determination of thermal neutron flux from (pu-be) source of silver foil
TRANSCRIPT
الزعليك محمود د. مفتاحالفيزياء بقسم باحث
المياه وتحليه المتجددة الطاقات بحوث مركز بتاجوراء
)طرابلس( الجماهيرية تاجوراء30878ص.ب. العظمى
بسيبسو مفتاح د. فرج الحسابات مجموعة ورئيس باحث
المفاعل بقسم الهيدروحرارية المياه وتحليه المتجددة الطاقات بحوث مركز
بتاجوراء )طرابلس( الجماهيرية تاجوراء30878ص.ب.
العظمىااللكتروني البريد :
الورقة عنوان النيتروني الفيض توزيع تحديد
ضــتعري من جـالنات راريـــالح ) النيتروني درــللمص الفضة ريحةــش
Pu – Be)
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Determination of Thermal Neutron Flux from (Pu-Be) Source of Silver Foil
تعريض من الناتج الحراري النيتروني الفيض توزيع تحديد (Pu – Be) النيتروني للمصدر الفضة شريحة
الخالصة
Puالمصدر) – Beيمكن الـتي النيترونـات مصـادر أهم من ( يعتـبر
لــه النيــتروني المصــدر الفضــة, وهــذا شــريحة تشــعيع في اســتخدامها
عمـره نصـف فترةولالستعادة, يةالالع القابلية أهمها من عالية مميزات
يمكن النيــترون كثافة شدة لقياسات النسبيةو المطلقة القيمة.طويلة
نيــترونيال لمجــالل الفضــة شــريحة تعــريض يتم عنــدما بدقــة حســابها
بالتنشــيط يعــرف مــا ذاوهــ ،اإلشــعاعي هانشــاط بقيــاس والقيــام
الفيض توزيــع تحديــد هــو البحثية الورقة هذه من الغرض إن اإلشعاعي.
النيتروني لمصدرل الفضة شريحة عيعـتـش من الناتج الحراري النيتروني
(Pu- Be)بجامعــة التشــعيع بمنظومــة الموجود (A.G.H.قســم ) الفيزيــاء
)مسافات(، أبعاد عدة عنـد (n/sec 106 × 5.5) حوالي ينتج والذي النووية
النيتـــروني للـفيض النظرية بالـنتـائج المعملية النتـائـج هـذه بمقـارنةو
(Dydejczyk 1991) بواسطة المـعـد(ANISN CODE) مـن عليـها المتحصل
النيــتروني الفيض توزيــع بين جيــد توافــق هناك بأن تبين جالنتائ , وهذه
من النــاتج النيــتروني الفيض قيم في الفــرق ولكن والنظــري المعملي
الفضــة لشــريحة اإلشــعاعي للنشــاط المطلقــة القيمــة قياس دقة عدم
.القراءات قياس ومعدل
Abstract
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Pu-Be source is one of the most important neutron source can be used for
activation of silver foil. This source has the following characteristics: high
reproducibility and long lifetime. The absolute and relative measurements of
neutron intensities can be precisely determined by activating a foil in the neutron
field and counting its radioactivity.
In the present work the experimentally thermal neutron flux obtained from
activation of silver foil at different distance from the axis of Pu-Be neutron source
with yield 5.5x106 n/s. was compared with theoretical estimated thermal flux
obtained from (ANISN CODE) prepared by (Dydejczyk 1991) the results conduct
good agreement was achieved between the special distribution of the
experimentally determined thermal neutron flux and theoretical estimated thermal
neutron flux. The difference between the values of neutron fluxes is probably due
to absolute measurement of silver foil activity.
1. Introduction
The neutron strength Q (i.e. the number of neutrons emitted by the source
per second) of laboratory neutron source are determined by using primary
standardization techniques such as the manganese activation method or the use of
the long counter. Such sources are extensively used for fast neutron calibration in
the laboratory. For thermal neutron calibration purposes a sigma pile or paraffin
or polyethylene mode rated system is normally used [1]. In the case of exposure to
slow neutrons, the increased blackening beneath a suitable filter such as cadmium,
arising from the (n,γ) reaction in the filter may be used for assessment of slow
neutron doses [2].
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For fast neutron doses a special film-pack which uses a nuclear emulsion
to record recoil proton tracks based on the emulsion is used [3]. The whole
assembly is made light-tight and part of it is covered with a cadmium filter to
absorb slow neutron which would otherwise give rise to confusing proton tracks
resulting from the (n,p) reaction with the nitrogen of the cellulose base. Thus there
are two types of film badge. One is for fast neutrons and the other for other types
of radiation. They are normally worn on the chest to enable a whole body dose
assessment to be made [4].
The biological effects of slow neutrons are considerably smaller than the
effect of neutrons in the energy range from 1-10 MeV. The existence of several
substances with very high capture crosses sections for thermal neutrons. And the
fact that this cross section has the same energy dependence as the slow neutrons
effects in tissues. The dosimetry of slow neutrons is less problematic than the
dosimetry fast neutrons [5].
All nuclear reactions which are triggered by slow neutrons such as (n, α),
(n, p), (n,γ) reactions which can in principle be employed for the detection of this
particles. The α-particles, protons or gamma-ray emitted in theses reactions are
the ones which produce the scintillation in suitable materials.
The problem of gamma-ray discrimination is facilitated by selecting
substance with large slow neutron cross section. The combined energy of the α-
particles and 3H nucleus formed in the reaction is 4.8MeV so that
gamma-ray discrimination is no problem [6].
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The cross section of boron for low neutrons is ten-times as greater as that
of lithium. The pulse height very small in this case because the combined energy
of α- particle and recoil nucleus is only 2.3 MeV and the efficiency is very poor
for the heavy particles [7].
Only detectors, which contain appreciable amounts of hydrogen, can be
expected to be real value for fast neutron dosimetry. All other elements are less
effective since the fractional energy transfer in collision between neutrons and
other atoms decreases rapidly with the atomic weight of the partner. Even
scintillates with highest hydrogen concentration attainable appear rather
inefficient considering the relatively strong effects of fast neutrons in tissues.
Scintillates which contain other light atoms like lithium have been investigated a
possible detectors for fast neutrons but apparently no conclusive results have yet
been reported [8]. The activation equipment describe in forth Arab conference on
peaceful uses of atomic energy, Tunis, 1998 (9).
2. The aim of this work
The fundamental aim of the present work is to determine the thermal
neutron fluxes of silver foil in the proximately of the Pu-Be neutron source, and
compared the experimentally results of thermal flux with the theoretically
estimated thermal flux from ANISN code which prepared by. Dydejczyk 1991.
3. Method of study
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The method used is the absolute method to determination of the thermal
neutron fluxes from the irradiation of silver foil, the absolute measurement of the
induced activity of the silver isotope (108Ag) is known and the detection efficiency
of the detector is also well known. It means the determined of the activity of silver
foil in the neutron field (Pu-Be) source and counting radioactivity. The neutron
activation analysis procedure are shown in Figure 1.
Figure 1. The main steps of the consecutive analysis.
The thermal flux of silver foil determine at different distance from
the source (2.75 cm every 0.5 cm to 5.75 cm) by using this formula [1].
(1)
Where A is the absolute activity of 108Ag, n is the number of radioactive nuclei, σ
is the cross section of (108Ag), which is equal 45 barn, λ is the decay constant of
(108Ag), where λ = Ln2/T1/2 and T1/2 for 108Ag = 2.42 min. The important energies of
the decay mode of silver is (energy) 1.77 MeV (97%), = 0.010 MeV, γ =
0.632 MeV (1.90%) and 0.511 MeV (0.20%) as showing in Figure 2, and Figure 3
shows the relation between the activity and irradiation time.
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Figure 2. The important energies of the decay of Silver (108Ag).
Figure 3. The relation between the activity and time.
4. Activation Procedure.
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The Activation of silver foil measurement of foil activity using a G.M
counter Procedure including the following steps:.
1. Preparation of the activation set-up for the measurement of a silver foil in the
polyethylene block, as shown in Figure 4.
2. The sample or foil of silver was irradiated for ta=12.2 min using Pu-Be
neutron source.
3. Preparation of the counting arrangement as shown in figure6, also G.M
counter was used to measure the background before counting the sample.
4. Measurement of the cooling time tc, the time between the end of irradiation
and the beginning of counting.
5. Placement of irradiated sample in the G.M counter and the performance
measurement.
6. Irradiate the silver foil and measurement of the activity of the foil at various
distances from the source, starting from 2.75 cm every 0.5 cm to 5.75 cm.
7. Determination of the activity of the foil (silver) at different distances from
the source.
8. Determination of the thermal flux at the same distance.
9. Plotting of the thermal neutron flux distribution curve.
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Figure 4. The arrangement of experiment setup.
5. Results of Experiment
The experimental results of the thermal neutron flux distribution are
obtained, form the activation measurement of the silver foil, and their theoretical
calculations are given in Table.1. Also in Figure 5 shows theoretical thermal
neutron flux distribution from the numerical solution of one-dimensional
Boltzmann transport equation using ANISN code, and Figure 6 shows the thermal
neutron flux distribution obtained both experimentally and theoretically.
Table.1 Results of the neutron flux distributions for distance between 2.75 - 5.75 form the
axis of the Pu-Be source.
Distance
from Source
Axis (cm)
Silver Foil
G.M counter ×
103 [n/cm .s]
Theoretical
estimation
× 104 [n/cm .s]
2.75 7.08 1.03
3.25 8.71 1.07
3.75 8.64 1.09
4.25 8.54 1.08
4.75 8.48 1.06
5.25 8.05 1.02
5.75 7.85 0.98
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For every measurement, the relative average error of silver foil was about
22 %. Table 2 gives the relative errors in percentage defined as the difference
between recorded neutron fluxes (experimentally) and estimated neutron flux
(theoretically) according to this equation:
(2)
Table.2 Relative errors in the determination of thermal neutron flux.
Distance
(cm)Relative Ag. Foil G.M counter
2.75 -31.0
3.25 -19.0
3.75 -21.0
4.25 -21.0
4.75 -20.0
5.25 -21.0
5.75 -20.8
Average error -21.8
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Figure 5. Theoretical thermal neutron flux distribution.
Figure 6. Thermal neutron flux distribution obtained both experimentally and
theoretically
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6. Discussion
Three main types of isotopic neutron sources can be distinguished namely
alpha-emitters, which produce neutron through an (α,n) reaction, Gamma-emitters
through a (γ,n) reaction and isotopes of heavy elements which undergo
spontaneous fission. The radioactive (α,n) source emit neutrons as a result of 9Be
(α,η) 12C nuclear reaction. This reaction produce neutrons with different energies
(non-mono-energetic neutrons) because of the following reasons: the alpha
particles have different initial energies and many of alpha particles lose part of
their initial energy by collision before they interact with a 9Be nucleus. The 12C
can be left in an excited state. The (α,n) reaction for 9Be is commercially induced
by the alpha-emitting radioisotopes 210Po, 239Pu, 226Ra and 241Am neutrons
produced by this reaction have a spectrum of energies ranging from 0 to about 11
MeV, the average neutron energy from these sources being about 4.5 MeV. Slow
neutrons can be obtained from this source by moderating fast neutrons in water,
paraffin or other hydrogen containing materials. Fast neutrons are slowed down to
thermal energy by elastic collisions, in the case of hydrogen the mean number of
collisions to thermals a fast neutron is 18 collisions. The alpha energies of 239Pu,
which decays with the half-life of 24300 years, are 5.10, 5.13, and 5.15 MeV.
Plutonium can be mixed with beryllium to produce Pu-Be source. The authors
applied this source in the present work. The source has the following
characteristics: high reproducibility, long lifetime, and they do not radiate gamma
rays. The absolute and relative measurements of neutron intensities can be
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precisely determined by activating a foil in the neutron field and counting its
radioactivity.
In the present work thermal neutron flux of the silver foil mined with mass
0.168 gm and thickness of 40 μm was determined. Silver exists in nature in two
isotopes of atomic masses 107 and 108.both isotopes undergo useful relation for
neutron activation analysis, the silver isotopes 108Ag is produced from 107Ag (η,γ)
108Ag nuclear reaction after a half-life (T1/2) of 2.42 min with cross-section 45 barn
and beta energy 1.77 MeV. The estimated thermal neutron flux distribution from
numerical solution of one-dimensional Boltzmann transport equation using
ANISN code, the ANISN code input program was prepared by Dydejczyk (1991)
[9]. The discrete coordinate simulation was performed for a cylindrical geometry
using an AMULTIGROUP cross-section library. The result of the simulation was
a thermal neutron flux distribution in the range from zero to 30 cm from the axis
of Pu-Be source, which is compared with the thermal neutron flux distribution
obtained experimentally by irradiation of silver foil. The shape of the spatial
distribution was similar even though obtained values slightly differ, because not
all necessary corrections were undertaken (correction coefficient between real
activity and recorded count rate). The relative errors in percentage defined as the
difference between recorded neutron flux (experimentally) and estimated neutron
flux (theoretically), for every measurement the relative average error for silver
foil was about 22%.
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7. Conclusion
The present work leads to the following conclusion
1. Good agreement was achieved between the special distribution of the
experimentally determined thermal neutron flux and theoretical estimated
thermal flux.
2. The difference between the numerical values of neutron fluxes is probably due
to the absolute measurement of foil activity.
3. The relative average error of silver foil measured was 22%
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8. References
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Verlag, Berlin, Gotottingen, Heidelberg, New York.
2. Dziunikowski B.(1989): Energy Dispersive X-ray Fluorescence Analysis,
Pwn-Polish Scientific Publishers, Warszawa.
3. Gardner R. P., and Ely R. L. (1987); Radioisotope Measurement
Application in Engineering. Reinhold Publishing Corp. New York,
Amsterdam, London.
4. Halliday D. (1985) Introductory Nuclear Physics. J. Wiley and sons Inc.
New York.
5. Foldiak G. (Ed.) (1986): Industrial Application of Radioisotopes.
Academe Kiado, Budapest.
6. Andesro P., Cunningham J.R., Hohlfeld, and Svensson H. (1985):
Absorbed dose determination in photon and electron beams. Technical
reports series No.277 IAEA , Vienna 1987.
7. Dziunikowski B. ,and Kalita S. (1989): Laboratory experiments in nuclear
techniques. Textbook No.1178 of the Academy of mining and metallurgy
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8. Hine G.J. and Brownell G.L. (1986): Radiation dosimetry, Academic
press Inc. publishers. New York.
9. Dydejczyk A. (1991): private communication.
10. Zealik, M. M., (1998): determination of Thermal and Epithermal Neutron
Flux from (Pu-Be) Source of In- foil, Fourth Arab conference on the
peaceful uses of Atomic Energy, Tunis, AAES.
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