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Development of new optically stimulated luminescence (OSL) neutron dosimeters
Radiation Measurements 43 (2008) 309 – 314 www.elsevier.com/locate/radmeas
Invited paper
Development of new optically stimulated luminescence (OSL) neutron dosimeters E.G. Yukihara a,∗ , J.C. Mittani a , F. Vanhavere b , M.S. Akselrod c a Department of Physics, Oklahoma State University, 145 Physical Sciences II, Stillwater, USA b SCK-CEN, Belgium Nuclear Research Centre, Boeretang 200, 2400 Mol, Belgium c Landuaer Stillwater Crystal Growth Division, 723 1/2 Eastgate, Stillwater, Oklahoma, USA
Abstract This paper demonstrates the possibility of using neutron converters incorporated into Al2 O3 : C dosimeters to produce new neutron-sensitive OSL dosimeters suitable for personal monitoring. A comparison between different neutron converters indicate that lithium-based compounds (6 LiF and 6 Li2 CO3 ) result in the best neutron sensitivity, probably because of the relatively large range of the 2.75 MeV triton produced by the 6 Li(n, )3 H neutron capture reaction. Based on these results, a new neutron-sensitive OSL composite material, identical to the one used in the Luxel姠 and InLight姠 dosimetry systems (Landauer Inc.) except for the smaller gamma sensitivity and enhanced neutron sensitivity, was produced and characterized. The gamma sensitivity of the new material is ∼ 35% the sensitivity of the regular Al2 O3 : C material (Luxel姠 ), due to reduced amount of Al2 O3 : C in its composition. The neutron sensitivity achieved in this new material is ∼ 60% of the neutron sensitivity of 6 LiF:Mg,Ti (TLD-600). © 2007 Elsevier Ltd. All rights reserved. Keywords: Neutron dosimetry; Optically stimulated luminescence; Neutron converters
1. Introduction In spite of the successful use of the optically stimulated luminescence (OSL) technique in personal dosimetry of photons (X-ray and gamma) and beta rays, the lack of a neutron sensitive OSL dosimeter prevents the application of the technique to neutron fields, representing the main disadvantage of OSL when compared to the thermoluminescence (TL) technique (McKeever and Moscovitch, 2003). Carbon-doped aluminium oxide (Al2 O3 : C), considered the reference OSL material in dosimetry because of its high sensitive and optical properties, has a neutron sensitivity lower than 7 LiF:Mg,Ti (TLD-700) (Klemic et al., 1996). Today, the OSL technique is the basis for the Luxel姠 and InLight姠 dosimetry systems (Landauer Inc.) (BZtter-Jensen et al., 2003) and 1.5 million users are monitored using the OSL technology. The introduction of a neutron sensitive OSL dosimeter would extend the advantages of the OSL technique, ∗ Corresponding author. Tel.: +1 405 744 5051; fax: +1 405 744 6811.
E-mail address: [email protected] (E.G. Yukihara). 1350-4487/$ - see front matter © 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.radmeas.2007.10.005
such as precise control of stimulation, fast readout that results in high productivity, wide range of linearity and high sensitivity, and possibility of re-estimation of absorbed dose (Akselrod and McKeever, 1999), to personal dosimetry in neutron fields. Neutron detection relies on secondary charged particles and photons created by neutron interaction with the detector itself or surrounding material. Elastic neutron scattering with hydrogen nuclei create recoil protons with energies comparable to the incident neutrons, whereas nuclear capture reactions such as the ones exemplified in Table 1 create various types of charged particles with energies of a few MeV, gamma rays, and conversion electrons. These secondary radiation deposit energy in the detector giving rise to a neutron induced signal. Thermoluminescence detectors (TLD) have been used for decades in personal neutron dosimetry. Materials such as LiF:Mg,Ti, LiF:Mg,Cu,P and Li2 B4 O7 : Mn can be prepared with higher concentration of 6 Li or 10 B isotopes, which have high cross-section for neutron capture (McKeever et al., 1995). Because the cross-section decreases with increasing neutron energies, these dosimeters are mostly sensitive to thermal and epithermal neutrons and, therefore, are used to detect the
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Table 1 Examples of isotopes of interest for neutron detection, their relative abundance, neutron capture cross-section for thermal neutrons, and products (Knoll, 2000; van Eijk, 2004) Isotope
Natural abundance (%)
(barns)
Products
6 Li
7.4
940
3H
(2.75 MeV) + 4 He (2.05 MeV)
10 B
19.8
3840
7 Li
(1.0 MeV) + 4 He (1.8 MeV) (0.83 MeV) + 4 He (1.47 MeV) + (0.48 MeV)
7 Li 157 Gd
15.7
255,000
158 Gd
+s + conversion e− + X-rays (29.182 keV)
155 Gd
14.8
60,900
156 Gd
+s + conversion e− + X-rays (39.199 keV)
neutrons scattered by the user’s body in the so-called albedo configuration (Piesch and Burgkhardt, 1985). The possibility of performing neutron imaging with the OSL technique using Al2 O3 : C sheets and Gd2 O3 neutron converters was recently demonstrated (Kobayashi et al., 2005), which hints to the possibility of using neutron converters to create a new OSL dosimeter based on Al2 O3 : C. This approach has the potential of shortening the lead time for the introduction of a useful neutron-sensitive OSL dosimeter in several years, since such a neutron dosimeter would have all the properties of Al2 O3 : C (optical sensitivity, absence of fading, stimulation and emission spectrum, etc.) with the desired neutron sensitivity. The question is whether such approach can produce a dosimeter with sensitivity high enough for personal dosimetry or not. This paper demonstrates the possibility of using neutron converters to produce new OSL dosimeters with neutron sensitivity suitable for personal dosimetry. We investigated the effect of different neutron converters and the size of Al2 O3 : C grains on the neutron sensitivity. Based on these results, we produced a new neutron-sensitive OSL composite material identical to the one used in the Luxel姠 and InLight姠 dosimetry systems (Landauer Inc.), except for the lower gamma sensitivity and enhanced neutron sensitivity. The gamma and neutron sensitivity of these new materials are compared to regular Al2 O3 : C dosimeter (Luxel姠 ) and 6 LiF:Mg,Ti (TLD-600). 2. Materials and methods 2.1. Samples The dosimeters used in this study were prepared with a powder mixture of Al2 O3 : C and one of the following neutron converters: lithium fluoride enriched with 6 Li (6 LiF), lithium carbonate enriched with 6 Li (6 Li2 CO3 ), gadolinium oxide (Gd2 O3 ), boric acid enriched with 10 B (H3 10 BO3 ), and high-density polyethylene (HDPE). Preliminary investigations on the performance of the neutron converters were performed using loose powder and cold-pressed pellets. Based on these preliminary results, a neutron sensitive composite material consisting of a thin layer of Al2 O3 : C (grain size < 38 m) and 6 LiF powder deposited in a clear polyester film, hereafter designated N-OSL tape, was manufactured and tested. This tape is the first version of a neutron sensitive OSL
dosimeter produced by the same processes used in the production of the regular Al2 O3 : C tape used in the Luxel姠 and InLight姠 badges (Landauer Inc.). In addition, 3×3×0.9 mm TLD chips of LiF:Mg,Ti enriched with 6 Li and 7 Li (TLD-600 and TLD-700, ThermoElectron Inc.) were used for comparison. 2.2. Radiation sources Preliminary neutron irradiations were carried out using an uncalibrated Pu–Be source to compare the sensitivity of different samples. To precisely determine the neutron sensitivity, irradiations were carried out at the SCK-CEN calibration room using a calibrated bare 252 Cf source (Vanhavere et al., 2001). All irradiations were carried out with the dosimeters mounted on a 30 × 30 × 15 cm PMMA phantom on the side facing the source (for more details, see Mittani et al., 2007). Irradiations with a shadow cone placed between the source and the phantom were carried out to estimate the contribution from room scattering on the results (ISO 8529-2, 2000). The gamma sensitivity of the dosimeters was obtained using a calibrated 60 Co source from SCK-CEN. Additional irradiations were carried out using a calibrated 90 Sr/90 Y beta source at OSU. 2.3. OSL and TL readouts OSL measurements were carried out using two types of readers: RisZ TL/OSL-DA-15 readers and an in-house built portable reader. Continuous-wave OSL (CW-OSL) were performed using the RisZ TL/OSL readers, green or blue LEDs for stimulation, and Hoya U-340 filters (transmission between 290 and 370 nm) in front of the photomultiplier tube (PMT) (BZtter-Jensen et al., 2000). Pulsed-OSL measurements were carried out using the in-house built portable reader, green LEDs for stimulation, and Corning 5-58 filters (transmission between 380 and 440 nm) in front of the PMT. It should be pointed out that the results obtained with the two readers are not directly comparable, since the Hoya U-340 filters transmits the UV emission band observed at 335 nm in addition to the main luminescence band of Al2 O3 : C, centered at 420 nm (Yukihara and McKeever, 2006b), whereas the Corning 5-58 filters block the UV emission. The relative intensity of the UV and F-center emission bands can vary with the radiation field (Yukihara and McKeever, 2006a; Yukihara et al., 2006).
E.G. Yukihara et al. / Radiation Measurements 43 (2008) 309 – 314
TL measurements of TLD-600 and TLD-700 dosimeters were carried out using the RisZ TL/OSL reader and Schott BG-39 filters (6 mm total thickness).
7
2.4. Data analysis
5
The OSL intensity over 600 s of stimulation was used as the signal S. The dosimeter’s response was calculated from the OSL signal S using the relationship:
4
R = Rn + R .
(2)
The gamma component R was estimated using the neutron insensitive dosimeters (Al2 O3 : C or TLD-700). The neutron response was also assumed to have two components, one due to the direct neutrons from the source and albedo neutrons scattered by the phantom (R0 ), plus the response due to neutrons scattered by the room (Rs ): Rn = R0 + Rs .
Rn/Rγ
(1)
where the calibration factor a was obtained by irradiation with known gamma doses. Therefore, the response of the dosimeter is the gamma dose required to produce an OSL signal of the same intensity as S. The neutron and gamma response of the dosimeters was separated using the dosimeter pair method. The response of the neutron sensitive detectors (e.g., Al2 O3 : C+ neutron converter or TLD-600) was assumed to have a neutron and a gamma component:
(3)
The contribution from room scattering Rs was estimated using dosimeters irradiated in the same conditions as before, but with a shadow cone mounted between the source and the phantom to block the neutrons coming directly from the source (ISO 8529-2, 2000). The neutron response obtained using the dosimeters irradiated with the shadow cone provide an estimate for Rs . The neutron sensitivity of the dosimeters was defined as the ratio between the neutron response and the reference neutron dose. 3. Results
6LiF
Gd2O3 6Li CO 2 3
6
H310BO3
3 2 1 0 0.0
0.5 1.0 1.5 2.0 2.5 mass of neutron converter/mass of Al2O3:C
3.0
Fig. 1. Comparison of the neutron sensitivity of mixtures of Al2 O3 : C with different neutron converters, expressed by the ratio between the neutron and gamma response for a 2 h exposure to a Pu–Be source neutron-to-gamma ratio Rn /R , as a function of the neutron converter to Al2 O3 : C mass ratio. The data was obtained using loose powder, and the OSL measurements were carried out using the RisZ reader, green stimulation, and Hoya U-340 filters in front of the PMT.
35 30 Response (mGy)
R = aS,
311
25 20 15 10 5 0
Fig. 2. Response of different samples exposed to a nominal neutron dose of 30 mSv of a bare 252 Cf source. The samples were irradiated on the face of a 30 × 30 × 15 cm PMMA phantom. The Al2 O3 : C+ neutron converter dosimeters were cold-pressed pellets, and the OSL measurements were carried out using the RisZ reader, green stimulation, and Hoya U-340 filters in front of the PMT.
3.1. Comparison of neutron converters Preliminary investigations were carried out to compare the performance achieved using different neutron converters. Aliquots of Al2 O3 : C powder mixed with different neutron converters were exposed to the Pu–Be source and the OSL signal was measured using the RisZ reader. Pure Al2 O3 : C was used to estimate the gamma contribution of the field, and the neutron-to-gamma ratio Rn /R was calculated as discussed in Section 2.4. Fig. 1 shows the neutron-to-gamma ratio Rn /R as a function of the mass ratio between neutron converter and Al2 O3 : C
in the mixture. Rn /R increases steadily with increasing proportion of neutron converter in the mixture. The largest neutron sensitivities were obtained using 6 LiF and 6 Li2 CO3 , followed by Gd2 O3 and H3 10 BO3 . The neutron sensitivity of the mixture containing 6 Li2 CO3 is approximately 70% of the neutron sensitivity of the mixture containing 6 LiF. This is caused by the fact that the number of Li atoms per mass of compound of Li2 CO3 is ∼ 70% the value of LiF. The neutron sensitivity obtained with H3 10 BO3 was very small, probably because the secondary particles produced by the
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10 B(n, )7 Li
reaction ( and 7 Li) have small range in Al2 O3 : C (a few microns or less). The high sensitivity obtained with lithium compounds is probably related to the triton produced in the 6 Li(n, )3 H reaction, which has a range of ∼ 20 m in Al2 O3 : C (considering an energy of 2.75 MeV). Based on these results, H3 10 BO3 was not included in further investigations. Although the data in Fig. 1 provides a comparison between the neutron sensitivity of the different samples, it does not show that the overall sensitivity obviously decreases with decreasing amounts of Al2 O3 : C in the mixture. An optimum compromise between neutron sensitivity and overall sensitivity was found for a neutron converter/Al2 O3 : C mass ratio between 0.6 and 1 (Mittani et al., 2007). To determine the neutron sensitivity achieved with various neutron converters, cold pressed pellets of Al2 O3 : C and neutron converter mixed in a 1:1 mass proportion were prepared
Al2O3:C grain sizes Regular <38 μm
3.5 3.0
Rn /Rγ
2.5 2.0 1.5 1.0 0.5 0.0 6Li
2F
6Li
2CO3
Gd2O3
Neutron converter Fig. 3. Comparison of the neutron sensitivity of mixtures of Al2 O3 : C with different neutron converters, expressed by the neutron-to-gamma ratio Rn /R for a 2 h exposure to a Pu–Be source. The data was obtained using OSL tapes produced in laboratory using Al2 O3 : C with either regular grain sizes (10.100 m) or small grain sizes (< 38 m). The OSL measurements were carried out using the RisZ reader, blue stimulation, and Hoya U-340 filters in front of the PMT.
and irradiated using the 252 Cf source. Single crystal Al2 O3 : C, regular Al2 O3 : C tape (Luxel姠 ), and TLDs were also irradiated in the same conditions for comparison. The responses of the various materials irradiated with a nominal neutron dose of 30 mSv are compared in Fig. 2. The response of the neutron insensitive dosimeters, Al2 O3 : C single crystals, Al2 O3 : C Luxel姠 , and TLD-700, varies between 1.96 and 2.31 mGy, giving an estimation for the gamma component. The highest response was from TLD-600 (34.9 mGy), followed by Al2 O3 : C+ 6 LiF (20.1 mGy). The response of Al2 O3 : C+ 6 Li CO is again approximately 70% of the response of the 3 2 Al2 O3 : C+ 6 LiF. In the case of the Al2 O3 : C + Gd2 O3 mixture the response was 6.3 mGy, which is higher than the gamma contribution estimated using Al2 O3 : C or TLD-700 dosimeters. However, the increased response is partly caused by the high atomic number of Gd and the resultant increased gamma sensitivity. The gamma contribution in this case needs to be estimated using the Al2 O3 : C + Yb2 O3 mixture, which has approximately the same effective atomic number but is not neutron sensitive (Pomerance, 1951). As expected, the response of Al2 O3 : C + Yb2 O3 mixture (3.7 mGy) was significantly higher than the response from the other gamma sensitive dosimeters. This preliminary investigation suggests that compounds containing 6 Li are the most promising neutron converters to develop the new OSL neutron dosimeters. The effect of 6 LiF and 6 Li CO neutron converters seems to differ only with respect 3 2 to the number of Li atoms per mass of compound. Another reason for choosing these neutrons converters is the fact that they have a low atomic number and do not alter the dosimeter’s sensitivity and effective atomic number as much as the Gd2 O3 neutron converters. The complete data on response and sensitivity of the various neutron converters can be found in Mittani et al. (2007). 3.2. Optimizing the neutron sensitivity Although the investigations in the previous section demonstrate the possibility of using neutron converters to produce neutron sensitive OSL dosimeters based on Al2 O3 : C, the
Fig. 4. Neutron sensitive OSL tape produced using a mixture of Al2 O3 : C and 6 LiF neutron converter.
E.G. Yukihara et al. / Radiation Measurements 43 (2008) 309 – 314
Table 2 Neutron response of the TLD pair (TLD-600 and TLD-700) and OSLD pair (new and regular OSL strip) to a nominal neutron dose of 30 mSv from a 252 Cf source
12 Gamma irradiation RegularAl2O3:C tape (LuxelTM) N-OSLtape
S - Sc (106 counts)
10
Dosimeter pair Rn (mSv)a (Dref = 45.6 mSv)
8
TLDs OSLDs
6
33.1 ± 1.3 20.7 ± 0.4
Rs (mSv)b (Dref = 15.6 mSv)
R0 (mSv)c (Dref = 30 mSv)
23.6 ± 0.7 14.3 ± 0.6
9.5 ± 1.5 6.4 ± 0.7
The irradiations were carried out with the dosimeters on a phantom at 1.5 m from the source. The uncertainties are the propagated values based on the readout of three dosimeters of each type. a Response to the total field (direct neutrons, albedo neutrons scattered from the phantom, and neutrons scattered by the room). b Response to the room scattering estimated using a shadow cone placed between the source and the phantom. c Response only to the source and to the albedo neutrons scattered by the phantom, obtained by R0 = Rn − Rs .
4
2
0 0
45
10
20
30 40 Hp(10) (mSv)
50
60
the grains which is not sensitive to neutrons, only to gammas. Using small grain sizes, it is possible to decrease the inactive volume and increase the neutron sensitivity compared to the gamma sensitivity. The effect of reducing the grain size in Al2 O3 : C powder was investigated using tapes prepared in laboratory using regular and fine Al2 O3 : C powder mixture mixed in a 1:1 mass ratio. Fig. 3 shows the neutron-to-gamma ratio for samples prepared using neutron converters and Al2 O3 : C powder of either regular grain sizes or grain sizes less than 38 m. Larger neutron-togamma ratios can be obtained using fine Al2 O3 : C powder in the case of 6 LiF and 6 Li2 CO3 . In the case of Gd2 O3 , no difference was observed as expected from the fact that the energy is deposited by conversion electrons and gammas, which irradiate the Al2 O3 : C grains more uniformly.
Neutron irradiation TLD-600 TLD-700 N-OSL tape
40 35 R - Rc (mSv)
313
Regular Al2O3:C tape (LuxelTM)
30 25 20 15 10 5 0 0
5
10
15 20 Hp(10) (mSv)
25
30
35
Fig. 5. (a) Gamma response of regular OSL tape and new neutron sensitive OSL tape, expressed by the total OSL signal S minus signal from control dosimeters Sc , as a function of personal dose equivalent Hp (10) for 60 Co irradiation. (b) Response of various dosimeters as a function of nominal personal dose equivalent Hp (10) for 252 Cf irradiation. The response includes the neutron and gamma components of the radiation field. The irradiations were carried out on a 30 × 30 × 15 cm PMMA phantom, at 1.5 m from the radiation source. The OSL measurements were carried out using the portable POSL system, with green stimulation, and Corning 5-58 filters in front of the PMT.
production of OSL tapes involves additional processes that degrade the neutron sensitivity. Improvements in the Al2 O3 : C+ neutron converter mixture were made to counteract some of these effects. One of these improvements is the use of Al2 O3 : C in fine powder. The grain sizes of the regular Al2 O3 : C powder used in the production of the standard Al2 O3 : C OSL tape range from 10 to 100 m. However, the 3 H produced by the 6 Li(n,)3 H neutron capture reaction have limited range in Al2 O3 : C. As a result, for large grains there may be a volume in the center of
3.3. Characteristic of new OSL neutron dosimeter Based on the information obtained in the previous sections, the N-OSL tape shown in Fig. 4 was fabricated using a mixture of Al2 O3 : C fine powder (< 38 m) and 6 LiF. Samples were prepared from this tape and irradiated with neutrons alongside with samples of regular Al2 O3 : C tape (Luxel姠 ) using the 252 Cf source at SCK-CEN. Samples from the new and regular OSL tape were irradiated with various doses using the 60 Co source to compare the sensitivity and linearity of the dosimeters. Fig. 5a shows the gamma response of both samples. The response of the new OSL strip is approximately 35% of the sensitivity of the regular Al2 O3 : C tape. This results gives an idea of the reduction in sensitivity caused by the reduced mass of Al2 O3 : C in the new OSL tape. However, the number is only a rough estimate, since the regular and neutron sensitive OSL tape were produced in different periods using Al2 O3 : C powder from different batches. To determine the neutron response, the samples were irradiated with various neutron doses using the 252 Cf source. TLDs were also irradiated for comparison. Fig. 5b shows the response of the various dosimeters as a function of the nominal neutron dose. The response of the regular Al2 O3 : C strip and TLD-700
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E.G. Yukihara et al. / Radiation Measurements 43 (2008) 309 – 314
Table 3 Neutron sensitivities of the TLD pair (TLD-600 and TLD-700) and OSLD pair (new and regular OSL strip) to a 252 Cf source Dosimeter pair
Rn /Dref a
Rs /Dref b
R0 /Dref c
TLDs OSLDs
0.727 ± 0.015 0.453 ± 0.008
1.48 ± 0.06 0.96 ± 0.03
0.337 ± 0.014 0.191 ± 0.006
The irradiations were carried out with the dosimeters on a phantom at 1.5 m from the source. The values and propagated uncertainties were calculated based on the responses shown in Error! Reference source not found.b, plus two additional irradiations with nominal neutron doses of 10 and 30 mSv with a shadow cone in between the source and the phantom. a Neutron sensitivity to the total neutron spectrum (direct neutrons, albedo neutrons scattered from the phantom, and neutrons scattered by the room). b Neutron sensitivity to the room scattering spectrum, which in these conditions contributes with ∼ 33% of the total neutron dose (Vanhavere et al., 2001). c Neutron sensitivity to the direct and albedo neutron spectrum.
are between 8.6% and 9.8% of the nominal neutron dose, indicating the gamma contribution of the radiation field. The response of the new OSL tape is linear and approximately 60% of the TLD-600 dosimeter’s response. Based on the data shown in Fig. 5, the total neutron response Rn , response to the neutrons scattered by the room, and response to the direct neutrons from the source and albedo neutrons scattered by the phantom R0 were calculated for the OSL and TL dosimeters as described in Section 2.4. The results for a nominal dose of 30 mSv are shown in Table 2. The neutron response of the OSLDs is ∼ 60% of the response of the TLDs in all cases. The neutron sensitivities of the TLDs and OSLDs shown in Table 3 were obtained dividing the neutron response to the respective reference neutron dose. 4. Conclusions The results presented in this paper demonstrate the possibility of producing new OSL dosimeters using a mixture of Al2 O3 : C and 6 LiF or 6 Li2 CO3 neutron converters. Higher neutron sensitivity of 6 Li compounds in comparison with 10 B converters seems to be related to the large range in Al2 O3 : C of the secondary 3 H particles, produced as a result of the 6 Li(n, )3 H neutron capture reaction. Other neutron converters resulted in lower neutron sensitivity and, in the case of Gd2 O3 , altered gamma sensitivity that requires proper compensation. Based on the investigations on the various neutron converters, a new neutron sensitive OSL tape made of Al2 O3 : C and 6 LiF was produced. This new OSL tape is identical to the regular Al2 O3 : C tape used in the Luxel姠 and InLight姠 badges, except for the gamma and neutron sensitivity. The gamma sensitivity was smaller (∼ 35%) of the regular tape’s sensitivity due to the decreased amount of Al2 O3 : C. However, the neutron response was significantly enhanced, reaching a value of ∼ 60% of the TLD-600 neutron sensitivity. The new OSL dosimeters developed in this study have the advantage of using the same reaction employed by the TLD-600 dosimeters, since both are based on the 6 Li(n, )3 H neutron
capture reaction. As a result, the neutron energy response and behavior of these new OSL dosimeters should be identical to the TLD-600 dosimeter, making all results on neutron dosimetry using TLD-600 immediately applicable to the neutron sensitive OSLDs. Although this approach still requires the use of the dosimeters in albedo configuration, the vast experience gained in albedo dosimetry using TLDs can be put to use with the new OSL dosimeters, with all the advantages of the OSL technique. Acknowledgments The authors would like to thank Bart Marlein and Ludo Melis for assistance with the irradiations at SCK-CEN. This work was supported by Landauer Inc. References Akselrod, M.S., McKeever, S.W.S., 1999. A radiation dosimetry method using pulsed optically stimulated luminescence. Radiat. Prot. Dosim. 81, 167–176. BZtter-Jensen, L., Bulur, E., Duller, G.A.T., Murray, A.S., 2000. Advances in luminescence instrument systems. Radiat. Meas. 32, 523–528. BZtter-Jensen, L., McKeever, S.W.S., Wintle, A.G., 2003. Optically Stimulated Luminescence Dosimetry. Elsevier, Amsterdam. ISO 8529-2, 2000. Reference neutron irradiations—part 2: calibration fundamentals related to basic quantities characterizing the radiation field. International Organization for Standardization. Klemic, G.A., Azziz, N., Marino, S.A., 1996. The neutron response of Al2 O3 : C, 7 Li:Mg,Cu,P, and 7 LiF:Mg,Ti TLDs. Radiat. Prot. Dosim. 65 (1–4), 221–226. Knoll, G.F., 2000. Radiation Detection and Measurements. Wiley, New York. Kobayashi, H., Satoh, M., Kobayashi, I., Morishima, H., 2005. Neutron imaging using an optically stimulated luminescence material: -Al2 O3 : C+ Gd2 O3 . IEEE Trans. Nucl. Sci. 52 (1), 360–363. McKeever, S.W.S., Moscovitch, M., 2003. On the advantages and disadvantages of optically stimulated luminescence dosimetry and thermoluminescence dosimetry. Radiat. Prot. Dosim. 104, 263–270. McKeever, S.W.S., Moscovitch, M., Townsend, P.D., 1995. Thermoluminescence Dosimetry Materials: Properties and Uses. Nuclear Technology Publishing, Ashford. Mittani, J.C.R., da Silva, A.A.R., Vanhavere, F., Akselrod, M.S., Yukihara, E.G., 2007. Investigation of neutron converters for production of optically stimulated luminescence (OSL) neutron dosimeters using Al2 O3 : C. Nucl. Instrum. Methods Phys. Res. B 260, 663–671. Piesch, E., Burgkhardt, B., 1985. Albedo neutron dosimetry. Radiat. Prot. Dosim. 10 (1–4), 175–188. Pomerance, H., 1951. Thermal neutron capture cross sections. Phys. Rev. 83 (3), 641–645. van Eijk, C.W.E., 2004. Neutron detection and neutron dosimetry. Radiat. Prot. Dosim. 110 (1–4), 5–13. Vanhavere, F., Coeck, M., Vermeersch, F., 2001. Procedures for neutron scattering corrections in a calibration facility with a non-symmetric set-up. Radiat. Prot. Dosim. 93 (1), 5–10. Yukihara, E.G., McKeever, S.W.S., 2006a. Ionization density dependence of the optically and thermally stimulated luminescence from Al2 O3 : C. Radiat. Prot. Dosim. 119, 206–217. Yukihara, E.G., McKeever, S.W.S., 2006b. Spectroscopy and optically stimulated luminescence of Al2 O3 : C using time-resolved measurements. J. Appl. Phys. 100 (8), 083512. Yukihara, E.G., Sawakuchi, G.O., Guduru, S., McKeever, S.W.S., Gaza, R., Benton, E.R., et al., 2006. Application of optically stimulated luminescence (OSL) technique in space dosimetry. Radiat. Meas. 41, 1126–1135.
Invited paper
Development of new optically stimulated luminescence (OSL) neutron dosimeters E.G. Yukihara a,∗ , J.C. Mittani a , F. Vanhavere b , M.S. Akselrod c a Department of Physics, Oklahoma State University, 145 Physical Sciences II, Stillwater, USA b SCK-CEN, Belgium Nuclear Research Centre, Boeretang 200, 2400 Mol, Belgium c Landuaer Stillwater Crystal Growth Division, 723 1/2 Eastgate, Stillwater, Oklahoma, USA
Abstract This paper demonstrates the possibility of using neutron converters incorporated into Al2 O3 : C dosimeters to produce new neutron-sensitive OSL dosimeters suitable for personal monitoring. A comparison between different neutron converters indicate that lithium-based compounds (6 LiF and 6 Li2 CO3 ) result in the best neutron sensitivity, probably because of the relatively large range of the 2.75 MeV triton produced by the 6 Li(n, )3 H neutron capture reaction. Based on these results, a new neutron-sensitive OSL composite material, identical to the one used in the Luxel姠 and InLight姠 dosimetry systems (Landauer Inc.) except for the smaller gamma sensitivity and enhanced neutron sensitivity, was produced and characterized. The gamma sensitivity of the new material is ∼ 35% the sensitivity of the regular Al2 O3 : C material (Luxel姠 ), due to reduced amount of Al2 O3 : C in its composition. The neutron sensitivity achieved in this new material is ∼ 60% of the neutron sensitivity of 6 LiF:Mg,Ti (TLD-600). © 2007 Elsevier Ltd. All rights reserved. Keywords: Neutron dosimetry; Optically stimulated luminescence; Neutron converters
1. Introduction In spite of the successful use of the optically stimulated luminescence (OSL) technique in personal dosimetry of photons (X-ray and gamma) and beta rays, the lack of a neutron sensitive OSL dosimeter prevents the application of the technique to neutron fields, representing the main disadvantage of OSL when compared to the thermoluminescence (TL) technique (McKeever and Moscovitch, 2003). Carbon-doped aluminium oxide (Al2 O3 : C), considered the reference OSL material in dosimetry because of its high sensitive and optical properties, has a neutron sensitivity lower than 7 LiF:Mg,Ti (TLD-700) (Klemic et al., 1996). Today, the OSL technique is the basis for the Luxel姠 and InLight姠 dosimetry systems (Landauer Inc.) (BZtter-Jensen et al., 2003) and 1.5 million users are monitored using the OSL technology. The introduction of a neutron sensitive OSL dosimeter would extend the advantages of the OSL technique, ∗ Corresponding author. Tel.: +1 405 744 5051; fax: +1 405 744 6811.
E-mail address: [email protected] (E.G. Yukihara). 1350-4487/$ - see front matter © 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.radmeas.2007.10.005
such as precise control of stimulation, fast readout that results in high productivity, wide range of linearity and high sensitivity, and possibility of re-estimation of absorbed dose (Akselrod and McKeever, 1999), to personal dosimetry in neutron fields. Neutron detection relies on secondary charged particles and photons created by neutron interaction with the detector itself or surrounding material. Elastic neutron scattering with hydrogen nuclei create recoil protons with energies comparable to the incident neutrons, whereas nuclear capture reactions such as the ones exemplified in Table 1 create various types of charged particles with energies of a few MeV, gamma rays, and conversion electrons. These secondary radiation deposit energy in the detector giving rise to a neutron induced signal. Thermoluminescence detectors (TLD) have been used for decades in personal neutron dosimetry. Materials such as LiF:Mg,Ti, LiF:Mg,Cu,P and Li2 B4 O7 : Mn can be prepared with higher concentration of 6 Li or 10 B isotopes, which have high cross-section for neutron capture (McKeever et al., 1995). Because the cross-section decreases with increasing neutron energies, these dosimeters are mostly sensitive to thermal and epithermal neutrons and, therefore, are used to detect the
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Table 1 Examples of isotopes of interest for neutron detection, their relative abundance, neutron capture cross-section for thermal neutrons, and products (Knoll, 2000; van Eijk, 2004) Isotope
Natural abundance (%)
(barns)
Products
6 Li
7.4
940
3H
(2.75 MeV) + 4 He (2.05 MeV)
10 B
19.8
3840
7 Li
(1.0 MeV) + 4 He (1.8 MeV) (0.83 MeV) + 4 He (1.47 MeV) + (0.48 MeV)
7 Li 157 Gd
15.7
255,000
158 Gd
+s + conversion e− + X-rays (29.182 keV)
155 Gd
14.8
60,900
156 Gd
+s + conversion e− + X-rays (39.199 keV)
neutrons scattered by the user’s body in the so-called albedo configuration (Piesch and Burgkhardt, 1985). The possibility of performing neutron imaging with the OSL technique using Al2 O3 : C sheets and Gd2 O3 neutron converters was recently demonstrated (Kobayashi et al., 2005), which hints to the possibility of using neutron converters to create a new OSL dosimeter based on Al2 O3 : C. This approach has the potential of shortening the lead time for the introduction of a useful neutron-sensitive OSL dosimeter in several years, since such a neutron dosimeter would have all the properties of Al2 O3 : C (optical sensitivity, absence of fading, stimulation and emission spectrum, etc.) with the desired neutron sensitivity. The question is whether such approach can produce a dosimeter with sensitivity high enough for personal dosimetry or not. This paper demonstrates the possibility of using neutron converters to produce new OSL dosimeters with neutron sensitivity suitable for personal dosimetry. We investigated the effect of different neutron converters and the size of Al2 O3 : C grains on the neutron sensitivity. Based on these results, we produced a new neutron-sensitive OSL composite material identical to the one used in the Luxel姠 and InLight姠 dosimetry systems (Landauer Inc.), except for the lower gamma sensitivity and enhanced neutron sensitivity. The gamma and neutron sensitivity of these new materials are compared to regular Al2 O3 : C dosimeter (Luxel姠 ) and 6 LiF:Mg,Ti (TLD-600). 2. Materials and methods 2.1. Samples The dosimeters used in this study were prepared with a powder mixture of Al2 O3 : C and one of the following neutron converters: lithium fluoride enriched with 6 Li (6 LiF), lithium carbonate enriched with 6 Li (6 Li2 CO3 ), gadolinium oxide (Gd2 O3 ), boric acid enriched with 10 B (H3 10 BO3 ), and high-density polyethylene (HDPE). Preliminary investigations on the performance of the neutron converters were performed using loose powder and cold-pressed pellets. Based on these preliminary results, a neutron sensitive composite material consisting of a thin layer of Al2 O3 : C (grain size < 38 m) and 6 LiF powder deposited in a clear polyester film, hereafter designated N-OSL tape, was manufactured and tested. This tape is the first version of a neutron sensitive OSL
dosimeter produced by the same processes used in the production of the regular Al2 O3 : C tape used in the Luxel姠 and InLight姠 badges (Landauer Inc.). In addition, 3×3×0.9 mm TLD chips of LiF:Mg,Ti enriched with 6 Li and 7 Li (TLD-600 and TLD-700, ThermoElectron Inc.) were used for comparison. 2.2. Radiation sources Preliminary neutron irradiations were carried out using an uncalibrated Pu–Be source to compare the sensitivity of different samples. To precisely determine the neutron sensitivity, irradiations were carried out at the SCK-CEN calibration room using a calibrated bare 252 Cf source (Vanhavere et al., 2001). All irradiations were carried out with the dosimeters mounted on a 30 × 30 × 15 cm PMMA phantom on the side facing the source (for more details, see Mittani et al., 2007). Irradiations with a shadow cone placed between the source and the phantom were carried out to estimate the contribution from room scattering on the results (ISO 8529-2, 2000). The gamma sensitivity of the dosimeters was obtained using a calibrated 60 Co source from SCK-CEN. Additional irradiations were carried out using a calibrated 90 Sr/90 Y beta source at OSU. 2.3. OSL and TL readouts OSL measurements were carried out using two types of readers: RisZ TL/OSL-DA-15 readers and an in-house built portable reader. Continuous-wave OSL (CW-OSL) were performed using the RisZ TL/OSL readers, green or blue LEDs for stimulation, and Hoya U-340 filters (transmission between 290 and 370 nm) in front of the photomultiplier tube (PMT) (BZtter-Jensen et al., 2000). Pulsed-OSL measurements were carried out using the in-house built portable reader, green LEDs for stimulation, and Corning 5-58 filters (transmission between 380 and 440 nm) in front of the PMT. It should be pointed out that the results obtained with the two readers are not directly comparable, since the Hoya U-340 filters transmits the UV emission band observed at 335 nm in addition to the main luminescence band of Al2 O3 : C, centered at 420 nm (Yukihara and McKeever, 2006b), whereas the Corning 5-58 filters block the UV emission. The relative intensity of the UV and F-center emission bands can vary with the radiation field (Yukihara and McKeever, 2006a; Yukihara et al., 2006).
E.G. Yukihara et al. / Radiation Measurements 43 (2008) 309 – 314
TL measurements of TLD-600 and TLD-700 dosimeters were carried out using the RisZ TL/OSL reader and Schott BG-39 filters (6 mm total thickness).
7
2.4. Data analysis
5
The OSL intensity over 600 s of stimulation was used as the signal S. The dosimeter’s response was calculated from the OSL signal S using the relationship:
4
R = Rn + R .
(2)
The gamma component R was estimated using the neutron insensitive dosimeters (Al2 O3 : C or TLD-700). The neutron response was also assumed to have two components, one due to the direct neutrons from the source and albedo neutrons scattered by the phantom (R0 ), plus the response due to neutrons scattered by the room (Rs ): Rn = R0 + Rs .
Rn/Rγ
(1)
where the calibration factor a was obtained by irradiation with known gamma doses. Therefore, the response of the dosimeter is the gamma dose required to produce an OSL signal of the same intensity as S. The neutron and gamma response of the dosimeters was separated using the dosimeter pair method. The response of the neutron sensitive detectors (e.g., Al2 O3 : C+ neutron converter or TLD-600) was assumed to have a neutron and a gamma component:
(3)
The contribution from room scattering Rs was estimated using dosimeters irradiated in the same conditions as before, but with a shadow cone mounted between the source and the phantom to block the neutrons coming directly from the source (ISO 8529-2, 2000). The neutron response obtained using the dosimeters irradiated with the shadow cone provide an estimate for Rs . The neutron sensitivity of the dosimeters was defined as the ratio between the neutron response and the reference neutron dose. 3. Results
6LiF
Gd2O3 6Li CO 2 3
6
H310BO3
3 2 1 0 0.0
0.5 1.0 1.5 2.0 2.5 mass of neutron converter/mass of Al2O3:C
3.0
Fig. 1. Comparison of the neutron sensitivity of mixtures of Al2 O3 : C with different neutron converters, expressed by the ratio between the neutron and gamma response for a 2 h exposure to a Pu–Be source neutron-to-gamma ratio Rn /R , as a function of the neutron converter to Al2 O3 : C mass ratio. The data was obtained using loose powder, and the OSL measurements were carried out using the RisZ reader, green stimulation, and Hoya U-340 filters in front of the PMT.
35 30 Response (mGy)
R = aS,
311
25 20 15 10 5 0
Fig. 2. Response of different samples exposed to a nominal neutron dose of 30 mSv of a bare 252 Cf source. The samples were irradiated on the face of a 30 × 30 × 15 cm PMMA phantom. The Al2 O3 : C+ neutron converter dosimeters were cold-pressed pellets, and the OSL measurements were carried out using the RisZ reader, green stimulation, and Hoya U-340 filters in front of the PMT.
3.1. Comparison of neutron converters Preliminary investigations were carried out to compare the performance achieved using different neutron converters. Aliquots of Al2 O3 : C powder mixed with different neutron converters were exposed to the Pu–Be source and the OSL signal was measured using the RisZ reader. Pure Al2 O3 : C was used to estimate the gamma contribution of the field, and the neutron-to-gamma ratio Rn /R was calculated as discussed in Section 2.4. Fig. 1 shows the neutron-to-gamma ratio Rn /R as a function of the mass ratio between neutron converter and Al2 O3 : C
in the mixture. Rn /R increases steadily with increasing proportion of neutron converter in the mixture. The largest neutron sensitivities were obtained using 6 LiF and 6 Li2 CO3 , followed by Gd2 O3 and H3 10 BO3 . The neutron sensitivity of the mixture containing 6 Li2 CO3 is approximately 70% of the neutron sensitivity of the mixture containing 6 LiF. This is caused by the fact that the number of Li atoms per mass of compound of Li2 CO3 is ∼ 70% the value of LiF. The neutron sensitivity obtained with H3 10 BO3 was very small, probably because the secondary particles produced by the
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10 B(n, )7 Li
reaction ( and 7 Li) have small range in Al2 O3 : C (a few microns or less). The high sensitivity obtained with lithium compounds is probably related to the triton produced in the 6 Li(n, )3 H reaction, which has a range of ∼ 20 m in Al2 O3 : C (considering an energy of 2.75 MeV). Based on these results, H3 10 BO3 was not included in further investigations. Although the data in Fig. 1 provides a comparison between the neutron sensitivity of the different samples, it does not show that the overall sensitivity obviously decreases with decreasing amounts of Al2 O3 : C in the mixture. An optimum compromise between neutron sensitivity and overall sensitivity was found for a neutron converter/Al2 O3 : C mass ratio between 0.6 and 1 (Mittani et al., 2007). To determine the neutron sensitivity achieved with various neutron converters, cold pressed pellets of Al2 O3 : C and neutron converter mixed in a 1:1 mass proportion were prepared
Al2O3:C grain sizes Regular <38 μm
3.5 3.0
Rn /Rγ
2.5 2.0 1.5 1.0 0.5 0.0 6Li
2F
6Li
2CO3
Gd2O3
Neutron converter Fig. 3. Comparison of the neutron sensitivity of mixtures of Al2 O3 : C with different neutron converters, expressed by the neutron-to-gamma ratio Rn /R for a 2 h exposure to a Pu–Be source. The data was obtained using OSL tapes produced in laboratory using Al2 O3 : C with either regular grain sizes (10.100 m) or small grain sizes (< 38 m). The OSL measurements were carried out using the RisZ reader, blue stimulation, and Hoya U-340 filters in front of the PMT.
and irradiated using the 252 Cf source. Single crystal Al2 O3 : C, regular Al2 O3 : C tape (Luxel姠 ), and TLDs were also irradiated in the same conditions for comparison. The responses of the various materials irradiated with a nominal neutron dose of 30 mSv are compared in Fig. 2. The response of the neutron insensitive dosimeters, Al2 O3 : C single crystals, Al2 O3 : C Luxel姠 , and TLD-700, varies between 1.96 and 2.31 mGy, giving an estimation for the gamma component. The highest response was from TLD-600 (34.9 mGy), followed by Al2 O3 : C+ 6 LiF (20.1 mGy). The response of Al2 O3 : C+ 6 Li CO is again approximately 70% of the response of the 3 2 Al2 O3 : C+ 6 LiF. In the case of the Al2 O3 : C + Gd2 O3 mixture the response was 6.3 mGy, which is higher than the gamma contribution estimated using Al2 O3 : C or TLD-700 dosimeters. However, the increased response is partly caused by the high atomic number of Gd and the resultant increased gamma sensitivity. The gamma contribution in this case needs to be estimated using the Al2 O3 : C + Yb2 O3 mixture, which has approximately the same effective atomic number but is not neutron sensitive (Pomerance, 1951). As expected, the response of Al2 O3 : C + Yb2 O3 mixture (3.7 mGy) was significantly higher than the response from the other gamma sensitive dosimeters. This preliminary investigation suggests that compounds containing 6 Li are the most promising neutron converters to develop the new OSL neutron dosimeters. The effect of 6 LiF and 6 Li CO neutron converters seems to differ only with respect 3 2 to the number of Li atoms per mass of compound. Another reason for choosing these neutrons converters is the fact that they have a low atomic number and do not alter the dosimeter’s sensitivity and effective atomic number as much as the Gd2 O3 neutron converters. The complete data on response and sensitivity of the various neutron converters can be found in Mittani et al. (2007). 3.2. Optimizing the neutron sensitivity Although the investigations in the previous section demonstrate the possibility of using neutron converters to produce neutron sensitive OSL dosimeters based on Al2 O3 : C, the
Fig. 4. Neutron sensitive OSL tape produced using a mixture of Al2 O3 : C and 6 LiF neutron converter.
E.G. Yukihara et al. / Radiation Measurements 43 (2008) 309 – 314
Table 2 Neutron response of the TLD pair (TLD-600 and TLD-700) and OSLD pair (new and regular OSL strip) to a nominal neutron dose of 30 mSv from a 252 Cf source
12 Gamma irradiation RegularAl2O3:C tape (LuxelTM) N-OSLtape
S - Sc (106 counts)
10
Dosimeter pair Rn (mSv)a (Dref = 45.6 mSv)
8
TLDs OSLDs
6
33.1 ± 1.3 20.7 ± 0.4
Rs (mSv)b (Dref = 15.6 mSv)
R0 (mSv)c (Dref = 30 mSv)
23.6 ± 0.7 14.3 ± 0.6
9.5 ± 1.5 6.4 ± 0.7
The irradiations were carried out with the dosimeters on a phantom at 1.5 m from the source. The uncertainties are the propagated values based on the readout of three dosimeters of each type. a Response to the total field (direct neutrons, albedo neutrons scattered from the phantom, and neutrons scattered by the room). b Response to the room scattering estimated using a shadow cone placed between the source and the phantom. c Response only to the source and to the albedo neutrons scattered by the phantom, obtained by R0 = Rn − Rs .
4
2
0 0
45
10
20
30 40 Hp(10) (mSv)
50
60
the grains which is not sensitive to neutrons, only to gammas. Using small grain sizes, it is possible to decrease the inactive volume and increase the neutron sensitivity compared to the gamma sensitivity. The effect of reducing the grain size in Al2 O3 : C powder was investigated using tapes prepared in laboratory using regular and fine Al2 O3 : C powder mixture mixed in a 1:1 mass ratio. Fig. 3 shows the neutron-to-gamma ratio for samples prepared using neutron converters and Al2 O3 : C powder of either regular grain sizes or grain sizes less than 38 m. Larger neutron-togamma ratios can be obtained using fine Al2 O3 : C powder in the case of 6 LiF and 6 Li2 CO3 . In the case of Gd2 O3 , no difference was observed as expected from the fact that the energy is deposited by conversion electrons and gammas, which irradiate the Al2 O3 : C grains more uniformly.
Neutron irradiation TLD-600 TLD-700 N-OSL tape
40 35 R - Rc (mSv)
313
Regular Al2O3:C tape (LuxelTM)
30 25 20 15 10 5 0 0
5
10
15 20 Hp(10) (mSv)
25
30
35
Fig. 5. (a) Gamma response of regular OSL tape and new neutron sensitive OSL tape, expressed by the total OSL signal S minus signal from control dosimeters Sc , as a function of personal dose equivalent Hp (10) for 60 Co irradiation. (b) Response of various dosimeters as a function of nominal personal dose equivalent Hp (10) for 252 Cf irradiation. The response includes the neutron and gamma components of the radiation field. The irradiations were carried out on a 30 × 30 × 15 cm PMMA phantom, at 1.5 m from the radiation source. The OSL measurements were carried out using the portable POSL system, with green stimulation, and Corning 5-58 filters in front of the PMT.
production of OSL tapes involves additional processes that degrade the neutron sensitivity. Improvements in the Al2 O3 : C+ neutron converter mixture were made to counteract some of these effects. One of these improvements is the use of Al2 O3 : C in fine powder. The grain sizes of the regular Al2 O3 : C powder used in the production of the standard Al2 O3 : C OSL tape range from 10 to 100 m. However, the 3 H produced by the 6 Li(n,)3 H neutron capture reaction have limited range in Al2 O3 : C. As a result, for large grains there may be a volume in the center of
3.3. Characteristic of new OSL neutron dosimeter Based on the information obtained in the previous sections, the N-OSL tape shown in Fig. 4 was fabricated using a mixture of Al2 O3 : C fine powder (< 38 m) and 6 LiF. Samples were prepared from this tape and irradiated with neutrons alongside with samples of regular Al2 O3 : C tape (Luxel姠 ) using the 252 Cf source at SCK-CEN. Samples from the new and regular OSL tape were irradiated with various doses using the 60 Co source to compare the sensitivity and linearity of the dosimeters. Fig. 5a shows the gamma response of both samples. The response of the new OSL strip is approximately 35% of the sensitivity of the regular Al2 O3 : C tape. This results gives an idea of the reduction in sensitivity caused by the reduced mass of Al2 O3 : C in the new OSL tape. However, the number is only a rough estimate, since the regular and neutron sensitive OSL tape were produced in different periods using Al2 O3 : C powder from different batches. To determine the neutron response, the samples were irradiated with various neutron doses using the 252 Cf source. TLDs were also irradiated for comparison. Fig. 5b shows the response of the various dosimeters as a function of the nominal neutron dose. The response of the regular Al2 O3 : C strip and TLD-700
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Table 3 Neutron sensitivities of the TLD pair (TLD-600 and TLD-700) and OSLD pair (new and regular OSL strip) to a 252 Cf source Dosimeter pair
Rn /Dref a
Rs /Dref b
R0 /Dref c
TLDs OSLDs
0.727 ± 0.015 0.453 ± 0.008
1.48 ± 0.06 0.96 ± 0.03
0.337 ± 0.014 0.191 ± 0.006
The irradiations were carried out with the dosimeters on a phantom at 1.5 m from the source. The values and propagated uncertainties were calculated based on the responses shown in Error! Reference source not found.b, plus two additional irradiations with nominal neutron doses of 10 and 30 mSv with a shadow cone in between the source and the phantom. a Neutron sensitivity to the total neutron spectrum (direct neutrons, albedo neutrons scattered from the phantom, and neutrons scattered by the room). b Neutron sensitivity to the room scattering spectrum, which in these conditions contributes with ∼ 33% of the total neutron dose (Vanhavere et al., 2001). c Neutron sensitivity to the direct and albedo neutron spectrum.
are between 8.6% and 9.8% of the nominal neutron dose, indicating the gamma contribution of the radiation field. The response of the new OSL tape is linear and approximately 60% of the TLD-600 dosimeter’s response. Based on the data shown in Fig. 5, the total neutron response Rn , response to the neutrons scattered by the room, and response to the direct neutrons from the source and albedo neutrons scattered by the phantom R0 were calculated for the OSL and TL dosimeters as described in Section 2.4. The results for a nominal dose of 30 mSv are shown in Table 2. The neutron response of the OSLDs is ∼ 60% of the response of the TLDs in all cases. The neutron sensitivities of the TLDs and OSLDs shown in Table 3 were obtained dividing the neutron response to the respective reference neutron dose. 4. Conclusions The results presented in this paper demonstrate the possibility of producing new OSL dosimeters using a mixture of Al2 O3 : C and 6 LiF or 6 Li2 CO3 neutron converters. Higher neutron sensitivity of 6 Li compounds in comparison with 10 B converters seems to be related to the large range in Al2 O3 : C of the secondary 3 H particles, produced as a result of the 6 Li(n, )3 H neutron capture reaction. Other neutron converters resulted in lower neutron sensitivity and, in the case of Gd2 O3 , altered gamma sensitivity that requires proper compensation. Based on the investigations on the various neutron converters, a new neutron sensitive OSL tape made of Al2 O3 : C and 6 LiF was produced. This new OSL tape is identical to the regular Al2 O3 : C tape used in the Luxel姠 and InLight姠 badges, except for the gamma and neutron sensitivity. The gamma sensitivity was smaller (∼ 35%) of the regular tape’s sensitivity due to the decreased amount of Al2 O3 : C. However, the neutron response was significantly enhanced, reaching a value of ∼ 60% of the TLD-600 neutron sensitivity. The new OSL dosimeters developed in this study have the advantage of using the same reaction employed by the TLD-600 dosimeters, since both are based on the 6 Li(n, )3 H neutron
capture reaction. As a result, the neutron energy response and behavior of these new OSL dosimeters should be identical to the TLD-600 dosimeter, making all results on neutron dosimetry using TLD-600 immediately applicable to the neutron sensitive OSLDs. Although this approach still requires the use of the dosimeters in albedo configuration, the vast experience gained in albedo dosimetry using TLDs can be put to use with the new OSL dosimeters, with all the advantages of the OSL technique. Acknowledgments The authors would like to thank Bart Marlein and Ludo Melis for assistance with the irradiations at SCK-CEN. This work was supported by Landauer Inc. References Akselrod, M.S., McKeever, S.W.S., 1999. A radiation dosimetry method using pulsed optically stimulated luminescence. Radiat. Prot. Dosim. 81, 167–176. BZtter-Jensen, L., Bulur, E., Duller, G.A.T., Murray, A.S., 2000. Advances in luminescence instrument systems. Radiat. Meas. 32, 523–528. BZtter-Jensen, L., McKeever, S.W.S., Wintle, A.G., 2003. Optically Stimulated Luminescence Dosimetry. Elsevier, Amsterdam. ISO 8529-2, 2000. Reference neutron irradiations—part 2: calibration fundamentals related to basic quantities characterizing the radiation field. International Organization for Standardization. Klemic, G.A., Azziz, N., Marino, S.A., 1996. The neutron response of Al2 O3 : C, 7 Li:Mg,Cu,P, and 7 LiF:Mg,Ti TLDs. Radiat. Prot. Dosim. 65 (1–4), 221–226. Knoll, G.F., 2000. Radiation Detection and Measurements. Wiley, New York. Kobayashi, H., Satoh, M., Kobayashi, I., Morishima, H., 2005. Neutron imaging using an optically stimulated luminescence material: -Al2 O3 : C+ Gd2 O3 . IEEE Trans. Nucl. Sci. 52 (1), 360–363. McKeever, S.W.S., Moscovitch, M., 2003. On the advantages and disadvantages of optically stimulated luminescence dosimetry and thermoluminescence dosimetry. Radiat. Prot. Dosim. 104, 263–270. McKeever, S.W.S., Moscovitch, M., Townsend, P.D., 1995. Thermoluminescence Dosimetry Materials: Properties and Uses. Nuclear Technology Publishing, Ashford. Mittani, J.C.R., da Silva, A.A.R., Vanhavere, F., Akselrod, M.S., Yukihara, E.G., 2007. Investigation of neutron converters for production of optically stimulated luminescence (OSL) neutron dosimeters using Al2 O3 : C. Nucl. Instrum. Methods Phys. Res. B 260, 663–671. Piesch, E., Burgkhardt, B., 1985. Albedo neutron dosimetry. Radiat. Prot. Dosim. 10 (1–4), 175–188. Pomerance, H., 1951. Thermal neutron capture cross sections. Phys. Rev. 83 (3), 641–645. van Eijk, C.W.E., 2004. Neutron detection and neutron dosimetry. Radiat. Prot. Dosim. 110 (1–4), 5–13. Vanhavere, F., Coeck, M., Vermeersch, F., 2001. Procedures for neutron scattering corrections in a calibration facility with a non-symmetric set-up. Radiat. Prot. Dosim. 93 (1), 5–10. Yukihara, E.G., McKeever, S.W.S., 2006a. Ionization density dependence of the optically and thermally stimulated luminescence from Al2 O3 : C. Radiat. Prot. Dosim. 119, 206–217. Yukihara, E.G., McKeever, S.W.S., 2006b. Spectroscopy and optically stimulated luminescence of Al2 O3 : C using time-resolved measurements. J. Appl. Phys. 100 (8), 083512. Yukihara, E.G., Sawakuchi, G.O., Guduru, S., McKeever, S.W.S., Gaza, R., Benton, E.R., et al., 2006. Application of optically stimulated luminescence (OSL) technique in space dosimetry. Radiat. Meas. 41, 1126–1135.