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Radiation Portal Monitor with 10B+ZnS(Ag) Neutron Detector Performance for the Detection of Special Nuclear Materials

Karen A. Guzmán-García1,*, Hector Rene Vega-Carrillo2, Eduardo Gallego1/2 Juan Antonio González3, Roberto Méndez4, Alfredo Lorente1/3 & Sviatoslv Ibañez-Fernandez1/4

1Departamento de Ingeniería Energética, ETSI Industriales Universidad Politécnica de Madrid, C. José Gutiérrez Abascal 2, 28006 Madrid, Spain

2Unidad Académica de Estudios Nucleares, Universidad Autónoma de Zacatecas C. Ciprés No. 10 98060, Zacatecas, Zac. Mexico

3Laboratorio de Ingeniería Nuclear, ETSI Caminos Canales y puertos Universidad Politécnica de Madrid, C. Prof. Aranguren 3, 28040, Madrid Spain

4Laboratorio de Patrones neutrónicos. CIEMAT, Av. Complutense 40, 28040, Madrid Spain *email: [email protected]

Abstract

In homeland security, neutron detection is used to prevent the smuggling of Special Nuclear Materials. Thermal neutrons are normally detected with 3He proportional counters, in the Radiation Portal Monitors, RPMs, however due to the 3He shortage new procedures are being studied. In this work Monte Carlo methods, using the MCNP6 code, have been used to study the neutron detection features of a 10B+ZnS(Ag) under real conditions inside of a RPM. The performance for neutron detection was carried out for 252Cf, 238U and 239Pu under different conditions. In order to mimic an actual situation occurring at border areas, a sample of SNM sited inside a vehicle was simulated and the RPM with 10B+ZnS(Ag) response was calculated. At 200 cm the 10B+ZnS(Ag) on RPM response is close to 2.5 cps-ng 252Cf, when the 252Cf neutron source is shielded with 0.5 cm-thick lead and 2.5 cm-thick polyethylene fulfilling the ANSI recommendations. Three different geometries of neutron detectors of 10B+ZnS(Ag) in a neutron detection system in RPM were modeled. Therefore, the 10B+ZnS(Ag) detectors are an innovative and viable replacement for the 3He detectors in the RPM.

Keywords: Special nuclear materials, Neutron, Radiation portal monitors, MCNP6

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Since terrorist attacks in the United States in September of 2001, heightened concern with regard to critical infrastructure security and methods necessary for guaranteeing the safety of the general public [Spence 2011]. The Radiation Portal Monitors (RPMs) deployed at ports, railways, airports and vehicle checkpoints, used to screen in vessel, vehicles, and individuals in order to thwart the illicit trafficking of SNM [Weltz et al., 2015]. The RPMs have played important role in preventing the illicit trafficking and transport of nuclear and radioactive materials [Kwak et al., 2010].

With an increase in the capabilities and sophistication of terrorist networks worldwide comes a corresponding increase in the portability of radiological or nuclear devices being detonated in any place of the world. One method to decrease the risk associated with this threat is to interdict the material during transport of goods. Current RPMs have limitations in their ability to detect shielded nuclear materials [Spence 2011]. But also still play an important role in national security.

Current RPM deployed at borders are generally equipped with two types of radiation sensors; high efficiency gamma detectors commonly based on plastic of Poly- Vinyl-Toluene (PVT), and neutron detectors that are exclusively based on helium-3, 3He proportional counters, consisted of 3He tubes surrounded in polyethylene for neutron detection, and plastic scintillators for detection [Peerani et al., 2012].

Neutron detection is important for homeland security efforts, including monitors national point of entry the presence of special nuclear material (SNM), which is defined as and enriched in 233U or 235U and is a fissile component of nuclear weapons [Weltz et al., 2015; Kouzes et al., 2009].

Most current neutron detection systems used for nuclear security are based on 3He (technology) proportional counter, these are robust, gamma insensitive, and remain an efficient, proven technology for detecting thermal neutrons. However, a global shortage of

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The aim of this work is to study the neutron detection features of the 10B+ZnS(Ag) detectors, using Monte-Carlo neutron-transport, with the MCNP6 code under real conditions inside on RPMs, in border areas, with a vehicle-based simulation in similarly to the RPMs benchmark simulation, in order to study the performance for each detector for detect a neutron source of 252Cf, and SNM, (HUE 70% 238U) and 239Pu under two different conditions; bare and shielding, for each detection system.

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2.- MATERIALS AND METHODS

10 2.1.-Description of the N-15, N-48 and N-15 plus of B+ZnS(Ag)

First of all three different detectors are going to be explained, they are detectors of 10B+ZnS(Ag), N-15 “nDetBrick”, N-48 and the N-15plus.

The N-15 and N-48 were manufactured by BridgePort Instruments, LLC. Both use a mixture of 10B high enrichment, with a detector ZnS(Ag), 10B+ZnS(Ag), as neutron detection. The screens are on polymethyl methacrylate, PMMA, with twofold, as a neutron moderator and as light guide, guide the pulses to a photomultiplier tube embedded high voltage supply, and a multichannel analyzer eMorpho® digital electronics.

The sensitive area of each detector is formed by 5 transparent layers of 10B+ZnS(Ag) deposited on 4 plates of PMMA of 23 x 36 x 0.635 cm ( N-15), 120 x 15.2 x 0.635 cm (N- 48). The PMMA acts as light guide and as moderator. All is surrounded by ~8µm thick aluminum mylar as light reflector. The internal configuration is shown in Figure 1.

Figure 1.- Internal configuration N-15 and N-48

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These detectors has similar geometry, both are rectangular and with the same internal array. The external size of the N-15 detector is 23 x 36 x 4 cm, and the external dimensions of the N- 48 detector are 141.5 x 16.7 x 6.35 cm. Each detector has an outer moderator t of high- density polyethylene, (0.94gr/cm3), HDPE. For the N-15 detector the moderator thickness is 24 mm in the front, lateral faces, and top, bottom 36 mm while 48 mm-thick in the back (24+36+48 mm). The N- 48 is 25 mm-thick in front, top, bottom, and lateral faces, while 50 mm-thick in the back (25+50 mm) as shown in Figure 2.

Figure 2.- Configuration N-15 and N-48 detectors, bare and moderated.

A new model of neutron detector was studied; prototype, [Guzmán-García et al., 2016] based on previous studies of the N-15 neutron detector, and was called N-15 plus, which is an improvement geometry respect to the manufactured detector, N-15 ―nDetBrick‖.

The difference respect to the actual N-15 is the PMMA thickness, from 0.635 to 0.800 cm and the sensitive area has 30% of 10B higher than the actual detector, The internal configuration can be seen in Figure 3.

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Figure 3.- N-15 plus model MCNP configuration.

In summary, the N-15 and N-48 are detectors manufactured and with similar internal geometry varying in the detection area and the N-15plus detector is an MCNP6 model based on previous studies from the N-15 detector [Guzmán-García et al., 2016]. These neutron detectors are studied as an alternative to the 3He detectors actual installed in the RPM, for this reason, models in RPM were made.

2.2.-Monte Carlo Calculations 2.2.1.-Radiation Portal Monitors, RPMs, models description

Using MCNP6 2.6.1 code version 6.1 [Pelowitz et al., 2014], models of three neutron detections system for RPMs were built including all the each detectors details such as: the sensitive layers, the PMMA, PTM and their moderators, where the PTM was modeled as ta empty cylinder of glass. Each neutron detections systems (N-15, N-48 and N-15plus) are described below.

A model was made with the detector N-15 ―nDetBrick‖, three N-15 detectors were positioning in the parallel inside on an iron structure of 200 x 40 x 15 cm dimensions, each detector has the moderator of 24+48 mm HDPE, the first detector (N-15C) was positioned to 75cm from the midpoint from the detector respect to the ground, Figure 4.

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Figure 4.- N-15, Neutron detector System.

Other model was made the detector N-48, was modeled together with a PVT gamma detector of 190 x 41 x 5 cm dimenstions, surrounded by aluminum and shielding with lead and polyethylene, both, N-48 neutron detector and PVT gamma detector are inside on an iron structure of 235 x 80 x 18 cm dimensions, this configuration can see in Figure 5.

Figure 5.- Radiation Portal Monitor, RPM, PVT gamma and N-48.

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The N-15plus model were complemented defining three N-15plus detectors with two PVT gamma detectors of 40 x 60 x 4 cm, the PVTs surrounded by aluminum and a lead/polyethylene shield, all inside on an iron structure of 200 x 90 x 15 cm, the N-15 plus detectors were positioning in parallel, the first detector were positioned to 75 cm from the midpoint from the detector respect to the ground, Figure 6.

Figure 6.- Radiation Portal Monitor, RPM, two PVT gamma and three N-15plus detectors.

2.2.2.- Model of vehicle with SNM in customs.

A vehicle-based simulation was performed similarly to the RPMs benchmark simulations, each RPM described above were defined as an emplaced stationary detector objects pass by the monitor, RPM, and are screened for the presence of radioactive materials. In this case we suppose cargo scanning scenario, and we defined a vehicle passing on the RPMs.

The vehicle was defined as the Annex A, informative for the test vehicle [CEI-IEC 2004]. The dimensions of the box are 530 x 270 x 200 cm, and the cab has 120 x 250 x170 cm. In the box different materials were defined as a scrap, with ~730 kg, these scrap materials are steel, cooper, bronze, lead, zinc, aluminum, brass, and iron, simulating a neutron source of 252Cf or SNM, HEU or 239Pu. Scrap materials can see in Table 1.

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Table 1.-Scrap, different material in the box inside the vehicle Density Weight Material [g/cm3] [kg] Steel 304 7.87 78.70 Steel 8.00 104.00 Cooper 8.96 116.48 Bronze 8.40 75.60 Lead 11.35 79.45 Zinc 7.133 ~ 135.53 Aluminum 2.6989 ~ 29.69 Brass 8.07 56.49 Iron 7.874 ~ 55.12 TOTAL ~ 731.05

In the vehicle center in the box in the middle of the ―scrap‖ were defined a sphere of ~1.87cm, ~500 g of U, 70% enriched HEU (2,144n/s) [Spence 2011], first bare, and then shielded in concentric spheres of polyethylene and lead. Other case was a sphere of ~1.87cm, ~515 g of 239Pu (40,000 n/s ) [Perani et al., 2012], first bare, and then shielded in concentric spheres of polyethylene and lead. Other case was a neutron source of 252Cf moderated with 0.7 cm of HDPE. The different materials in the box and these materials were in randomly defined position and Figure 7.

Figure 7.- Scrap and the source

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The sources are characterized by a Maxwell energy distribution. In the use of MCNP, the Watts fission spectrum is used an alternative to the Maxwell fission spectrum for induced fission of the 252Cf, 239Pu and 238U. The Watts fission spectrum is modeled by the following equation:

( ) ( ) (√( ) (1)

In order to determine the H*(10) received by a driver when transport these type of materials in the interior of the cab. The driver was defined as the BOMAB Phantoms [Nuclear Technology Services, Inc 2016] that is an acronym for Bottle Manikin Absorber, theses BOMAB are designed to imitate, as nearly possible, human tissue for the purpose of calibrating whole body counters, these phantom consists on the cylinders of various shapes (circular and elliptical) dimensions which together comprise the average person. High density polyethylene, approximately 4.76 mm tick, is used in all parts of the phantom the dimensions are shown in the Table 2. Figure 8.

Figure 8.- BOMAB phantom

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Table 2.-BOMAB Phantom dimensions Quantity Shape Cross section Vertical height (cm2) (cm)

Head 1 Ellipse 19x14 20

Neck1 Circle 13Ø 10

Thorax 1 Ellipse 36x20 20

Lumbar 1 Ellipse 36x20 20

Thighs 2 Circle 16 Ø 40

Legs 2 Circle 13 Ø 40

Arms 2 Circle 10.5 Ø 60

For each case in three different positions, first in front of the N-15 detection systems, then in front of the RPM-N-48, and finally in the RPM-N-15plus, all these model can see in Figure 9.

In all the Monte Carlo calculations the number of histories was large enough to obtain uncertainties less than 5%. For the calculations, the cross sections were taken from the ENDF/B-VI library, where the S(α, β) treatment was included to take into account the thermalized neutron interactions [Vega-Carrillo et al., 2014].

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Figure 9.- Vehicle-based model with all RPMs in 3 different positions.

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3.- RESULTS

Table 3 shows the calculated detectors response for each source in the three positions described above, the response are the amount of 10B(n,α)7Li reactions per neutron form the source.

Table 3.- Results of each position and with the different materials System Source Position 1 Position 2 Position 3 conditions 252Cf* 2.31 ± 0.02 1.27± 0.06 0.35 ± 0.09

Neutron detector HEU** 2.59 ± 0.04 1.33 ±0.06 0.24 ±0.10 system three N-15 HEU Shield 1.54 ± 0.05 0.78 ±0.07 0.21 ± 0.13 “nDetBrick” 239 detectors Pu*** 54.24 ± 0.04 27.48 ± 0.05 7.21 ± 0.10 239 Pu Shield 33.55 ± 0.05 16.97 ± 0.07 4.40 ± 0.14 252Cf* 1.36 ± 0.02 1,77 ± 0.02 0.64 ± 0.02 LEFT HEU** Radiation Portal 1.37 ± 0.02 1.79 ± 0.01 0.65 ± 0.02 Monitor with N- HEU Shield 0.82 ± 0.02 1.07 ± 0.02 0.38 ± 0.03 48neutron 239 detector and PVT Pu*** 28.63 ± 0.02 37.67 ± 0.02 13.69 ± 0.02 gamma detector 239 Pu Shield 17.82 ± 0.02 23.56 ± 0.02 8.45 ± 0.03 252Cf* 1.35 ± 0.02 1,79 ± 0.02 0.65 ± 0.02 RIGHT Radiation Portal HEU** 1.35 ± 0.02 1.80 ± 0.02 0.66 ± 0.02 Monitor with N- HEU Shield 0.81 ± 0.02 1.08 ± 0.02 0.39 ± 0.03 48 neutron 239 detector and PVT Pu*** 28.94 ± 0.02 37.67 ± 0.02 13.72 ±0.02 gamma detector 239 Pu Shield 17.89 ± 0.02 23.84 ± 0.02 8.50 ± 0.03 252Cf* 0.53 ± 0.02 1.59 ± 0.05 2.63 ± 0.04 Radiation Portal HEU** 0.54 ± 0.08 1.64 ± 0.03 2.67 ± 0.04 Monitor with N- 15plus and two HEU Shield 0.31 ± 0.11 0.96 ± 0.06 1.57 ± 0.05 PVT gamma 239 Pu*** 22.11 ± 0.04 22.10 ± 0.04 55.62 ± 0.04 detector 239 Pu Shield 6.78 ± 0.11 21.25 ± 0.06 12.20 ± 0.05

Detector in front the source N-15 “nDetBrick” N-48 N-15 plus *1 ng 252Cf produces 2,340 n/s [Vega-Carrillo 1988] **500g HEU 70% produces 2,144 n/s [Spencer 2011] ***500g 239Pu produces 40,000 n/s [Perani 2012]

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The MCNP results are by emitted neutron, so in each case the results were multiplied by the corresponding intensity source, in 252Cf for nano-gram of californium, HEU 70% for 500 g and 239Pu for 515 g.

Table 4 shows the comparative of each neutron detection system respect to each source, when the vehicle pass in front of them.

Table 4.- Comparative (RPMs) N-15, N-48 and N-15plus N-15 N-48 (right) N-15 plus Source “nDetBrick” Description (one neutron (three neutron conditions (three neutron detector) detectors) detectors) 252Cf* 2.31 ± 0.02 1,79 ± 0.02 2.63 ± 0.04

Comparative HEU** 2.59 ± 0.04 1.80 ± 0.02 2.67 ± 0.04 with three different HEU Shield 1.54 ± 0.05 1.08 ± 0.02 1.57 ± 0.05 systems when the source is in 239 front of each Pu*** 54.24 ± 0.04 37.67 ± 0.02 55.62 ± 0.04

239Pu Shield 33.55 ± 0.05 23.84 ± 0.02 12.20 ± 0.05

Table 5 shows the H*(10) for neutrons, for the driver for each case of positions and materials described in the model.

The MCNP results are by emitted neutron, and are gives in pSv/s, the conversion was made for give the results in µSv/h, in each case the results were multiplied by the corresponding intensity source.

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Table 5.- H*(10)neutrons calculated for each material

252Cf HEU HEU Shield 239Pu 239Pu Shield Section µSv/h µSv/h µSv/h µSv/h n n n n µSv/h (10-4) (10-4) (10-4) (10-4) (10-4) n Head (1) 4.94 ± 0.39 5.21 ±0. 41 2.72 ±0.29 110 .04± 15 59.21 ± 1.2

Neck (1) 4.61 ± 0.69 4.55 ± 0.66 2.44 ± 0.46 97.97 ±2.5 54.20 ± 2.0

Thorax (1) 3.17 ± 0.14 3.34 ± 0.15 1.75 ± 0.10 70.26 ±5.4 39.33 ± 0.45

Lumbar (1) 3.11 ± 0.16 3.29 ± 0.17 1.74 ± 0.21 69.3 ± 6.5 38.01 ± 0.46

Legs (2-2) 12.60 ±1.46 13.06 ± 1.49 6.90 ± 0.10 275.59 ±55 152.25 ± 46

Arms (2) 7.83± 0.11 8.20 ± 0.12 4.36 ± 0.80 172.24 ± 40 95.81 ± 3.3

TOTAL 36.21 ± 1.90 37.65 ± 0.21 19.96 ± 1.4 796.12 ± 75 438.88 ±62

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4.- DISCUSSION

Table 3 shows the calculated response of three neutron detection systems inside a RPM, the N-15, N-48 and N-15plus, in counts per neutron multiplied by the intensity of each source, the neutrons source was located at 200 cm from the detector, in the position 1 the N-15 detector was in the center and we can see than in this case is the more efficient, and the same behavior occur for each RPM when the source is just in front of the neutron detectors in each system, in this table we can see the response of each RPM en all positions.

Table 4, is in summary all response only when the neutron source is just in 200 cm in front and we can see that the N-15plus, RPM, shows the best response, in efficiency, and it can detect a 2.5cps/ng 252Cf as the ANSI required [ANSI 2006], the model propuse of improvement of N-15 can be detect 2.5 cps/ng 252Cf.

Table 5, shows the H*(10) ambient dose equivalent due to neutrons calculated, when the driver transport, a neutron source of 252Cf+HDPE, a Sphere of HEU and 239Pu, bare and shielded by borated polyethylene and lead. In the case that the driver unaware that the box contain some nuclear material mixed in the scrap. The Marjory received dose is when the vehicle transport 239Pu, and in all cases in the legs received the more H*(10).

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5.- CONCLUSIONS

The increased probability of nuclear attack and the scarcity of the elemental material of 3He for the neutron detectors, forced to seek interesting alternatives. In this paper three neutron detectors have been studied in similar situation of operation, these models are based in previous studies [Guzman-Garcia et al., 2016a; Guzman-Garcia et al., 2016b].

The reactions in 10B was calculated using Monte Carlo methods with MCNP6 code, each total response were adjusted with their own efficiency factor determined previously, to determine the counts rates in each case.

The 10B+ZnS(Ag) detectors are an interesting alternative to replace 3He detectors in RPMs. The N- 15 detector is considered suitable for portable backpack systems. The N-48 detector is close to be considered a replacement for 3He detectors in RPM. But the N-15 plus as an improvement in the geometry of the detector raising the amount of 10B and the PMMA- thick, increases the detector efficiency aiming to reach 2.5 cps/ng 252Cf, defined in the ANSI standards a goal to use this type of detectors as an alternative in RPMs [ANSI 2006].

Acknowledgments The first autor, K.A. Guzmán-García, thanks the scholarship granted by the CONACyT and the COZCyT, Mexico.

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REFERENCES

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Pelowitz, Denise B., Fallgren, Andrew J. and McMath, Garret E. (2014). MCNP6TM User's Manual, Code Version 6.1, LA-CP-14-00745, Rev. 0, Los Alamos National Laboratory, Los Alamos, New Mexico, June 2014.

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