Neutron capture cross-section measurement for the 186W(n,γ)187W reaction at 0.0536 eV energy

ARTICLE IN PRESS

Applied Radiation and Isotopes 66 (2008) 1235–1239 www.elsevier.com/locate/apradiso

Neutron capture cross-section measurement for the reaction at 0.0536 eV energy

186

W(n,g)187W

M.S. Uddina,, M.H. Chowdhuryb,c, S.M. Hossaina, Sk.A. Latifa, M.A. Hafiza, M.A. Islama, A.K.M. Zakariaa, S.M. Azharul Islamc a

Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Savar, GPO Box No. 3787, Dhaka-1000, Bangladesh b Department of Physics, Comilla Victoria Government College, Comilla, Bangladesh c Department of Physics, Jahangirnagar University, Savar, Dhaka, Bangladesh Received 27 November 2007; received in revised form 27 January 2008; accepted 29 January 2008

Abstract The thermal neutron-induced activation cross section for the 186W(n,g)187W reaction was measured at 0.0536 eV neutron energy using TRIGA Mark-II research reactor, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The 197Au(n,g)198Au monitor reaction induced in a high-purity gold foil was used to determine the effective neutron beam intensity. The activities induced in sample and monitor foils were measured nondestructively by a high-resolution HPGe g-ray detector. The present experimental cross-section value is the first one at 0.0536 eV. The obtained new cross section that amounts to 26.671.6 b is 2% higher than the recently reported data in ENDF/B-VII and 5% lower than that of JENDL-3.3. r 2008 Elsevier Ltd. All rights reserved. Keywords: Reactor; Thermal neutron; 0.0536 eV energy; Cross section; Activation technique

1. Introduction Tungsten is a structural material used in various parts of fusion reactors. Nuclear reactors have been considered promising as long-lived heat sources for light-weight space powder systems. To meet the requirements of compactness and high-temperature operation, tungsten has been suggested for use both as a fuel-element material and as a shielding material. To evaluate the merits of tungsten for these applications, accurate and complete nuclear data are required, particularly for neutron capture cross sections (Shook and Bogart, 1968). Because of the large temperature difference between reactor start-up and operation, the energy dependence of the capture cross section must be known over an appreciable energy interval. Particularly, the thermal neutron-induced activation cross section for the 186W(n,g)187W reaction is of great importance not only for design and development of nuclear reactors, but also Corresponding author. Tel.: +880 171 5363326.

E-mail address: [email protected] (M.S. Uddin). 0969-8043/$ - see front matter r 2008 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2008.01.013

for the activation analysis and other theoretical and experimental studies concerning the interaction of neutrons with tungsten. Up to now, the capture cross sections have been measured with various neutron sources, based on reactors, Van-de-Graaff accelerators and electron linear accelerators. A number of authors have reported that the cross section for this reaction at 0.0253 eV neutron energy and a large discrepancy is present among them. In most of these studies, thermal neutron cross sections were determined experimentally by the activation method using cadmium ratios of the investigated material and a reference material (monitor), for which usually gold is employed. Karadag and Yucel (2004) reported the thermal neutron cross section for the 186W(n,g)187W reaction, which was measured by the activation method using the 55Mn(n,g)56Mn reaction as a single comparator, where the irradiation was performed in an isotropic neutron field of the 241Am–Be neutron sources. In these studies, neutron beams with broad spectrum (thermal, epithermal and fast energies) were used for irradiation of target.

ARTICLE IN PRESS M.S. Uddin et al. / Applied Radiation and Isotopes 66 (2008) 1235–1239

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The present work was undertaken to measure the cross sections of the 186W(n,g)187W reaction only in the interaction of monoenergetic thermal neutrons at 0.0536 eV with natural tungsten using the 3 MW TRIGA Mark-II research reactor at the Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. To achieve 0.0536 neutron energy, it must have to arrange a special facility with monochromator outside the reactor. The activation technique can be employed to obtain capture cross sections, because the decay scheme of the product nucleus is known and its states with suitable half-life are present. The results were compared with the available reported experimental data and evaluated data of ENDF/B-VII (2007) and JENDL-3.3 (2002). 2. Experimental technique 2.1. Neutron source The neutrons used in this experiment were generated in TRIGA Mark-II reactor at the Atomic Energy Research Establishment, Dhaka, Bangladesh. The TRIGA is a research reactor having a maximum continuous thermal power output of 3 MW. The experimental facilities around TRIGA reactor include four neutron beam tubes, namely tangential, piercing, radial-1 and radial-2 beam port. In the present experiment, the piercing beam port was utilized. The neutrons coming out of the reactor through this beam port are of various wavelengths. They have been monochromatized before sending them on the target for experiment. This can very effectively be done using a suitable single crystal of Cu(2 0 0). A schematic diagram of the arrangement for monochromatization of reactor neutrons and the experimental setup are shown in Fig. 1. The reactor neutrons (thermal and epithermal) of piercing beam port are coming through a collimator with sufficient shielding to the monochromator, which is also surrounded

Inpile Collimator Reactor Beam (Thermal+epithermal) Reactor

Shield Collimator Au-monitor

Monochromator

W-Target Monochromatic Beam (Thermal, 0.0536 eV)

Fig. 1. A schematic view of the monochromatic system and experimental arrangement for irradiation.

with enough shielding as shown in Fig. 1. The monochromatic neutrons escape through another collimator to target for irradiation. A monochromatic neutron beam of wavelength l ¼ 1.236 A˚ obtained by Bragg reflection from a Cu (2 0 0) monochromator was used for irradiation of target materials. The corresponding neutron energy at this wavelength is 0.0536 eV. 2.2. Sample irradiation A foil of tungsten (1.2 cm diameter  200 mm thick; purity 99.99%) of natural isotopic composition: 180W(0.1270.01%), 182 W(26.5070.16%), 183W(14.3170.04%), 184W(30.647 0.02%), and 186W(28.4270.19%) and thin gold foils (1.2 cm diameter  25 mm thick; purity 99.99%) were irradiated with neutrons of 0.0536 eV energy for 5 h. Gold foils were used to measure the effective neutron flux using the 197Au(n,g)198Au monitor reaction (68.571.7 b at 0.0536 eV (Yamamoto et al., 1996; Pavlenko and Gnidak, 1975; Haddad et al., 1964)). Two gold foils of approximately same size and weight were attached at the front and back of the tungsten foil, respectively to check the difference in neutron beam intensity between entrance and exit of the target. To follow the effect of epithermal neutrons and fast neutrons, the cadmium-covered gold foil and bare aluminum foil were also irradiated. 2.3. Gamma-ray measurement and data analysis The activities of the radioisotopes produced in the target and the monitor foils were measured nondestructively using high-purity germanium (HPGe) gamma-ray spectroscopy (Canberra, 15% relative efficiency, 1.8 keV resolution at 1332.5 keV of 60Co) coupled with a digital gamma spectrometry system (ORTEC DSPEC jr TM) and Maestro data acquisition software. The spectrum analysis was done using the program GammaVision 5.0 (EG&G Ortec) and Hypermat PC software. Measurements were started about 10 h after the end of irradiation. Each sample was recounted three times, giving enough intervals to avoid disturbance by overlapping gamma-lines from undesired sources and to determine the experimental cross-section value with adequate precision and accuracy. The thickness of the material used in the present experiment was limited for the self-absorption of captured gamma rays in the sample. If this absorption becomes large, the fraction of the gamma rays escaping may become a function of the neutron capture cross section of the sample. This arises from the fact that a capture event occurring near the front face (which is more likely when the cross section is relatively high) can have an appreciably higher probability of being detected than one that originates in the interior due to the exponential nature of the gamma attenuation. The efficiency versus energy curve of the HPGe gammaray detector for the counting distance was determined using the standard point sources, 133Ba, 109Cd, 22Na, 60Co, 57 Co, 54Mn and 137Cs. The neutron beam intensity was determined from the measured activities induced in gold

ARTICLE IN PRESS M.S. Uddin et al. / Applied Radiation and Isotopes 66 (2008) 1235–1239

monitor foils using the 197Au(n,g)198Au monitor reaction. The cross section of the monitor reaction, 68.5 71.7 b at 0.0536 eV, was evaluated from the trend line of experimental data reported by Yamamoto et al.(1996), Pavlenko and Gnidak (1975) and Haddad et al.(1964). The monitor foils were irradiated simultaneously and measured with the same detector and in a comparable geometry as the tungsten target. The gamma-ray count rates were converted to decay rates by correcting for the gamma-ray intensities and the efficiency of the detector using the following formula: R¼

NI g

eltc ð1

lC ,  eltm Þ  ð1  elti Þ

(1)

where l is the decay constant, s1; C is the total counts of gamma-ray peak area; N is the number of target atoms, atom; e is the peak efficiency; Ig is the branching ratio of gamma ray; tc is the cooling time, s; tm is the counting time, s; ti is the irradiation time, s. If the neutron has a pure monoenergetic spectrum, the cross section at the peak neutron energy s(Epeak) can simply be obtained as sðE peak Þ ¼

R , fðE peak Þ

(2)

where, f(E) ¼ neutron fluence, n cm2 s1. The decay data of the radioactive products were taken from the NUDAT database (National Nuclear Data Center, information extracted from the NuDat database, http://www.nndc.bnl.gov/nudat2). The decay data are quoted in Table 1. The uncertainties considered in order to derive the total uncertainty in cross section are quoted in Table 2. It should have to be mentioned that the calculated uncertainty in the neutron flux is 4%. The overall uncertainty in the cross section is around 6% (1s). The neutron flux was determined individually for both of two Au-foils irradiated at the front and back of W-foil, and the obtained values are 1.423  105 and 1.408  105 n cm2 s1, respectively. Therefore, the neutron absorption in W-foil during irradiation is about 1%. No peaks for the 198Au and 24Na [27Al(n,a)24Na] radionuclides were found in gamma-ray spectrum in the irradiated Cd-covered gold and aluminum foils, respectively. This confirms the negligibility of the epithermal and fast neutrons in the used beam during irradiation.

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3. Results and discussion Thermal neutron capture cross section of the W(n,g)187W reaction was measured by activation technique at 0.0536 eV neutron energy. The result is given in Table 3 together with other experimental values at

186

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 g-ray counting Decay branching ratio Efficiency calibration Sample mass Isotopic abundance Half-life Cross section of the monitor reaction

1–2 3 3 0.01 0.19 0.06 2.5

Total

6

Table 3 Neutron capture cross section for the

186

W(n,g)187W reaction

References (year)

Neutron energy (eV)

Neutron capture cross section (b)

This work, 2007 ENDF/B-VII, 2007 Mughabghab, 2006 Mughabghab, 2006 Karadag and Yucel, 2004 JENDL-3.3.2002 Kafala et al., 1997 De Corte and Simonits, 1989 Knopf and Waschkowski, 1987 Simonits et al., 1984 Anufriev et al., 1981 Heft, 1978 Gleason., 1977 Erdtmann, 1976 Hogg and Wilson, 1970 Damle et al., 1967 Gillette, 1966 Friesenhahn et al., 1966 Lyon, 1960 Pomerance, 1952 Seren et al., 1947

0.0536 0.0253 0.0536 0.0253 0.0253 0.0253 0.0253 0.0253

26.671.6 37.5 26.6a70.5 38.170.5 39.572.3 39.6 42.870.8 38.771.9

0.0253

38.570.8

0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253 0.0253

3771.8 3773 36.670.8 3771.5 37.871.5 4071.5 35.470.8 33 37.871.2 41.3 34.172.7 34.276.8

a

Expected 1/v cross section based on Mughabghab’s evaluation.

Table 1 Decay characteristics of the investigated nuclides and the contributing reactions Nuclear reaction

Half-life

Gamma-ray energy (keV)

Branching ratio (%)

186

23.7270.06% h

197

2.69572.1E-4% d

479.5570.022% 685.7370.04% 411.871.7E4%

21.870.7% 27.370.9% 95.5

W(n,g)187W Au(n,g)198Au

ARTICLE IN PRESS M.S. Uddin et al. / Applied Radiation and Isotopes 66 (2008) 1235–1239

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479.55 keV (187W)

1240

Counts

930

685.73 keV (187W)

0.0253 eV. The uncertainty in the measured cross section is also given. The 187W radionuclide was identified by the 479.55 and 685.73 keV individual gamma-lines. For example, a spectrum for the radioactive tungsten target is shown in Fig. 2. The cross-section values determined individually for these gamma-lines are consistent with each other. Up to now, a number of authors reported thermal neutron capture cross section for the 186W(n,g)187W reaction at 0.0253 eV. Recently, evaluated data are also reported in ENDF/BVII, (2007) and JENDL-3.3 (2002) libraries. In these references, a mixed neutron beam of thermal and epithermal energies was used for activation. They corrected the activation due to the epithermal neutrons by the Cd cut-off energy technique, i.e., the cadmium ratio foil activation method was used for the determination of thermal and epithermal neutron fluxes to which the target materials were exposed. The method involved the irradiations of bare foils and the foils covered with cadmium. Since the neutrons used in irradiation cover a wide range of energy, the bare foil is activated both by thermal and epithermal neutrons. On the other hand, the cadmiumcovered foil is activated only with epithermal neutrons because of high thermal neutron absorption cross section of cadmium. Activity due to thermal flux could, therefore, be derived by simply subtracting the values of cadmiumcovered foils values from those of the bare foils. It is interesting that all the existing experimental values are reported at 0.0253 eV neutron energy. In the present experiment, the neutrons that came out of the reactor were monochromatized before irradiation and contained 0.0536 eV energy. As shown in Fig. 1, a facility was established to obtain only thermal neutrons of 0.0536 eV energy. To measure thermal neutron capture cross section at this energy, both the target and Aumonitor foils were irradiated with the monoenergetic neutron beam with single direction. Therefore, our technique is very easy, where the difficult Cd cut-off method for separation of thermal neutron flux from epithermal is not required. We report thermal neutron capture cross section for the 186W(n,g)187W reaction at

0.0536 eV neutron energy. The evaluated value of ENDF/ B-VII (2007) is about 7% lower than that of JENDL-3.3 (2002). It should have to be mentioned that both these data files have shown the energy dependence of cross section in the thermal energy region. We have collected the value at 0.0536 eV from their excitation function curve to compare with the measured one. The measured value is 5% lower than JENDL-3.3 (2002) and 2% higher than ENDF/B-VII (2007). The deviation of the present result from the evaluated values is within the estimated uncertainty. Friesenhahn et al. (1966) have shown the energy dependence of the capture cross section over the appreciable energy interval 0.01–10 eV. The present value is on the trend of Friesenhahn et al. (1966). A large difference between the values at 0.0253 and 0.0536 eV was found. Neutron capture cross section in the thermal region ordinarily obeys the 1/v law, where v is the speed of incident neutron. Mughabghab’s latest evaluated cross section (Atlas of Neutron Resonances, Elsevier, 2006) is 38.170.5 b at 0.0253 eV energy. Assuming a 1/v crosssection dependence, this gives 26.670.5 b at 0.0536 eV and agrees precisely with the present measured value. 4. Conclusions The neutron capture cross section for the 186W(n,g)187W reaction relative to the 197Au(n,g)198Au monitor reaction was measured at 0.0536 eV neutron energy by activation technique using the TRIGA Mark-II research reactor, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. We report new cross-section amounts to 26.671.6 b, which is consistent with the data of Friesenhahn et al. (1966) and recently reported evaluated data in ENDF/B-VII (2007) and JENDL-3.3 (2002). As far as we know, there are no experimental data available at our investigated energy. So far we are the only one who carried out experiment with monoenergetic pure thermal neutrons with single direction for cross-section measurement. The obtained value will be useful for the proper estimation of radioactivity and radiation safety analysis in fusion reactors. Acknowledgments The authors thank the manager, Reactor Operation and Maintenance Unit (ROMU) and reactor operation crew, for their help in performing irradiations. The authors are highly grateful to the staff of the Reactor and Neutron Physics Division for their cordial help in carrying out the experiment.

620

310

References 0.00

250.00

500.00 Energy (keV)

750.00

1000.00

Fig. 2. Gamma-ray spectrum for the radioactive tungsten.

Anufriev, V.A., Babich, S.I., Kolesov, A.G., Nefjodov, V.N., Poruchikov, V.A., 1981. Neutron resonances of 186W in the energy range up to 300 eV. At. Energy 50 (1), 67.

ARTICLE IN PRESS M.S. Uddin et al. / Applied Radiation and Isotopes 66 (2008) 1235–1239 Damle, P.P., Fabry, A., Jacquemin, R., 1967. Study of the reaction 186W(n,g)187W. Report from Euratom-countries +Euratom to EANDC-76, 107(2). De Corte, F., Simonits, A., 1989. k0-Measurements and related nuclear data compilation for (n,g) reactor activation analysis IIIb: tabulation. J. Radioanal. Nucl. Chem. 133, 43. Erdtmann, G., 1976. Neutron Activation Tables. Verlag Chemie, Weinheim. ENDF/B-VII Library. Database version of 30 October 2007. Friesenhahn, S.J., Haddad, E., Froehner, F.H., Lopez, W.M., 1966. The neutron capture cross section of the tungsten isotopes from 0.01 to 10 electron volts. Nucl. Sci. Eng. 26, 487. Gillette, J.H., 1966. Preprint: ORNL-4013, vol. 5. Oak Ridge National Laboratory. Gleason, G., 1977. Thermal Neutron (n,g) Cross Sections and Resonance Integrals—part 2. Private Communication to NEA-Data Bank. ORL, USA. Haddad, E., Walton, R.W., Friesenhahn, S.J., Lopez, W.M., 1964. A high efficiency detector for neutron capture cross section measurements. Nucl. Instr. and Meth. Phys. Res. 31, 125. Heft, R.E., 1978. A consistent set of nuclear parameter values for absolute INAA. International Conference on Computers in Activation Analysis and Gamma ray Spectroscopy, vol. 495. Mayaguez, Puerto Rico. Hogg, C.H., Wilson, W.L., 1970. Reactor thermal-neutron cross sections of 64Zn, 68Zn, 121Sb, 123Sb, 55Mn, and 186W. Idaho Nuclear Corporation Reports: 1317, p. 53. JENDL-3.3. Japan Atomic Energy Research Institute. Nuclear Data Center. Copyright, 2002 JAERI.

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Karadag, M., Yucel, H., 2004. Measurement of thermal neutron crosssection and resonance integral for the 186W(n,g)187W reaction by the activation method using a single monitor. Annal. Nucl. Eng. 31, 1285. Kafala, S.L., MacMahon, T.D., Borzakov, S.B., 1997. Neutron activation for precise nuclear data. J. Radioanal. Nucl. Chem. 215 (2), 193. Knopf, K., Waschkowski, W., 1987. Interaction of neutrons with tungsten and its isotopes (in German). Z. Naturforsch. A 42 (9), 909. Lyon, W.S., 1960. Reactor neutron activation cross sections for a number of elements. Nucl. Sci. Eng. 8, 378. Mughabghab, S.F., 2006. Resonance Parameters and Thermal Cross Sections, Z ¼ 1–100, Atlas of Neutron Resonances. Elsevier, Amsterdam. Pavlenko, E.A., Gnidak, N.L., 1975. Neutron capture cross section measurement for hafnium and vanadium isotopes. All Union Conf. Neutron Phys. 3, 171. Pomerance, H., 1952. Thermal neutron capture cross sections. Phys. Rev. 88 (2), 412. Seren, L., Friedlander, H.N., Turkel, S.H., 1947. Thermal neutron activation cross sections. Phys. Rev. 72 (10), 888. Shook, D.F., Bogart, D., 1968. Effective resonance integrals of separated tungsten isotopes from reactivity measurements. Nucl. Sci. Eng. 31, 415. Simonits, A., De Corte, F., Elnimr, T., Moens, L., Hoste, J., 1984. Comparative study of measured and critically evaluated resonance integral of thermal cross-section ratios. J. Radioanal. Nucl. Chem. Artic. 81 (2), 397. Yamamoto, S., Kobayashi, K., Fujita, Y., 1996. Application of BGO scintillators to absolute measurements of neutron capture cross sections between 0.01 and 10 eV. J. Nucl. Sci. Tech. 33, 815.