Experimental cross section of the 71Ga(n,γ)72Ga reaction at 0.0334 eV energy

Nuclear Instruments and Methods in Physics Research B 336 (2014) 1–5

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Experimental cross section of the energy

71

Ga(n,c)72Ga reaction at 0.0334 eV

N. Afroze a,b, M.S. Uddin a,⇑, S.M. Hossain a, M.A. Islam a, M.A. Shariff a, A.K.M. Zakaria a, T.K. Datta a, S.M. Azharul Islam b a b

Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, GPO Box No. 3787, Savar, Dhaka 1000, Bangladesh Department of Physics, Jahangirnagar University, Savar, Dhaka, Bangladesh

a r t i c l e

i n f o

Article history: Received 25 March 2014 Received in revised form 13 June 2014 Accepted 16 June 2014

Keywords: Thermal neutron 0.0334 eV energy Activation technique Gallium (Ga) Neutrons powder diffractometer Reactor

a b s t r a c t The cross section of the 71Ga(n,c)72Ga reaction at 0.0334 eV was measured for the first time using monochromatic neutrons from powder diffractometer at TRIGA Mark II nuclear reactor. The 197Au(n,c)198Au reaction was used to monitor the neutron beam intensity. The HPGe c-ray spectrometry was used to determine the radioactivity of the product radionuclides. The obtained cross section value amounted 3.42 ± 0.27 b is about 95% consistent with JENDL-4, but about 17% and 14% lower than that of the ENDF/B-VII and TENDL-2012 data libraries, respectively. The measured value at 0.0334 eV and the previous measured value at 0.0536 eV would be useful to confirm the reliability of the data evaluated by 1/v relation in the above libraries. Ó 2014 Elsevier B.V. All rights reserved.

1. Introduction The investigations of neutron induced reactions are of considerable interest, not only for their importance to fundamental research in Nuclear Physics, Reactor Physics and Astrophysics, but also for practical applications in nuclear technology, dosimetry, radiation safety, development of radiation detector, improving evaluated data libraries, etc. These tasks require improved nuclear data and high precision cross sections for neutron induced reactions. The (n,c) reaction in a nuclear reactor is of great interest with respect to the production of several radioisotopes. It has been, therefore, extensively investigated, but all the measurements have generally been done at the average thermal energy 0.025 eV [1–8]. A technique which might be included in the list of spectrometric methods is that of Cadmium filtering. The absorption cross section of Cd is special in that it is large for energies less than 0.2 eV and small for higher energies. It can, therefore, be used to selectively filter out thermal energy neutrons. Comparison between activation foils or wires shielded with Cd to those without is used to separate contributions to the neutron spectrum from thermal and epithermal neutrons. This technique has, however, the following problems,

⇑ Corresponding author. Tel. +880 1715363326. E-mail address: [email protected] (M.S. Uddin). http://dx.doi.org/10.1016/j.nimb.2014.06.016 0168-583X/Ó 2014 Elsevier B.V. All rights reserved.

which are responsible for the inclusion of uncertainties in the measured cross section: (i) The Cd shielding allows correction for the thermal contribution, but correction of the high energy component is difficult. It should be mentioned that the fast neutrons with energies from epithermal up to 20 MeV amount to 36% of the total number of neutrons in the spectrum for TRIGA Mark II nuclear reactor [9]. Therefore, special attempt is needed to correct the effect of fast neutrons, which are not taken care of by Cd-shielding. (ii) The diameter of the activation Ga2O3 pellet and Cd shield were about 1.3 and 2 cm, respectively. The counting is performed ex-situ, meaning that activation rates are determined after removing the activated samples from the irradiation environment. The large activity produced in Cdcover creates radiation hazard, which demands long cooling time for handling of radioactive sample and a large part of the produced radionuclide 72Ga (T1/2 = 14.1 h) decays out before counting and increasing statistical uncertainty. (iii) During the irradiation in the core of a reactor, the sample is exposed to a neutron field (thermal, epithermal and fast neutrons). If sample covered with Cd and sample without cover are irradiated together, thermal neutrons cannot reach one side of the bare sample due to Cd-filtering, which adds a large uncertainty in the measured cross section.

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Thus, it needs the detail study on cross sections at broad energies in thermal region (obey the 1/v law, where v is the speed of incident neutron) to investigate the energy dependence of neutron capture cross section with adequate precision and accuracy. Uses of guided neutron beams from the steady state research reactor can be the solution to address the problem; in certain cases neutron beam choppers, various neutron filter- or crystal-based spectrometers can be applied in order to provide with well-defined energy selection. The experimental determination of neutron capture cross sections for various targets using the reflected monochromatized (Cu200 plane) beam (0.0536 eV) from the triple axis spectrometer (TAS) installed at 3 MW TRIGA Mark-II research reactor is well established [10–15]. The measurements of neutron capture cross sections at average thermal energy 0.025 eV in the reactor environment is a common practice. The measurement of cross sections using neutrons out of average thermal energy is rare due to the non-availability of monoenergetic neutrons. Recently, a high resolution neutron powder diffractometer (locally called SAND) facility is installed in a radial beam port of the above reactor for neutron scattering experiments. This provides a relatively clean neutron beam of energy 0.0334 eV and the neutron beam intensity was sufficient to make good activation of gallium target. It was, therefore, possible to perform cross section measurement utilizing this new experimental facility. The possibility of using this new facility for determination of neutron capture cross section has already been assessed and reported to elsewhere [16]. The present work deals with measurement of cross section of the 71Ga(n,c)72Ga reaction at energy of 0.0334 eV. This study was carried out as a part of our systematic measurements on neutron capture cross sections using the monochromatic reflected neutrons at TRIGA Mark II research reactor [10–15]. Gallium (Ga) was selected as target because it is a semiconductor element and used as main component in radiation detector. Due to a very low melting point and a very high boiling point, it also becomes a promissory element among various candidates for liquid–metallic coolant.

2. Experimental technique 2.1. Neutron source The monochromatic neutrons of 0.0334 eV energy were used for activation of sample. The neutrons of various wavelength produced in the TRIGA Mark II reactor installed at the campus of Atomic Energy Research Establishment, Savar, Dhaka were monochromatized by neutron powdered diffractometer namely SAND, which is installed at a radial beam port of the reactor. A detail of the diffractometer as well as neutron monochromator facility has already been reported in [14]. A brief schematic diagram on the arrangement for monochromatization of reactor neutrons and experimental setup is shown in Fig. 1. The neutrons of various energies coming from reactor core were passed through a sapphire crystal with size of 12.7 cm in diameter and 12.24 cm in length mounted in the tapered section at the front (upstream) end of the collimator to filter fast neutron before entering the monochromator. Monochromatization can very effectively be done by Bragg reflection from a Si(115) single crystal. The monochromator was well shielded. Thus shielding with size of 228  172  176 cm3 was consisted with seven steel jacketed, internally reinforced, heavy concrete filled blocks. The monochromator is fabricated from 9 single crystal silicon slabs of 1.45  0.53  19.05 cm3 that have been cut from the same 0.6 cm thick wafer of silicon. The monochromator is positioned with the bending screw at the top, appropriate for use of the Si(115) reflection. The thickness, offset angle and bending radius of silicon slabs

have been selected to optimize the intensity and diffractometer resolution for the Si(115) reflection of the monochromator at 97o take-off angle and a sample–monochromator distance of 128.62 cm yielding a wave length of k = 1.5656 Å, which corresponds to 0.0334 eV neutron energy. The uncertainty in wave length as well as energy was derived from refinement of Ni standard; it was amounted to 0.004%. After reflection the monochromatic neutrons pass through a collimator with inner size of 94.62  11.4  5 cm3 and then focus on the sample. It should be pointed out that the size of neutron beam at the sample position was 94.62  11.4  5 cm3. The inside of collimator is made by boronated polythene and outside by stainless steel. 2.2. Sample preparation and irradiation The gallium oxide (99.99% purity) powder for gallium target of natural isotopic composition (69Ga-60.108% and 71Ga-39.892%) was pressed at a pressure of 5 tons per cm2 to prepare a pellet. The weight of the pellet was 1.192 g with a diameter of 1.3 cm. The pellet was covered by polyethylene bag and then sandwiched between two thin pure gold foils (25 lm thick, 61 and 59 mg weight). The Au-foils were used to monitor beam intensity at both entrance and exit of gallium oxide pellet during the irradiation. A Cd-covered Au-foil was also irradiated to check any contribution of epithermal neutrons. The sample was irradiated with unidirectional monoenergetic neutrons of 0.0334 eV for 3 h and 40 min at the outer end position of the last collimator. During irradiation the reactor was operated at 2 MW power. 2.3. Gamma-ray measurement and data analysis The radioactivity induced in gallium oxide and monitor foils was measured nondestructively using the high-purity germanium (HPGe) gamma-ray spectroscopy (Canberra, 25% relative efficiency, 1.8 keV resolution at 1332.5 keV of 60Co). The detector was coupled with digital gamma spectrometry system (ORTEC DSPEC jr™) and Maestro data acquisition software. A typical c-ray spectrum for the irradiated gallium oxide sample is shown in Fig. 2. Each irradiated sample was counted 3 times giving enough intervals to avoid interference from gamma-lines of undesired sources and to obtain cross section value with adequate precision and accuracy. All samples were counted directly on the surface of the detector to have good counting statistics. The extended size (1.3 cm) of the samples demands a correction in the efficiency of the detector. To make this correction a sample of 1.3 cm diameter was irradiated in the core of reactor to obtain sufficient activity and then counted both at surface and 15 cm from the detector surface. By considering the activity at 15 cm as standard value, where the sample-size effect on the efficiency was negligible, the detector efficiency for the extended sample on the surface was determined. The efficiency versus energy curve of the HPGe gamma-ray detector was determined using the standard point sources, 57Co, 60 Co, 133Ba, 137Cs and 152Eu. The same efficiency curve was used to determine the activities of both the 198Au and 72Ga radionuclides induced in monitor and sample target, respectively. From the measured count rate of the product radionuclide, the reaction rate was deduced by correcting for the gamma-ray intensities and the efficiency of the detector using the following formula:

R¼

k C N e Ic ektc ð1  ektm Þ  ð1  ekti Þ

ð1Þ

where, k = decay constant (s1), C = total counts of gamma-ray peak area, N = number of target atoms, e = detector efficiency,

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N. Afroze et al. / Nuclear Instruments and Methods in Physics Research B 336 (2014) 1–5

Fig. 1. A schematic view of (a) monochromatic system, and (b) experimental arrangement of sample for irradiation.

Table 2 Sources of uncertainties considered in order to derive the total uncertainty in the measured cross section. Sources of uncertainty

Uncertainty (%)

Statistical uncertainty of c-ray counting Peak area analysis Efficiency calibration Sample mass Half-life Isotopic abundance Self-shielding factor Neutron flux Total

1–4 2 4 0.01 0.01 0.016 0.01 6 8

uncertainties are given in Table 2. The overall uncertainty in the cross section is around 8%. 2.4. Neutron flux measurement Fig. 2. A typical c-ray spectrum for the irradiated gallium oxide sample; measuring time-1500 s and cooling time-1000 s.

From the measured activity of 198Au produced via the Au(n,c)198Au reaction in the two Au-foils the neutron flux was determined. The Au-foil covered with Cd was counted and there was no epithermal neutron contribution observed. The problem with the monitor reaction lies in the selection of the standard cross section, because no recommended cross section value has been reported at 0.0334 eV energy. The standard cross section of the 197 Au(n,c)198Au reaction at the above energy was evaluated from the experimental data reported by Yamamoto et al. [18], Pavlenko et al. [19], Haddad et al. [20] and the data reported in ENDF/B-VII [21] library by a fitting procedure. Chowdhury et al. have described the detail evaluation of the monitor reaction cross section [10]. The obtained value for the 197Au neutron absorption cross section is 86.26 ± 2.1 b at 0.0334 eV. The neutron beam intensities obtained at entrance and exit of the target are 7.54  104 and 6.50  103 n cm2 s1, respectively. To follow the neutron attenuation along the Ga2O3 target, the total thickness (1.4 mm) was divided into four segments and the neutron fluxes both at entrance and exit of each segment were calculated by the following equation. 197

Ic = gamma-ray intensity, tc = cooling time(s), tm = counting time (s) and ti = irradiation time (s). From the reaction rate and the measured neutron flux (n cm2 s1), /(E), the cross-section r(E) can simply be obtained as

rðEÞ ¼

R /ðEÞ

ð2Þ

The decay data of the residual radionuclides were taken from the NUDAT (2009) [17] database http://www.nndc.bnl.gov/nudat2 and are given in Table 1. The combined uncertainty in the experimentally determined cross section was estimated by taking the square root of the quadratic sum of the individual uncertainties. The sources of Table 1 Decay characteristics of the investigated nuclides and the contributing reactions.* Nuclear reaction

Halflife

Gamma-ray energy (keV)

Branching ratio (%)

I ¼ Io expðltÞ

71

14.1 h

834.03 629.96 411.8

95.45 24.8 95.5

where, Io is the incident flux and I is the flux of neutron at exit of sample matrix of a thickness, t. The constant l is the linear attenuation coefficient. The various types of interactions with matter are combined into a total attenuation coefficient:

72

Ga(n,c) Ga

197

*

Au(n,c)198Au

2.695 d

Data were taken from NUDAT (2009) database [17].

ð3Þ

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N. Afroze et al. / Nuclear Instruments and Methods in Physics Research B 336 (2014) 1–5

l ¼ lscatter þ lneutron capture þ lfission þ . . .

ð4Þ

From the neutron fluxes at entrance and exit of Ga2O3 sample measured in this work, the value of l of sample target was determined. By using this value the neutron attenuation along the each segment was calculated. The average of the mean values in the four segments was used to determine cross section. The neutron attenuation along the Ga2O3 was estimated to 13.72%. The average flux value obtained by considering the exponential degradation in gallium oxide is 99.83% consistent with that of simple arithmetic mean of the measured entrance and exit values. 2.5. Data corrections The data were corrected for efficiency, coincidence loss, neutron and gamma attenuation. The uncertainties involved in all parameters needed to convert the count rates into decay rates were considered to deduce uncertainty in cross section. Due to weak activity, the uncertainty due to random coincidences as well as pulse pile up loss was negligible. The effect of real coincidences was also considered. The correction factor for gamma-ray attenuation in the sample at a given gamma-ray energy at a fixed geometry for the case of a cylinder, coaxially positioned with the detector was calculated by considering both the attenuation coefficients for gallium and oxygen elements using the following equation.

Fg ¼

lx 1  elx

ð5Þ

where l is the linear attenuation coefficient (cm1) and x is the sample thickness (in cm). The mass attenuation coefficient (l/q; cm2/g) for the gallium and oxygen elements was collected from the Handbook of Chemistry and Physics [22]. The total mass attenuation coefficient (cm2/g) for the compounds (Ga2O3) was converted to linear attenuation coefficient (cm1) by multiplying with the density, q (g/cm3) of sample. The correction factors for gamma-ray attenuation along the gallium oxide pellet were found to be 1.031 and 1.035 for the 834.03 and 629.96 keV gamma-rays emitted from the decay of the 72Ga radionuclide, respectively. The activities of the product radionuclides were weak. The irradiated samples, therefore, were counted directly on the surface of the detector to have good counting statistics. In this case, the correction factor to an efficiency loss of 7.5% for extended sample was taken into account.

Table 3 Experimental cross section for the special energies.

71

Ga(n,)72Ga reaction. Bold means our work and

Year

References

Neutron energy (eV)

Neutron capture cross section (b)

2014 2008 2012 2011 2003 2000 1984 1975 1960 1952 1947

This work Uddin et al. [13] Krane [7] Son et al. [23] Karadag et al. [2] Long et al. [6] Koester et al. [5] Gleason [1] Lyon [8] Pomerance [3] Seren [4]

0.0334 0.0536 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253

3.42 ± 0.27 2.75 ± 0.14 4.12 ± 0.18 4.45 ± 0.25 4.41 ± 0.18 4.62 ± 0.09 3.67 ± 0.10 4.40 ± 0.20 6.18 ± 0.62 4.90 ± 0.40 3.36 ± 0.67

table. The measured value at 0.0334 eV is shown in Fig. 3 together with the values of 0.0253 eV [1–8,23] and 0.0536 eV neutrons [13] and the data in the ENDF/B-VII [21], JENDL-4 [24] and TENDL-2012 [25] files. A number of authors have reported experimental thermal neutron capture cross section for the 71Ga(n,c)72Ga reaction and those values are at 0.0253 eV neutron energy [1–8]. Their reported values varied from 3 to 6 b. Most of these values are very old and it is very difficult to extract precious value from these reported values. At 0.0334 eV neutron, the cross section of the 71Ga(n,c)72Ga reaction measured in this work is 3.42 ± 0.27 b. This value converts to about 4.0 b at the energy of 0.0253 eV assuming the cross sections to follow the 1/v dependence in the thermal region. Recently, Krane [7] reported cross section of the above reaction at energy of 0.0253 eV amounted to 4.12 ± 0.18 is consistent with this work. The value measured in this work at 0.0334 eV is very close to the data curve (Fig. 3) of JENDL-4 library based on calculation using the 1/v relationship. In previous, the cross section of this reaction at 0.0536 eV was measured by Uddin et al. [13] in the same laboratory, which is also in excellent consistent with this library. The precise experimental data at both the 0.0334 and 0.0536 eV energies have, therefore, established the 1/v correlation firmly for the above reaction. The excitation functions in thermal energy region of the ENDF/B-VII and TEND-2012 data files are about 17% and 14%, respectively, higher than that of the data point of this work. Therefore, the measured value at 0.0334 eV will be useful for validating the evaluated data of these two data files.

3. Results and discussion Experimental investigation on cross section of the nuclear reaction, 71Ga(n,c)72Ga, was performed using the 0.0334 eV neutrons. The sample preparation and reactor irradiations were done properly. The neutron flux density as well as the neutron flux gradient was determined with adequate precision and accuracy. The radioactivity of the radionuclides produced both in Ga-target and Aumonitor foil was measured precisely via c-ray spectrometry. Thereby all the necessary precautions and corrections regarding pile-up and coincidence losses, self-absorption in the source, efficiency of the detector, proper use of the decay data, etc. were taken into account. The cross section value obtained will take place as a new experimental reference point. In order to validate the measured value, we constructed the excitation function for the above reaction, filling in the data available at other energies. The new value shall be certainly welcomed with enthusiasm by the evaluators and theorists. The cross section of the 71Ga(n,c)72Ga reaction measured in this work at 0.0334 eV and that of at 0.0253 eV reported in literature are collected in Table 3. The value at 0.0536 eV obtained in the previous work of our laboratory reported in [13] is also given in the

Fig. 3. Neutron capture cross sections for the 71Ga(n,c)72Ga reaction; (a) Solid circle-this work at 0.0334 eV and (b) solid triangle at 0.0536 eV (previous work, Uddin et al. [13]).

N. Afroze et al. / Nuclear Instruments and Methods in Physics Research B 336 (2014) 1–5

4. Conclusion The cross section of the 71Ga(n,c)72Ga reaction at 0.0334 eV neutron was measured; it is amounted to 3.42 ± 0.27 b. As far as we know, the measured value at 0.0334 eV is the first experimental value. The new value was compared with those given in three data files, ENDF/B-VII, JENDL-4 and TENDL-2012, derived theoretically or from systematic. The new value is certainly interesting to the evaluators and theorists. In short the result presented here constitutes a very good contribution to the field of nuclear data. Acknowledgments The authors thank to the Director, Reactor Operation and Maintenance Unit (ROMU) and reactor operation crew for their help in performing irradiations. The authors are highly grateful to Dr. S.M. Yunus, Director of the Institute of Nuclear Science and Technology and to the Head Dr. Imtiaz Kamal and other scientists of the Reactor and Neutron Physics Division of the same institute, for their cordial advice in carrying out this research. The authors gratefully acknowledge Prof. Dr. S.M. Qaim of Research Center Juelich, Germany for his valuable advice and suggestions to properly present this work. References [1] G. Gleason, Thermal neutron cross sections and (n, c) resonance integrals. Part I, Radiochem. Radioanal. Lett. 23 (1975) 317. [2] M. Karadag, H. Yucel, M. Tan, M. Ozmen, Measurement of thermal neutron cross-sections and resonance integrals for 71Ga(n,c)72Ga and 75As(n,c)76As by using 241Am–Be isotopic neutron source, Nucl. Instr. Meth. Phys. Res. A 501 (2003) 524. [3] H. Pomerance, Thermal neutron capture cross sections, Phys. Rev. 88 (1952) 412. [4] L. Seren, H.N. Friedlander, S.H. Turkel, Thermal neutron activation cross sections, Phys. Rev. 72 (1947) 888. [5] L. Koester, K. Knopf, W. Waschkowski, A. Kluever, Interactions of neutrons with gallium and its isotopes, J. Zeits. Fur Phys. Sec. A 318 (1984) 347. [6] H. Long, G. Gang, Y. Xiang, L. Lin, Z. Rong, Measurements of thermal neutron capture cross section, At. Energy Sci. Technol. 34 (2000) 456. [7] K.S. Krane, The decays of 70,72Ga to levels of 70,72Ge and the neutron capture cross sections of 69,71Ga, Appl. Radiat. Isot. 70 (2012) 1649. [8] W.S. Lyon, Reactor neutron activation cross sections for a number of elements, Nucl. Sci. Eng. 8 (1960) 378.

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