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Neutron spectroscopy by time of flight method and determination of neutron beam December 20 2009 1 000 000 Calculated spectra Measurements of Jan 20 2009 Neutron flux density a.u. 10 000 Prepared by: 100 1 0,01 Sameh Hassan , Yomna Abd El-Moaty 1E-4 1E-4 0,01 1 100 10 000 Energy, eV Supervisors: L Pikelner, V.Shvetsov FLNP, Dubna Summer Student Practice, 2009, JINR Dubna 1 Motivation - Studying the processes of the interaction of slow neutrons with nuclei ( 1- 105 ev) -Radioactive capture with gamma emission is the most common type of reaction at certain energies for slow neutrons. This (n,) reaction often results in product nuclei which are radioactive. For example: 59 1 60 60 Co n Co * Co 27 0 27 27 So it is a method of studying dependence of neutron cross-section on its energy. 2 3 Targets and methods For example to study total neutron cross-sections of tungsten (W) The time-of-flight (TOF) method is used to measure the transmission of the sample Sample Size [mm3] Density [g/cm3] Atomic Mass [a.m.u] Purity [%] W 100x100x0.2 19.3 183.85 99.98 4 5 The neutron Time of Flight (TOF) spectrum 8000 46.26[eV] 183 183 27.03[eV] 6000 Sample (W) Beam Open Beam Background W W 18.8[eV] 186 W Counts 183 7.6[eV] W 4.15[eV] 182 W 4000 2000 0 0 50 100 150 200 250 300 350 400 TOF Channels (2 s/ch) Fig.1 ) The neutron TOF spectrum for sample-in and open beam along with background level of W sample 6 Spectrum processing The total neutron cross-section is determined by measuring the transmission of neutrons through the samples. Thus the neutron total cross-section is related to the neutron transmission rate 1 T(E) as follows: ( E ) ln T ( E ) N N is the atomic density per cm2 in the sample. T (E) [ N BGS] [ N0 BGO] N and N0 are the foreground counts for the sample in (sample beam) and out (open beam), BGS and BGO are the background counts for sample in and out respectively. The atomic density N in the sample can be calculated from the formula N ( t N A ) / A[/ cm2 ] 7 By inserting the values of N and transmissions from the fig.1 we measured the total cross-sections of the entire samples depending on neutron energies. 100000 natural W (n,tot) Present data R.E.SCHMUNK R.E.CHRIEN J.A.HARVEY W.SELOVE ENDF/B-VI Total Cross-Section in barn 10000 1000 100 10 1 0.1 0.01 0.1 1 10 100 Neutron Energy in eV The measured cross-sections are compared with the evaluated ones from ENDF/B-VI and some other published data 8 Detection system Our detection system for Ɣ rays emitted during neutron capture is liquid scintillation detector consists of six photo multiplier tubes surrounding the sample 9 Scintillation detector 10 Transmission measurements Sample Source Flight path L, m dt Detector Collimator L Lm En eV 5.227 10 3 t s ∆E= 2∆t E t ∆L 2 ∆E = 2∆L E L ∆E= 2.77. 10 -2 ∆t(μs) E3/2 L(μs) 11 12 nσ0Г/Δ 13 Breit-Wigner formula for cross section 100000 c E E0 2 2 4 / 2 2.86.10 9 cm E (ev) c Capture cross section :total resonance width 1000 100 10 1 0.1 0.01 :width of neutron resonance n E natural W (n,tot) Present data R.E.SCHMUNK R.E.CHRIEN J.A.HARVEY W.SELOVE ENDF/B-VI 10000 capture Total Cross-Section in barn 2 g n 0 :energy at the center of the resonance 0.1 1 10 100 Neutron Energy in eV 14 Application of neutron cross section 1) Finding the neutron flux at certain energy 2) Determine resonance parameter ГƔ,Гn 0.1E (ev) M 15 Experiment Sample :Ta181 n = 1.5*1021 nuclei /cm2 Time of irradiation:360 min curve 16 Results Δ 4.3ev nσ0Г/ Δ A ∑N n(E)ζƔ flux n(E)ζ/ flux 0.048 33 0.655 27558 4.8.104 7.5 0.64 10.3ev 0.0754 8.5 0.414 10640 2.7.104 3.3 0.82 13.95ev 0.0878 1.776 0.1756 3485 2.1.104 2.6 0.81 17 Measured flux 7 10 6 10 5 10 4 10 2 n/cm /sec/eV 3 10 2 10 1 10 0 10 -1 10 -2 10 -3 10 -4 10 -5 10 10 -4 10 -3 10 -2 10 -1 10 0 10 1 10 2 E, eV 10 3 10 4 10 5 10 6 10 7 18 Conclusion This method is good in measuring flux as we get almost the same fraction in our three resonances The relation between the energy and flux is inversely proportional Resonances are corresponding to the energy states in our sample 19 thanks 20