X International Symposium/XX National Congress on Solid State Dosimetry “Augusto Moreno y Moreno” September 24-27; 2007. Puebla, Pue. Mexico

SPECTROMETRY AND DOSIMETRY OF A NEUTRON SOURCE

Héctor René Vega-Carrillo*/**, Eduardo Manzanares-Acuña*, Víctor M. Hernández-Dávila*/**, Jaime Ramírez González*, Rubén Hernández Villasana*, Alejandro Chacón Ruiz*/** Unidades Académicas de *Estudios Nucleares e **Ingeniería Eléctrica Universidad Autónoma de Zacatecas C. Ciprés 10, Fracc. La Peñuela, 98068 Zacatecas, Zac. México [email protected]

Resumen

Using Monte Carlo methods the spectrum, dose equivalent and ambient dose equivalent of a 239PuBe at several distances has been determined. Spectrum and both doses, at 100 cm, were determined experimentally using a Bonner sphere spectrometer. These quantities were obtained by unfolding the spectrometer count rates using artificial neural networks. The dose equivalent, based in the ICRP 21 criteria, was measured with the area neutron dosemeter Eberline model NRD, at 100, 200 and 300 cm. All measurements were carried out in an open space to avoid the room return. With these results it was found that this source has a yield of 8.41E(6) n/s.

Keywords: Spectrum, Dose, Neutron source, Monte Carlo

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Vega-Carrillo, H.R. et al. Spectrometry and dosimetry of a neutron source

1. INTRODUCTION

In 1932, Chadwick did show the existence of neutron [1, 2]; since then neutrons have been widely utilized in science and technology.

Neutrons are artificially produced by nuclear fission and nuclear fusion in reactors, using accelerating devices that induce nuclear reactions involving charged particles and gamma-rays, and with alpha and gamma radiations from radioactive nuclides. Also, neutrons can be found in the environment, those produced by the nuclear reactions between the cosmic rays and the nuclei present in the atmosphere, and those induced by nuclear reactions occurring between alpha particles and environmental materials; the alpha particles come from naturally-occurring heavy nuclei decay. [3, 4]

Depending of how neutrons are produced its energy distribution, also named spectrum, ranges from few tenths of eV to several GeV. The knowledge of neutron spectrum is vital for dosimetry, which it is carried out by any of the following categories: time of flight, nuclear reaction, recoil protons and integral systems.

The isotopic neutron sources are small, compact, portable, easy to handle and do not require of high voltage. In these sources neutrons are produced by (γ, n) or (α, n) exoenergetic nuclear reactions. In this group are also those neutron sources produced during the spontaneous fission of some heavy nuclei. [5]

Photoneutron sources are those that produce neutrons using the (γ, n) reaction. During its decay some radioisotopes produce γ-rays that collide with D or Be producing neutrons. The threshold energy for this reaction is 1.67 MeV for Be and 2.33 MeV for D; involved nuclear reactions are 9Be(γ, n)8Be and 2H(γ, n)1H. In these sources the γ-emitters are surrounding with a thick layer of the target nuclei. Nevertheless the photons are monoenergetic the resulting neutron spectrum is spread in energy due to the angle between incident photon and emitted neutron and due to the thickness of target. Some of the radionuclides used to produce such photons are 24Na, 88Y, 124Sb and 140La.

Alphaneutron sources produce neutrons through the (α, n) reaction. When an alpha particle collides with 9Be it is produced following nuclear reaction: 4He+9Be → 1n + 12C* + Q. The Q- value is 5.7 MeV that is shared between the reaction products. The intermediate state of this reaction is the formation of 13C compound nucleus. Neutrons are produced through different channels; those with higher probability are shown in figure 1.

The alphaneutron sources are produced by mixing the 9Be with the α-emitter radioisotope. Nevertheless 9Be is widely used there are other target nuclei that are suitable to produce these neutron sources such as 10B, 7Li, 19F, 13C and 18O. Between those α-emitting nuclei are 210Po, 226Ra, 227Ac, 228Th, 238Pu, 239Pu, 241Am, 242Cm, 244Cm, etc.

Spontaneous fission sources are those neutron sources produced when heavy nuclei disintegrate by spontaneous fission releasing several neutrons in the process. Nevertheless many transuranium elements, like 242Pu, 238U, 244Cm, 252Cf, etc undergo α-decay and suffer spontaneous fission. One

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X International Symposium/XX National Congreso on Solid State Dosimetry of the widely used spontaneous fission neutron source is the 252Cf, which is used bare or 252 moderated with D2O. During Cf decay only 3.2% goes through spontaneous fission shown in figure 2.

Figure 1. Neutron production during (α, n) reaction in 9Be.

Figure 2. Spontaneous fission of 252Cf.

Isotopic neutrons sources are widely utilized in several activities including neutron activation [6], teaching [7], and monitor calibration [8]. For calibration the International Organization for 241 241 252 252 Standarization (ISO), recommend to use AmBe, AmB, Cf and D2O-moderated Cf sources [9]. However, worldwide other isotopic sources are utilized with different purposes.

Calibration, safety protocols and potential applications of neutron sources require knowing the physical features of neutron source, its strength (neutron yield), spectrum, and the effect of laboratory conditions upon neutron spectrum and dosimetric quantities. This information is also useful to design the storage and handling procedures of neutron sources.

The 239PuBe sources containing small quantities of 241Pu tend to increase the source’s neutron yield because 241Pu decays to 241Am that also emits alpha particles. Thus the neutron source will tend to increase its neutron yield with time. The neutron yield of 239PuBe sources with impurities of 241Pu can be estimated using equation 1 [10].

Q = Qo – 7.4E(6) m f [ Exp(-λ1 t) - Exp(-λ2 t)] (1)

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Vega-Carrillo, H.R. et al. Spectrometry and dosimetry of a neutron source

Here, Qo is the neutron emission rate, in neutrons/second, at time of fabrication, m is the mass of 241 -1 Pu in the source, f is the mass fraction of Pu present at fabrication, λ1 = 5.25E(-2) y and λ1 = 1.60E(-3) y-1.

The aim of this work was to use Monte Carlo methods to determine the yield, fluence rate, neutron spectrum, dose equivalent and the ambient equivalent dose due to neutrons of a 239PuBe neutron source; and to compare these results with those obtained with measurements.

2.- MATERIALS AND METHODS

In our facility we have a 239PuBe source that is part of a subcritical nuclear reactor. The nominal original activity of 239Pu was 185 GBq and consists in 3 cylindrical pellets. It was manufactured at 1968 having 79.979 grams of Pu, no information is available about the amount of 241Pu and the source strength.

The cylindrical pellets are together inside an aluminium container, which is handled with a long solid cylindrical-shape piece of lucite.

2.1.- Calculations

A model of the neutron source was built and neutron transport calculations were carried out with MCNP 4C code [11]. The source term used for calculations was obtained from the International Atomic Energy Agency, IAEA, [12]. Neutron spectrum, ΦE(E), at 50, 100, 200 and 300 cm from the source, in air, was calculated.

With the neutron spectra the dose equivalent, H, and the ambient dose equivalent, H*(10), were determined using equation 2.

Emax ∆ = ∫ Φ E ()()E δ E dE (2) E min

Here, ∆ is the dose (H or H*(10)), δ(E) is the fluence-to-dose conversion coefficients for H or H*(10). H and H*(10) were determined under the ICRP21 [13] and ICRP74 criteria [14] respectively.

2.2.- Experiments

Using a Bonner sphere spectrometer (BSS) with a 0.4 Ø × 0.4 cm2 6LiI(Eu) scintillator, the neutron spectra, at 100 cm form the source, was obtained. Polyethylene spheres utilized during measurements were 0, 2, 3, 5, 8, 10 and 12 inches-diameter. From the count rates measured with the BSS the neutron spectrum, total fluence rate, H, and H*(10) were estimated using artificial neural networks technology. [15-17].

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X International Symposium/XX National Congreso on Solid State Dosimetry

H was measured using the area neutron dosemeter (old term neutron remmeter) EBERLINE model NRD.

All measurements were carried out in an open space, with no walls, and at a height above ground of 200 cm in order to avoid the neutron room return [18, 19].

Spectrum, total fluence rate and the dose equivalent obtained experimentally were compared with calculated results. In the aim to match those quantities a correction factor due to the presence of 241Pu was estimated.

3. RESULTS AND DISCUSSION

The neutron spectrum of 239PuBe in air, obtained with the Monte Carlo calculations, is shown in figure 3. Here can be noticed that the in-scatter effect due to the air presence in negligible and neutron spectrum only reduces, following the 1/r2 rule, as the source-to-detector distance increases.

10-4

10-5 239PuBe

10-6 50 cm 100 cm

] 200 cm -1 10-7 u 300 cm ∆ - -2 10-8

10-9 (E) [ cm E

Φ 10-10 E 10-11

10-12

10-13 10-5 10-4 10-3 10-2 10-1 100 101 102

Neutron energy [ MeV ]

Figure 3. Neutron spectrum of 239PuBe at different distances in air.

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Vega-Carrillo, H.R. et al. Spectrometry and dosimetry of a neutron source

In figure 4 are the measured and calculated neutron spectra at 100 cm. Both spectra are alike.

50

239PuBe 40

] Experiment -1 Monte Carlo u ∆

- 30 -1

20 (E) [ cm E Φ

E 10

0 10-5 10-4 10-3 10-2 10-1 100 101 102

Neutron energy [ MeV ]

Figure 4. Measured and calculated neutron spectrum of 239PuBe at 100 cm.

From Monte Carlo calculation, at 100 cm the source strength is 1 n/s, Φ = 9.53883E(-6) n/cm2-s, H = 1.26996E(-6) rem/h and H*(10) = 1.36801E(-5) µSv/h. The corresponding dose-to-fluence conversion factors for 239PuBe are 398 pSv-cm2 for H*(10) and 3.70E(-8) rem-cm2 for H. Both factors are in agreement, in 6% [8] and 4% [20] respectively, with coefficients reported in the literature.

Using the mass of 239Pu and its half life, 24110 ± 30 y [21], the original activity of 239Pu is 1.84E(11) Bq, thirty five years latter this activity is reduced in 0.1% due decay. Using the specific yield factors for 239PuBe sources given by IAEA [22] and the NCRP [10], the neutron source strength of 239PuBe is 7.95E(6) and 7.94E(6) n/s respectively. This neutron yield was utilized with the Monte Carlo results to obtain the neutron fluence rate, H and H*(10) being 76 n/cm2-s, 10.0 mrem/h and 108.6 µSv/h respectively.

At 100 cm the neutron fluence rate, H and H*(10), obtained from the BSS are 81 ± 4 n/cm2-s, 10.6 ± 0.5 mrem/h, and 114.2 ± 5.7 µSv/h respectively; while the H measured at 100 cm with the

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X International Symposium/XX National Congreso on Solid State Dosimetry area neutron dosemeter Eberline NRD is 10.5 ± 0.4 mrem/h. Comparing these results with those derived from Monte Carlo calculation significant differences are noticed.

Probable explanation of the differences is attributed to neutron yield increase due to the presence of 241Pu. To estimate the amount of this isotope we used equation 1 obtaining that a 0.1% mass fraction allows matching these quantities. Doing the proper corrections the source strength is (8.41 ± 0.84)E(6) n/s, at 100 cm the neutron fluence rate is 80 ± 8 n/cm2-s, H is 10.7 ± 1.1 mrem/h and H*(10) is 115 ± 12 µSv/h.

The corrected source strength was also tested by using the Monte Carlo calculation for H and the corresponding experimental values measured at 200 and 300 cm, this comparison is shown in figure 5, where an overall 10% uncertainty was given to quantities derived from Monte Carlo calculations.

12

10 Calculated Measured

8

6

H [ mrem/h ] mrem/h [ H 4

2

0 50 100 150 200 250 300 350

Source-to-detector distance [ cm ]

Figure 5. Calculated and measured dose equivalent of 239PuBe at 100, 200 and 300 cm.

By comparing the dose equivalent calculated at 100, 200 and 300 cm with those measured can be noticed that they agree well, therefore the source strength is is 8.41E(6) n/s. With this value any other Monte Carlo calculation, where the neutron source can be perturbed by materials around it, can be performed.

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Vega-Carrillo, H.R. et al. Spectrometry and dosimetry of a neutron source

4. CONCLUSIONS

Quantities like source strength, spectrum, fluence rate and ambient dose equivalent are required to characterize a neutron source. Those quantities have been determined for a 239PuBe neutron source by means of Monte Carlo calculations and experiments.

During manufacturing alphaneutron sources could have impurities that along its radioactive decay might end in α-emitter isotopes that are added to the main α-emitter radionuclide. In the case o a 239PuBe source the presence of small amounts of 241Pu isotope decaying by β-emission ends in 241Am that decays emitting α particles. The build up of 241Am produce an increase in the 239PuBe source.

By Monte Carlo calculations, where a detail model of the neutron source was utilized, the neutron spectrum, fluence rate, H and H*(10) were determined. From this calculations the dose- to-fluence conversion factors are 3.70E(-8) rem-cm2 and 398 pSv-cm2.

The 239PuBe neutron spectra at 50, 100, 200, and 300 cm were calculated using MCNP 4C. Calculated neutron spectrum at 100 cm agrees well with the spectrum obtained with BSS.

The dose equivalent was measured at 100 cm with an area neutron dosemeter being 10.5 ± 0.4 mrem/h. This was utilized with the Monte Carlo results to obtain the neutron yield, from this it was estimated that this source had 0.1% mass fraction of 241Pu.

Corrections due to the 241Pu presence were carried out and corrected values for source strength, fluence rate, H and H*(10) were estimated. These quantities compares well with those measured with BSS. Thus the neutron yield is 8.41E(6) n/s. At 100 cm the neutron fluence is 80 n/cm2-s, H is 10.7 mrem/h and H*(10) is 115 µSv/h that compare well with 81 n/cm2-s, 10.6 mrem/h and 114.2 µSv/h obtained with BSS.

Verification was extended by comparing the Monte Carlo derived values for H at 200 and 300 cm with those measured with the area neutron dosemeter.

AKNOWLEDGMENTS

This work is part of research project SYNAPSIS, partially supported by CONACYT under contract SEP-2005-C01-46893.

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REFERENCES

1. Chadwick, J. (1932). Possible existence of a neutron. Nature 192: 312. 2. Chadwick, J. (1932). The existence of a neutron. Proceedings of the Royal Society of London A 136: 692-708. 3. Vega-Carrillo, H.R. and Manzanares-Acuña, E. (2004). Background neutron spectrum at 2420 m above sea level. Nuclear Instruments and Methods in Physics Research A 524(1- 3): 146-151. 4. Vega-Carrillo, H.R. and Torres-Muhech, C. (2002). Low energy neutrons from a 239PuBe isotopic neutron source inserted in moderating media. Revista Mexicana de Física 48(5): 405-412. 5. Garg, A.N. and Batra, R.J. (1986). Isotopic neutron sources in neutron activation analysis. Journal of Radioanalytical and Nuclear Chemistry, Articles 98(1): 167-194. 6. Naqvi,A.A. and Al-Nagadi, M.M. (2004). Performance comparison of an 241Am-Be neutron source-based PGNAA setup with the KFUPM PGNAA setup. Journal of Radioanalytical and Nuclear Chemistry 260(3): 641-646. 7. DuBard, J.L. and Gambhir, A. (1994). Dust off the neutron howitzer to teach nuclear physics!. American Journal of Physics 62(3): 255-257. 8. Józefowickz, K., Golnik, N., Tulik, P. and Zielczynski, M. (2007). Radionuclide neutron sources in calibration laboratory-Neutron and gamma doses and their changes in time. Radiation Protection Dosimetry. (doi:10:1093/rpd/ncm029). 9. ISO. (2001). Reference neutron radiations. Characteristics and methods of production. International Organization for Standarization. ISO-Standard 8529, Part1. 10. NCRP. (1991). Calibration on survey instruments used in radiation protection for the assessment of ionizing radiation field and radioactive surface contamination. National Council on Radiation Protection and Measurements Report No.112. Bethesda MD USA. 11. Briesmeister, J.F. (Editor). (2000). MCNPTM–A general Monte Carlo N-particle transport code. Version 4C. Los Alamos Nacional Laboratory Report LA-13709-M. 12. Griffith, R.V., Palfalvi, J. and Madhvanath, U. (1990). Compendium of neutron spectra and detector responses for radiation protection purposes. International Atomic Energy Agency Technical Report Series No. 318. Vienna. 13. ICRP. (1971). Data for protection against ionizing radiation from external sources: Supplement ICRP publication 15. International Commission on Radiation Protection Publication 21.

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14. ICRP. (1996). Conversion coefficients for use in radiological protection against external radiation. International Commission on Radiological Protection publication 74. Annals of the ICRP 26(3/4). 15. Vega-Carrillo, H.R., Manzanares-Acuña, E., Hernandez-Davila, V.M., Barquero, R., Iñiguez, M.P., Mendez, R., Palacios, F., Arteaga-Arteaga, T. and Ortiz-Rodriguez, J.M. (2006). Neutron spectrometry using artificial neural networks. Radiation Measurements 41: 425-431. 16. Vega-Carrillo, H.R., Hernandez-Davila, V.M., Manzanares-Acuña, E., Gallego, E., Lorente, A. and Iñiguez, M.P. (2007). Artificial neural networks technology for neutron spectrometry and dosimetry. Radiation Protection Dosimetry. (doi: 10.1093/rpd/ncm084). 17. Vega-Carrillo, H.R., Hernandez-Davila, V.M., Manzanares-Acuña, E., Mercado, S., G.A., Gallego, E., Lorente, A., Perales-Muños, W.A. and Robles-Rodriguez, J.A. (2006). Artificial neural networks in neutron dosimetry. Radiation Protection Dosimetry 118(3): 251-259. 18. Vega-Carrillo, H.R., Manzanares-Acuña, E., Iñiguez, M.P., Gallego, E. and Lorente, A. (2007). Study of room-return neutrons. Radiation Measurements 42(3): 413-419. 19. Vega-Carrillo, H.R., Manzanares-Acuña, E., Iñiguez, M.P., Gallego, E. and Lorente, A. (2007). Spectrum of isotopic neutron sources inside concrete wall spherical cavities. Radiation Measurements. (doi: 10.1016/j.radmeas.2007.06.006).

20. Anderson, M.E. and Neff, R.A. (1972). Neutron spectra of different size 239Pu-Be(α, n) sources. Nuclear Instruments and Methods 99: 231-235. 21. BNL. [On line]. Brookhaven National Laboratoy. National Nuclear Data Center. NuDat 2.3. < http://www.nndc.bnl.gov/nudat2/ >. [Consult: July 2007]. 22. Hoste, J. (1988). Isotopic neutron sources for neutron activation analysis. International Atomic Energy Agency. IAEA-TECDOC-465. Vienna.

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