Page 117 Integral cross section measurements using TRIGA reactor and Am/Be neutron source M.S. Uddin 1*, S.M. Hossain 1, M.R. Zaman 2, S. Sudár 3 and S.M. Qaim 4 1 Institute of Nuclear Science and Technology (INST), Atomic Energy Research Establishment, Savar, GPO Box-3787, Dhaka-1000, Bangladesh 2 Department of Applied Chemistry and Chemical Engineering, University of Rajshahi, Bangladesh 3 Institute of Experimental Physics, Debrecen University, H-4001 Debrecen, Hungary 4 Institut für Neurowissenschaften und Medizin, INM-5: Nuklearchemie, Forschungszentrum Jülich, D-52425 Jülich, Germany Abstract : *Corresponding author: md.shuzauddin@yahoo.com The spectrum of fast neutrons having energies from 0.5 to 20 MeV in the core of the 3 MW TRIGA Mark II reactor at Savar, and the neutron spectrum over the energy range of 1.5 to 11 MeV of the Am/Be source at Rajshahi were unfolded by activating the metal foils Al, Ti, Fe, Ni, Nb, Mo and In, thereby inducing threshold nuclear reactions covering the whole energy range of the spectrum, and then performing necessary iterative calculations utilizing the activation results and the code SULSA. The two neutron sources were then conveniently used for integral cross section measurements of the neutron threshold reactions 27 Al(n, ) 24 Na, 46 Ti(n,p) 46 Sc, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc, 54 Fe(n,p) 54 Mn and 64 Zn(n,p) 64 Cu. The result of integral measurement was compared with the integrated cross section, obtained from the known excitation functions of those reactions (given in several evaluated data files) averaged over the respective neutron spectrum. Keywords: TRIGA Mark II research reactor, Am/Be neutron source, Neutron Spectrum unfolding, Spectrum average cross section. 1. INTRODUCTION Among the various types of nuclear research reactors operating today, the TRIGA reactors occupy a special place because of their enhanced inherent safety and provision for high thermal neutron fluxes. The nuclear reactor TRIGA Mark II at Savar has been used till now for medical radioisotope production, neutron activation analysis, neutron scattering, neutron radiography, etc. For all those areas of work purely thermal neutron fluxes are preferable. So far not much research related to fast neutrons has been performed, partly due to a lack of knowledge on the fast neutron spectrum available in the reactor core. We undertook to characterize the fast neutron component of the spectrum. Recently an Am/Be neutron source has been installed at Rajshahi University. Though the neutron spectrum is generally known [cf.1], a careful check in case of a new facility is necessary. Since fast neutrons can induce neutron threshold reactions, we decided to measure integral cross sections of several such reactions. Several data libraries [2-5] report evaluated excitation functions of fast neutron interactions with materials like Ti, Ni, Mo, Fe, Zn, etc. The integral cross section measurements in a well-defined neutron field could serve as a test of reliability of evaluated excitation functions. 2. MATERIALS AND METHOD High purity foils of several metals, namely Al, Ti, Fe, Ni, Nb, Mo and In (Goodfellow; purity >99 %) were cut in circular discs with diameter of 1 cm and placed in an Al-container, and finally in an irradiation vial. All foils in a stack were irradiated together in the reactor with neutrons at the dry central thimble (DCT). The irradiation was performed for 40 min at 1 MW. A short 10 min irradiation of In foil was also performed in the reactor core at the same power. Samples for irradiations with neutrons of Am/Be source were prepared by cutting discs of 2 cm diameter and irradiated for 300 h. The details on the foils irradiated and the investigated activation products are given in [cf. 6,7]. The decay data were taken from the NUDAT database (National Nuclear Data Center, BNL, USA) [8]. The radioactivity induced in each irradiated foil was measured non-destructively using a high-purity germanium (HPGe) -ray detector associated with a digital gamma spectrometry system (ORTEC DSPEC jr TM ) and Maestro data acquisition software at the Institute of Nuclear Science and Technology (INST), Savar, Dhaka, Bangladesh.
Page 118 The samples irradiated in the core of the reactor were counted several times over a period of a few months to check the half-lives of the activation products. All the samples were counted at 20 cm from the detector surface, where the sample-size effect on the efficiency is negligible. At that counting position, the probability of coincidence losses are expected to be negligible. The radioactivity produced in the irradiation at Am/Be source was also measured at the Institut für Nuklearchemie, Forschungszentrum Jülich, Germany, using a low-level HPGe detector as described in [7]. Due to weak activities, the samples were placed directly on the surface of the detector, where an efficiency loss for the extended sample and loss of counts due to real coincidences had to be corrected. For this purpose, an independently prepared active enough sample was also counted at 10 cm from the detector surface, where both the sample-size effect on the efficiency and the coincidence loss were negligible. The obtained activity at 10 cm was considered as standard value, using which the detector efficiency for the extended sample on the surface combined with the coincidence loss was calculated. The analysis of the -ray spectra was done using the GammaVision software (Version 6.01). The activities of a number of threshold reactions [6,7], covering the full energy range from threshold to 20 MeV for the reactor and up to 11 MeV for the Am/Be neutron source, served as the basic data for the calculation. They were given as inputs in the iterative code SULSA [9] for unfolding. Uddin et al. [6,7] have reported details on fast neutron spectrum unfolding of Am/Be neutron source and TRIGA Mark II nuclear research reactor. The 58 Ni(n,p) 58 Co reaction induced in the Ni-monitor foil was used to measure the total fast neutron flux in reactor for energies above 0.5 MeV. At TRIGA reactor the cross section value averaged for the unfolded spectrum (0.5 20 MeV) was calculated using the known excitation function of the reaction; it amounted to 91.2 5.5 mb [6]. In the case of Am/Be source, the same reaction was used as flux monitor. By using a value of 391 19 mb [7] for the spectrum-averaged cross section of this reaction, the flux of neutrons of the Am/Be source above 1.5 MeV was calculated. From the measured count rate of each product radionuclide and the flux of neutrons, the neutron-spectrum averaged cross section <σ> of each investigated reaction, namely 27 Al(n, ) 24 Na, 46 Ti(n,p) 46 Sc, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc, 54 Fe(n,p) 54 Mn and 64 Zn(n,p) 64 Cu, was deduced by using the usual activation formula. The combined uncertainty in the experimentally determined cross section was estimated by taking the square root of the quadratic sum of the individual uncertainties which were very similar to those described earlier [6,7,10,11]. 3.1 Characterization of the fast neutron spectrum 3. RESULTS AND DISCUSSION The details on unfolding of fast neutron spectrum of both the TRIGA reactor and Am/Be source are reported in [6,7]. Here, only a brief discussion is given. The unfolded neutron spectrum over the energy range of 0.5 MeV to 20 MeV at the DCT of TRIGA Mark II reactor for 1 MW power is shown in Fig. 1. We ascribe an uncertainty of about 6 % to the various regions of the spectrum. For comparison, the 235 U-fission spectrum [12] is also shown. It appears instructive to compare the fast neutron spectrum of a TRIGA reactor determined in this work with the high-energy part of a pure 235 U-fission spectrum [cf. 12], also shown in Fig. 2 after normalization to a power of 1 MW. Whereas the energy region above 1 MeV coincides well with the TRIGA reactor spectrum, in the energy range between 0.1 and 1.0 MeV, there is considerable deviation. The TRIGA spectrum is somewhat higher in the energy region between 0.5 and 1.5 MeV than the pure 235 U-fission spectrum. There could be several reasons, e.g. difference in fuel cladding, presence of ZrH 1.6, lower enrichment of 235 U in TRIGA (i.e. higher concentration of 238 U), presence of a thick graphite plate in the vicinity of the core, which thermalizes and reflects neutrons, etc.
Page 119 Fig.1: Fast neutron spectrum of TRIGA Mark II reactor unfolded at the irradiation position DCT and 1 MW power. For comparison, the pure 235 U-fission spectrum [12] is also shown. Fig.2: Fast neutron spectrum of Am/Be source: solid line- unfolded in this work and dashed line-reported by Lövestam et al. [1].
Page 120 Table -1. Measured and calculated neutron spectrum averaged cross sections Spectrum-averaged cross section < > (mb) Reaction TRIGA Mark II Reactor (0.5 20 MeV) < > meas (This work) < > Recom (Calamand) ENDF/B-VII.0 IRDF-2002 JEFF-3.2 JENDL-4.0 Am/Be neutron source (1.5 11 MeV) < > meas (This work) ENDF/B-VII.0 IRDF-2002 JEFF-3.2 JENDL-4.0 < > meas < > meas < > meas < > meas < > meas < > meas < > meas < > meas 27 Al(n, ) 24 Na 0.67 0.06 0.73 0.05 0.71 1.07 0.70 1.04 0.67 1.00 0.69 1.02 9.7 1.2 10.2 1.05 10.3 1.06 10.2 1.05 9.7 1.00 46 Ti(n,p) 46 Sc 11.6 0.95 12.5 0.90 11.24 0.97 10.33 0.90 10.4 0.90 11.0 0.95 100 10 99 0.99 99 0.99 93 0.93 99 0.99 47 Ti(n,p) 47 Sc 16.7 1.37 20 2.30 15.4 0.92 15.0 0.90 16.2 0.90 14.9 0.90 63 6 62 0.98 64 1.01 65 1.03 62 0.98 48 Ti(n,p) 48 Sc 0.28 0.023 0.31 0.03 0.28 1.00 0.28 1.00 0.27 0.97 0.28 1.00 4.6 0.6 4.6 1.00 4.8 1.04 4.4 0.96 4.6 1.00 54 Fe(n,p) 54 Mn 72.5 6.00 82.5 5.00 66.6 0.92 66.5 0.92 66.05 0.91 67.44 0.93 311 26 314 1.01 312 1.00 314 1.01 314 1.01 64 Zn(n,p) 64 Cu 28.9±2.0 29.30 1.01 31.0 1.07 29.4 1.02 29.4 1.02 143±13 141 0.99 143 1.00 148 1.03 142 0.99
Page 121 The unfolded neutron spectrum of the Am/Be source is shown in Fig. 2 together with the standard spectrum given by Lövestam et al. [1]. The shapes of the two spectra above 1.5 MeV are the same. The magnitudes, on the other hand, differ by about 20 % between 2 and 4 MeV. However, the cross section of the monitor reaction 58 Ni(n,p) 58 Co averaged for our spectrum agreed within ± 6 % with the averaged value for the standard spectrum [1]. This deviation is within the estimated uncertainty in the monitor cross section. The part of the neutron spectrum below 1.5 MeV is rather uncertain and the region below 0.37 MeV is unknown. It is very hard to characterize the low energy part of the spectrum because there are very few monitor reactions available. For validation of excitation functions of neutron threshold reactions, however, that part of the spectrum is unimportant. 3.2 Spectrum averaged cross sections A number of threshold reactions have been studied. For comparison, we report here only those reactions which are common for TRIGA reactor and Am/Be source. The experimentally determined integral cross sections, averaged over the fast neutron spectrum of both TRIGA reactor and Am/Be source are given in Table 1. The total uncertainty of each value amounts to about 9 %. The integrated reaction cross sections obtained from the curves of evaluated data files and their ratios to the measured values are also given in Table 1. Our measured integral data in the TRIGA reactor are also compared with the normalized and recommended values for the pure fission spectrum (cf. Calamand, [13]). Our measured integral values are generally somewhat lower than the Calamand recommended values (within 8 15 %). This is interpreted to be due to the difference between the TRIGA neutron spectrum and the pure fission spectrum below about 1.5 MeV. The details of comparison between our measured values and the integrated values for several reactions are reported in [6,7]. Here, only for six reactions, namely 27 Al(n,α) 24 Na, 46 Ti(n,p) 46 Sc, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc, 54 Fe(n,p) 54 Mn and 64 Zn(n,p) 64 Cu, the <σ> cal. /<σ> meas. ratio for the evaluated data libraries, ENDF/B-VII.0, IRDF- 2002, JEFF-3.2 and JENDL-4.0 lies mostly between 0.93 and 1.07, showing that the agreement between the integrated data and the integrally measured data is within ±7%. Only in a few cases a deviation up to 10 % was observed. This agreement reflects the status of the excitation functions of the six reactions considered. 4. CONCLUSION In most of the cases, the measured integral data were found to be in good agreement (within 4 %) with the data integrated from the evaluated excitation functions. Only in a few cases a deviation up to 10 % was observed. This reflects the good status of the excitation function. Thus the two fast neutron fields characterized in this work appear to be very suitable to check the reliability of the evaluated excitation functions of neutron threshold reactions. ACKNOWLEDGEMENTS We thank Professor H. H. Coenen of Juelich, Germany, for his support of this cooperation, and Dr. I. Spahn and Mr. S. Spellerberg, also at Juelich, for some experimental help. The authors thank the operation crew of the TRIGA Mark II nuclear reactor, Savar, Dhaka, for their help in performing irradiation of the samples. M. S. Uddin specially thanks the Alexander von Humboldt Foundation in Germany for financial support to conduct this research work in Bangladesh, and the Radiation Testing and Monitoring Laboratory, Chittagong, for granting him leave of absence to conduct this experiment at BAEC facilities at Savar, Dhaka.
Page 122 REFERENCES 1. G. Lövestam, M. Hult, A. Fessler, J. Gasparro, P. Kockerols, K. Okkinga, H. Tagzirla, F. Vanhavere and J.S.E. Wieslander, Radiat. Meas. 44, 72-79 (2009). 2. O. Bersillon, L.R. Greenwood, P.J. Griffin, W. Mannhart, H.J. Nolthenius, R. Paviotti-Corcuera, K.I. Zolotarev, E.M. Zsolnay, P.K. McLaughlin and A. Trkov, International Reactor Dosimetry File 2002 (IRDF-2002), International Atomic Energy Agency, Vienna, Austria, Technical Reports Series No. 452, December 2006. 3. ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology, Nucl. Data Sheets 107, 2931 (2006). 4. Joint Evaluated Fission and Fusion File, Version 3.2, released 2013, http://www.oecd-nea.org/db forms/data/eva/evatapes/jeff_31/jeff32. 5. K. Shibata, O. Iwamoto, T. Nakagawa, N. Iwamoto, A. Ichihara, S. Kunieda, S. Chiba, K. Furutaka, N. Otuka, T. Ohsawa, T. Murata, H. Matsunobu, A. Zukeran, S. Kamada and J. Katakura, JENDL-4.0: A New Library for Nuclear Science and Engineering, J. Nucl. Sci. Technol. 48 (1), 1 (2011). 6. M.S. Uddin, S. Sudár, R. Khan, S.M. Hossain, M.A. Zulquarnain and S.M. Qaim, Radiochim. Acta 101, 613 (2013). 7. M.S. Uddin, M.R. Zaman, S.M. Hossain, I. Spahn, S. Sudár and S.M. Qaim, Appl. Radiat. Isot. 68, 1656 (2010). 8. NUDAT 2.6 Decay Radiation Database, National Nuclear Data Center, Brookhaven National Laboratory (2009), http://www.nndc.bnl.gov/nudat2. 9. S. Sudár, A solution for the neutron spectrum unfolding problem without using input spectrum. IAEA- INDC (HUN)-026/L (1989) IAEA, Vienna, Austria. 10. M.S. Uddin, M.R. Zaman, S.M. Hossain and S.M. Qaim, Radiochim. Acta 102, 473 (2014). 11. M.S. Uddin, I. Spahn, S.M. Hossain, M.Rumman. Zaman, M.Rakib. Zaman and S.M. Qaim, Radiochim. Acta (2014) (in press). 12. R.L. Walsh, NEANDC(E) 241/L; INDC (FR) 70/L (1987). 13. A. Calamand, Handbook on Nuclear Activation Cross Sections. Technical Report No. 156, IAEA, Vienna, (1974) p.273.