 42 a

2.1.1. 252Cf source modeling 20mg of 252Cf with 400MBq (10.7 mCi) activity, producing 4.6  107 neutron per seconds, is inside the standard X.1 capsule code CVN7 are modeled by the Monte Carlo code (TechnologyA, Sourc, 2004). Fig. 3 illustrates the source geometry. Watt Fission Spectra is used for neutron energy spectra and experimental data reported by Glassel (Glassel et al., 1989), is used for the gamma energy spectrum of MCNP source model. The spectra for neutron and gamma are given in Eqs. (3) and (4) and plotted in Figs. 3 and 4 respectively. =

241 237 95 Am/93 NP

sample substance and the detector box containing two NaeI detectors (7.62 Cm x 7.62 Cm (3  3 inch)).


industrial facilities, the use of radioisotope neutron sources requires periodic replacement of the source. The current 252Cf source in CBX™ analyzers, while desirable from a point of view of spectral characteristics, is difficult to prepare and replace. Thus, a more readily available replacement radioisotope source is sought to attain as closely as possible the specifications and features of 252Cf. According to the previous research, AmeBe sources has shown good conformity with 252Cf (Naqvi et al., 2006). AmeBe sources produce neutron as result of Beða; nÞ reaction in which a is the He nucleus emitted from 241 Am spontaneous decay. Subsequently, the Beða; nÞ reaction produces a 12 C nucleus, a single neutron and gamma ray resulting from the decay of the excited state of 12 C.


(4) In Eqs. 3 and 4, E is in terms of MeV and the ratio of gamma to neutron for each fission is 2.132. 2.1.2. Gamma shield modeling Gamma radiated from 252Cf has a continuous energy spectrum of 0e10 MeV. Thus, when this radioisotope is used as a PGNAA source, its high intensity gamma rays could overlap the characteristic prompt gammas of (nth,g) interactions. The gamma shield should be designed in a way that neither reduces the neutron flux nor generates high energy and intensity prompt gamma rays through (nth,g) interaction. Conventional materials used as gamma shield are lead and bismuth. To determine the optimum shield thickness and material, source is enclosed by cubes of different sizes and materials and gamma and neutron fluxes, exterior to the cubes, are evaluated. 2.1.3. Moderator modeling Since 252Cf is a source with a relatively high fast neutron flux and given that the PGNAA technique is based thermal neutron interaction with sample, neutrons are needed to be slowed down (moderated) before reaching to sample (Zhang and Tuo, 2013; Abdelmonem et al., 2007). Accordingly, the objective is to increase thermal neutron and reduce fast neutron fluxes in PGNAA analysis. Meanwhile the moderator should not increase the background gamma. In this research, the materials considered being suitable as moderator are: high density polyethylene (HDPE), paraffin, graphite and heavy water. The choice of the best moderator and the optimal thickness was made on the basis of two goals: 1. Maximizing thermal neutron flux minimizing fast neutron flux, 2. Maximizing characteristic prompt gamma flux for each element. For both goals, various thicknesses of the aforementioned 4 moderators are modeled in Monte Carlo and their moderating properties are investigated utilizing Sða; bÞ cross-section library of MCNPX. To reach the first goal, Thermal and fast neutron fluxes as well as gamma ray flux obtained from neutron collision with 4 moderating materials of different thicknesses are calculated and compared. In the second goal, characteristic prompt gamma flux resulting from thermal neutron interaction with the sample material is evaluated with the Monte Carlo code.


K. Hadad et al. / Progress in Nuclear Energy 90 (2016) 204e211

Fig. 1. Workflow in CBX™ optimization and source substitution modeling.

Fig. 2. Geometry of the CBXTM analyzer.

2.1.4. Evaluation of prompt gamma rays spectra Prompt gamma ray spectra are evaluated for three samples typical in the Kangan cement plant (Table 1). Identification of elements according to their characteristic gamma ray energy is presented in Table 2 (Firestone et al., 2004). In PGNAA, each characteristic peak in gamma spectrum belongs to an element. Consequently, elemental identification and its concentration could be estimated from the gamma ray spectra with proper calibration. 2.2. Geometrical modeling of CBX™ analyzer with AmeBe 2.2.1. AmeBe source modeling Retaining the physical geometry of CBX analyzer, AmeBe source with the same activity as 252Cf is modeled in MCNPX. Two standard X.14 capsule code AMN25 (TechnologyA, Sourc, 2004) (370 GBq

Fig. 3. Standard X.1 capsule (Cf-252). (Pelowitz et al., 2008).

each and 20 Ci total) would provide an approximate neutron emission of 4.0  107 n/s. This is the closest neutron emission comparable to 4.6  107 n/s emission of the previous neutron source (i.e. 252Cf). Data for standard capsule of AmeBe (X.14 model) is obtained from ISO/DIS8529-1 (International Standard, 2000). Figs. 5 and 6 represent the source geometry and its neutron and gamma energy spectra which are modeled in SDEF command, embedded in MCNPX code input (Firestone et al., 2004; Vitorelli et al., 2005). The majority of gamma rays emitted from AmeBe source result

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Fig. 4. Normalized neutron and Gamma spectra of



Fig. 5. Standard X.14 capsule (AmeBe) (Pelowitz et al., 2008).

Table 1 Elemental Compositions in 3 typical Kangan cements (weight percent).

Si Al Fe Ca Mg S K Na O




0.007793 0.00500 0.002258 0.362643 0.00882 0.00068 0.000991 0.000807 0.167023

0.046433 0.010833 0.010400 0.312714 0.006540 0.002480 0.006526 0.00088 0.202726

0.052080 0.007367 0.012384 0.293357 0.018540 0.002600 0.002313 0.001100 0.206526

Table 2 Identification of elements (Abdelmonem et al., 2007).









Gamma Energy (MeV)

sðEg ÞðbarnsÞ

Si Si Fe Ca Fe Fe Al

3.54 4.93 6.02 6.42 7.63 7.65 7.72

0.12 0.31 0.23 0.18 0.65 0.55 0.05

from 241 Am decay with approximately 36% of gamma having 60 KeV and 42% of 14 KeV energies. Half thickness required to shield 60 KeV gammas is about 1 cm and 1 mm for aluminum and lead respectively. This amount of lead has a very small impact on the neutron flux which have energies from 10 KeV to 7 MeV with an average energy of 4.5 MeV. Moreover, 12C* produced from neutron collision with Be, emits gamma ray energy of 4.438 MeV and gamma to neutron density ratio is 0.596. Following the substituting AmeBe, gamma shield and moderator are modeled and optimized to achieve the goals mentioned in 2.1.2 and 2.1.3. 2.2.2. Optimized geometry for AmeBe Source position in CBX analyzer is optimized for its 252Cf source. With the new AM-Be source, its position, Since the geometry was optimized for 252Cf source, AmeBe source optimization needs to be performed. For this, thickness of moderator, source location and source-moderator distance are varied. Giving two described methods in the flowchart, the appropriate thickness of the

Fig. 6. a) Neutron and b) Gamma energy spectra for AmeBe (Firestone et al., 2004; Vitorelli et al., 2005).

moderator and source location are determined. The spectrum of new setup is compared with original CBX™ with 252Cf source. 3. Results Based on flow chart illustrated in Fig. 1, the results are presented


K. Hadad et al. / Progress in Nuclear Energy 90 (2016) 204e211

here. 3.1. CBX™ modeling with original


Cf source

3.1.1. Shielding design As shown in Fig. 7, increasing the shield thickness leads to reduction in gamma flux as well as increasing the neutron flux which is due to (n, 2n) and (n, 3n) reactions in shielding material. Lead produces slightly higher neutron flux than bismuth (Figs. 7 and 8). Increasing the lead thickness up to 4 cm, shields 94% of the gamma rays. Meanwhile, neutron energy shifts to lower spectrum and the shield acts as a neutron moderator (Vega-Carrillo et al., 2006). 3.1.2. Moderator design Based on description in section 2.1.3 the moderator is designed in two steps as follows. Moderator design based on maximizing thermal neutron flux. Fig. 9 represents thermal neutron flux versus moderator thickness. The reason for thermal neutron flux reduction in high thickness of moderator is that a large number of thermal neutrons being absorbed in the moderator. The energy spectrum of moderated neutrons with D2O and graphite, are placed in epithermal region which shows the lack of ability of these moderators for thermal moderation. Moreover with D2O and Graphite, optimum thickness for maximum thermal neutrons yield is too high (30 and 50 cm respectively). Due to high hydrogen content in HDPE and paraffin, higher moderating power is presumable which results in lower moderator thickness (~5 cm). Among 4 moderators studied, HDPE yields the highest thermal neutron flux. As shown in Fig. 9, thermal neutron yields of HDPE rises to a maximum value and falls to its minimum value as the moderator thickness increased. The rise and fall in thermal neutron flux with thickness is the due to domination of scattering cross section and absorption cross sections respectively. As shown in Fig. 10, paraffin and HDPE produce 2.223 MeV gamma ray as a result of (n; g) reaction which is absent with D2O and Graphite moderators.

Fig. 8. Relative neutron flux versus variation in thickness of gamma shield relative to the air.

Fig. 9. Thermal neutron flux for 4 moderator substances versus moderator thickness. Moderator design based on maximizing prompt gamma from sample material. Optimized moderator thickness and type is also determined through maximizing prompt gamma flux emitted from a cement sample of 20 cm thick. In Fig. 11 (a,b,c), the emitted characteristic prompt gamma flux from calcium, iron and silicon

Fig. 10. Gamma flux for 4 moderator substances versus moderator thickness.

Fig. 7. Relative gamma flux versus variation in thickness of gamma shield relative to the air.

(present in Kangan cement samples) versus moderator thickness for each type of moderator are shown. With increasing the moderator thickness (Fig. 11), the prompt gamma flux increases to its maximum value and then by further increase, the flux falls off. Initial increase in prompt gamma flux is due to the increase in thermal neutron flux incident on the sample material. Since HDPE has the highest prompt gamma flux for every

K. Hadad et al. / Progress in Nuclear Energy 90 (2016) 204e211

209 Element analysis using the optimized HDPE moderator. Each element has a characteristic peak in the energy spectrum, which indicates the presence of the element in the sample. In addition, smaller peaks in the spectrum are the result of cumulative intensity of prompt gamma of different elements in the sample which have the same characteristic energy. This analysis is illustrated in Fig. 12. Prompt gamma spectrum by changing the composition of the sample. According to Table 1, since the percentage of elements in the sample is different, characteristic peak intensities are different. For example (Fig. 13), iron characteristic energy of 7.65 MeV, with respect to its different percentage in three materials under study, has different intensity. 3.2.


Cf source substitution

In order to assess the CBX™ analyzer capability to work with different neutron source, we substitute its original source (i.e. 252Cf) with AmeBe neutron source. AmeBe is placed in the same position as original 252Cf. With the same activity for both sources (i.e. 4:6  107 n=s), AmeBe produces higher counts of prompt gamma rays than 252Cf. Due to higher average energy of neutrons emitted from AmeBe source, the possibility of inelastic interactions rises and PGNAA method is capable of detecting elements such as oxygen, which is not identifiable by 252Cf. Likewise for calcium element, due to inelastic collisions, two peaks of 3.74 MeV and 3.91 MeV are identifiable (Fig. 14). 3.3. PGNAA analyzer design based on AmeBe source The size of original container box of CBX™ analyzer is preserved and the following modifications are made to improve characteristic prompt gamma flux from the cement sample: a) neutron source position with respect to moderator, b) moderator thickness and position with respect to the source position. The best position for AmeBe source as well as the optimum moderation thickness is investigated based on highest thermal neutron flux in front of the cement sample. Fig. 15 illustrates the variation of thermal neutron flux versus source position for different moderator thicknesses. As evident from Fig. 15, when source located within 0.5e1.5 cm form the moderator, maximum

Fig. 11. The prompt gamma flux of a) calcium, b) iron and c) silicon elements in DGW sample.

element in the sample, it could be selected as the optimum moderator.

3.1.3. Analysis based on prompt gamma spectrum from 252Cf source Presence of two high intensity peaks of 2.223 MeV and 2.61 MeV belonging to hydrogen and lead, lowers the resolution of the characteristics peaks of cement elements in the spectrum. Since this two peak are less than 3 MeV and are out of energy range mentioned in Table 2, they could be marginalized to raise the resolution in elemental peaks.

Fig. 12. Prompt gamma spectra obtained from the DGW sample on the basis of source.




K. Hadad et al. / Progress in Nuclear Energy 90 (2016) 204e211

Fig. 13. Prompt gamma spectra obtained from three samples on the basis of

Fig. 14. Comparison of prompt gamma spectrum from DGW Sample for

thermal neutron flux occurs. As moderator thickness increases, the thermal flux increases as well and saturates to a steady value within 6e8 cm thickness. Thicker moderator increases and absorbs thermal flux, therefore further increase in moderator thickness has minimum effect. The result of new CBX™ analyzer design with AmeBe source is presented in Fig. 16. As seen in Fig. 16, the new CBX™ design, based on AmeBe neutron source, yields higher elemental prompt gamma characteristic flux than the original CBX™ with 252Cf or AmeBe sources. Meanwhile the new design utilizing AmeBe source could identify more elements which may be useful for other applications of this analyzer.



Cf source.

Cf and AmeBe Sources.

4. Conclusion In this study we investigated the CBX™ PGNAA analyzer in order to: 1 Improve its shield for gamma and neutron, 2 Improve its characteristic prompt gamma flux output, 3 Possibility of substituting its neutron source with a longer halflife source (i.e. AmeBe neutron source). Both lead and bismuth, due to their shielding properties and availabilities, are investigated. A 4 cm lead shield is shown to reduce the activity as much as 94% (Figs. 7 and 8).

K. Hadad et al. / Progress in Nuclear Energy 90 (2016) 204e211


cement samples (obtained from Kangan Cement Plant) indicates that the improved CBX™ system conclusively detects the main elements present in the sample and is distinguishable from background (Fig. 13). We then substituted 252Cf with AmeBe neutron source and studied the PGNAA spectrum. The results indicate that the new AmeBe source produces higher yield of prompt gamma characteristic flux than original 252Cf source (Fig. 14). This system was also improved by optimizing source location and moderator thickness and position (Fig. 15). The new system revealed to produce the highest yield of prompt gamma ray flux among all systems (Fig. 16). The optimized system could be implemented effortlessly in CBX™ thermal analyzer to achieve longer life and robust detection and analysis of cement samples. References Fig. 15. Thermal neutron flux versus variation in source position for different moderator thickness where M2 through M8 represents 2 through 8 cm moderator thickness.

Fig. 16. Comparison of prompt gamma spectrum from DGW Sample for AmeBe Source in new design.


Cf and

The improvement of the moderator of the original design using 4 different moderators is studied in two steps. At first, the moderator is improved to yield the highest thermal neutron (Fig. 9) and second, highest yield of prompt gamma characteristic flux (Fig. 11). The result indicates that 5 cm HDPE satisfies both objectives. Modeling of the whole analyzer with DGW, DPW and DP6

Abdelmonem, M.S., Naqvi, A.A., Al-Misned, G., Al-Ghamdi, H., 2007. Performance improvement of a PGNAA setup due to change in moderator design: a Monte Carlo study. J. Radioanal. Nucl. Chem. 271 (3), 685e690. Ali-Abdallah, M., Mansour, N.A., 2011. Neutron activation analysis of cement bulk samples. Adv. Appl. Sci. Res. 2, 613e620. Firestone, R.B., et al., 2004. Database of prompt gamma rays from slow neutron capture for elemental analysis. Lawrence Berkeley Natl. Lab. Glassel, p., Schmid-Fabian, R., Schwalm, D., Habs, D., Helmolt, H.U.V., 1989. 252Cf fission revisited e new insights into the fission process. Nucl. Phys. 502, 315e324. IAEA, 2000. Emerging New Applications of Nucleonic Control Systems in Industry. International Standard ISO/DIS 8529-1: Reference Neutron Radiations- Part 1, 2000. Naqvi, A.A., Abdelmonem, M.S., Al-Misned, G., Al-Ghamdi, H., 2006. New sourceemoderator geometry to improve performance of 252Cf and 241AmeBe source-based PGNAA setups. Nucl. Instrum. Methods Phys. Res. Sect. A Accel. Spectrom. Detect. Assoc. Equip. 562 (1), 358e364. Naqvi, A.A., Maslehuddin, M., Garwan, M.A., Nagadi, M.M., Al-Amoudi, O.S.B., Raashid, M., 2011. Estimation of minimum detectable concentration of chlorine in the blast furnace slag cement concrete. Nucl. Instrum. Methods Phys. Res. Sect. B Beam Interact. Mater. Atoms 269 (1), 1e6. Naqvi, A.A., Al-Matouq, F.A., Khiari, F.Z., Isab, A.A., Rehman, K.U., Raashid, M., 2012. Prompt gamma tests of LaBr 3: Ce and BGO detectors for detection of hydrogen, carbon and oxygen in bulk samples. Nucl. Instrum. Methods Phys. Res. Sect. A Accel. Spectrom. Detect. Assoc. Equip. 684, 82e87. Naqvi, A.A., Al-Matouq, F.A., Khiari, F.Z., Gondal, M.A., Isab, A.A., 2013. Optimization of a prompt gamma setup for analysis of environmental samples. J. Radioanal. Nucl. Chem. 296 (1), 215e221. Panjeh, H., Izadi-Najafabadi, R., 2011. Body Composition Analyzer Based on PGNAA Method. INTECH Open Access Publisher. Pelowitz, D.B., Hendricks, J.S., Durkee, J.W., Fensin, M.L., James, M.R., 2008. MCNPX 2.7. A Extensions. Los Alamos National Security. LLC, Los Alamos, NM. LA-UR08e07182. IAEA Technology QSA,Sources, 2004 chapter 2. ~ a, E., Hern vila, V.M., Gallego, E., Vega-Carrillo, H.R., Manzanares-Acun andez-Da Lorente, A., 2006. Neutron shielding for a 252 Cf Source. In: Annual Meeting XVII Annual SNM Congress, First American IRPA Congress. Vitorelli, J.C., Silva, A.X., Crispim, V.R., Da Fonseca, E.S., Pereira, W.W., 2005. Monte Carlo simulation of response function for a NaI (Tl) detector for gamma rays from 241 Am/Be source. Appl. Radiat. Isot. 62 (4), 619e622. YANG, J., YANG, Y., LI, Y., LIU, M., CHENG, Y., 2012. Monte-Carlo simulation of cement neutron field distribution characteristics in PGNAA. Nucl. Sci. Tech. 6 (006). Zhang, J., Tuo, X., 2013. PGNAA Neutron Source Moderation Setup Optimization. arXiv Preprint arXiv:1309.1308.