Measurement of Neutron Capture Cross Sections of139
La, 152Sm and 191,193Ir at 55 and 144keV
Vuong Huu Tan
Vietnam Atomic Energy Commission
59-Ly Thuong Kiet, Hanoi, Vietnam

Tran Tuan Anh, Nguyen Canh Hai and Pham Ngoc Son
Dalat Nuclear Research Institute01-Nguyen Tu Luc, Dalat, Vietnam

Tokio Fukahori
Nuclear Data Center, Nuclear Science and Engineering Directorate
Japan Atomic Energy Agency
Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, JapanThe neutron capture cross sections of 139La, 152Sm and 191,193Ir at average energies of55keV and 144keV have been measured relative to the standard capture cross sections of197
Au by means of the activation method. The neutron beams were derived by filteredtechniques from the horizontal channel No.4 of the research reactor at the Dalat Nuclear
Research Institute, Vietnam. A fast-digital gamma-ray spectroscopy in compacted with a 58%
efficient HPGe detector has been used for measurements of gamma-ray spectra from theactivated samples. The correction factors for multi-scattering, self-shielding and resonancecapture effects of neutron in the irradiating samples were calculated by Monte-Carlotechnique. The present results have been compared with the previous measurements, and theevaluated data from JENDL-3.3 and ENDF/B-VI.8 libraries.
Key words: neutron, capture cross section, nuclear reaction data, filtered neutron beam,
resonance neutron capture, correction

1. Introduction
Accurate measurements of neutron capture cross sections for most of nuclides are currently
necessary for the calculations of neutron transport, the assessments of the reactor safety, theinvestigations of high-burn-up core characteristics, the decay heat power predictions, and for thenuclear transmutation study. In keV energy region, the (n,γ) cross sections of the nuclides at or nearmagic neutron number, N=50, 82 and 126 are special important for the study on the s-process reaction
chain for nucleosynthesis. However, the present status of experimental data for capture cross sections
is still inadequate both in quality and in quantity. Therefore, it is important to perform the precisely
measurements of capture cross sections for those nuclides, particular in keV energy region [1-6].
In the present experiment, we performed the measurements of capture cross section of 139La, 152Sm
and 191, 193Ir on the filtered neutron beams of 55keV and 144keV, relative to the standard capture crosssection of 197Au by the activation method. The neutron beams were derived from the horizontal
channel No.4 of the research reactor at the Dalat Nuclear Research Institute (DNRI), by using thefiltered compositions of 98cmSi + 35g/cm2S + 0.2g/cm2B10 and 98cmSi + 1cmTi + 0.2g/cm2B10 for
55keV and 144keV respectively [6,7]. The neutron energy resolution, FWHM, is 8keV at 55keV peak,and 22keV at 144keV peak [7].
Beside determining the corrections for neutron multi-scattering and self-shielding in irradiatedsamples, it is important to concern that the large resonance capture cross sections of the standards andsamples, in the slow neutron background region above the Cd-Cutoff energy, strongly contribute to the2006 Symposium on Nuclear Data

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uncertainty of experimental results. Therefore, the correction factors for slow neutron resonancecapture in the present experiments were also calculated by the general least square method for strong
resonance captures at 4.92eV of 197Au, 8.06eV of 152Sm, 1.3eV of 193Ir, 0.67eV, 5,38eV and 6.15eV of191
Ir, and 72.3eV of 139La. Furthermore, a low background and fast-digital spectroscopy with a highefficiency, 58%, HPGe detector has been used for detection of gamma-ray spectrum from theirradiated samples, and the statistical uncertainties are expected to be less than 1%.
2. Experiments
The measurements for neutron capture cross sections of 139La, 152Sm and 191, 193Ir at the energies of55keV and 144keV were performed on the filtered neutron beams of DNRI. The neutron beams werecollimated to 3cm in diameter by using the usual materials of LiF, Cd, B4C, Pb and borated paraffin.
The physical properties of these beams are given in Table 1. Since the neutron spectra had been
experimentally measured before with a recoil-proton counter [6], in this work, the neutron transport
and unfolding methods [8,9] were applied to obtain the exact spectra, shown in Fig. 1 and 2, whichhave been used for calculation of the average quantities and of the correction factors.
The samples were prepared from the natural oxide powders, 99.99% purity, of La2O3, Sm2O3 andIrO2. In order to diminish the effect of water on the samples [10], each collected amount of thepowders was dried up at about 105oC for several hours before weighting and pressing into pellets.
Then the pellets were covered by thin polyethylene foils. The standard gold foils with 2.54cm indiameter and 0.01mm in thickness were used as the neutron flux monitors. Each sample wassandwiched between two gold disks, and the sample groups were wrapped in Cd covers with 0.5mm inthickness with aim to reject most of thermal neutron background. The irradiation time was 70 hours
for every sample group.
The specific activities of the samples and the gold disks were measured by using the fast-digitalspectroscopy system and with the high efficiency HPGe detector. The system was calibrated by using
standard radioisotope sources and a multi-nuclides standard solution, supported by IAEA. Each ofirradiated samples, standards and calibrated sources was measured under the same conditions.
Table 1 The properties of the filtered neutron beams [6]Neutron energy(keV)
55
144

Filter combination
98cmSi + 35g/cm2S + 0.2g/cm2B102

1098cmSi +1cmTi + 0.2g/cm B

400

160

350

140

Relative intensity

Relative intensity

180

Transpo rt calculatio n300
250
200
150

Neutron energy (eV)Fig. 1 Neutron spectrum of the 55keV
filtered beam

6.0E+04

Transport calculation

60

20

5.5E+04

Unfolding spectrum

8040

2006 Symposium on Nuclear Data

22 keV100

50

5.0E+04

8 keV

6120

100

4.5E+04

5.61 x 105

200

Unfo lding spectrum

04.0E+04

FWHM

2.14 x 10500450Flux density(n/cm2/s)

08.0E+04

1.2E+05

1.6E+05

2.0E+05

Neutron energy (eV)Fig. 2 Neutron spectrum of the 144keV
filtered beam

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3. Data Analysis
During irradiation in the neutron beam with energy spectrum Φ(E), the reaction rate, R, of samples
is defined as follows:

R = N ∫ Φ(E)σ a(E)dE ,(1)where N is the number of nuclei in sample, and σa(E) is the neutron capture cross section at energy E.
The average neutron capture cross section, <σa>, and neutron flux, <Φ>, are defined as following:< σ a >= ∫ σ a ( E ) Φ ( E ) dE / ∫ Φ ( E ) dE ; < Φ >=∫ Φ ( E ) dE .

Applying these average quantities, the integrating equation (1) can be rewritten:

R = N < σ a >< Φ >(2)
The activity, A, of the sample at the end of neutron irradiation is given by expressions:

A = R (1 − exp( − λ t1 )) ,A=(3)Cf c λ,ε γ I γ exp(−λt 2 )(1 − exp(−λt 3 ))(4)

where C denotes the net counts of the corresponding gamma peak, and t1, t2 and t3 are irradiating,cooling and measuring times, respectively. λ is the decay constant of the product nucleus, εγ thedetection efficiency of detector, Iγ the intensity of interesting γ-ray line and fc is the correction factors.
Finally, from equations (2), (3) and (4), the average capture cross sections, <σa>x, for the samples ataverage neutron spectrum <Φ> can be obtained relative to that of 197Au standard by the following
relations:< σ a >=f (λ , t ) =C x f (λ , t ) x f cx I γAu ε γAu N Au < σ a > Au
C Au f (λ , t ) Au f cAu I γxε γx N xλ(1 − exp(−λt1 )) exp(−λt 2 )(1 − exp(−λt 3 ))

,(5)

(6)

where the superscript ‘x’ denotes the nucleus of sample. Calculating from the ENDF/B-VI.8 datalibrary, the average standard capture cross section of 197Au respects to the 55keV beam’s spectrum is414.61mb, and for that to the 144keV beam’s spectrum is 277.21mb. The relevant decay data ofproduct nuclei, used in this work, are given in Table 2.
Productnucleus
198
Au
140
La
153
Sm192
Ir194
Ir

Table 2 Decay properties of the product nuclei [11]Intensity per decay
Half-life
γ-ray energy
(%)
(keV)
2.6952±0.0002 d
411.8
95.6±0.1
1.6781±0.0003 d
487.0245.5±0.6
46.50±0.21h103.2
29.3±0.1
73.827±0.013 d
316.5
82.7±0.2
19.28±0.13
h328.4513.1±1.7

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The correction factors for the neutron self-shielding, multi-scattering and the effects of strong
resonance capture of samples with slow neutron background were calculated by Monte-Carlo method[9, 12]. In which, the effect of isotopic impurities and oxygen in the samples and the loss energy byscattering of neutron were taken into account. In the resonance capture corrections, the background
spectra of neutron beams were measured by unfolding method and resonance thin-foil activation
technique. The data used for the correction calculation were taken from JENDL-3.3 [13] and ENDF/BVI.8 [14]. The calculated correction factors are given in Table 3.
Table 3 Correction factors for multi-scattering, self-shielding and background resonance capture ofneutron in the samples
55keV region
144keV region
Nuclides
SelfMultiResonance
SelfMultiResonanceshieldingscattering
captureshieldingscattering
captureAu-1970.99850.99010.42690.99880.99290.5338
La-1390.99620.97850.62270.99860.982
0.7531
Sm-1520.99880.9856
0.28160.9991
0.99170.4890Ir-1910.99590.9782
0.4937
0.99680.98280.6593
Ir-1930.99590.9774
0.5214
0.99680.98260.6944
4. Results and Discussion
In the present work, the new values of average neutron capture cross sections of 139La, 152Sm and191, 193
Ir at incident neutron energies of 55keV and 144keV are reported with errors about 5-6.5%. The
results are given in Table 4. The uncertainties in the present measurements are mainly due to thestatistical errors (0.1-2%), the uncertainties of γ-ray detection efficiency (3.5%), the reference crosssection (∼3%) and the correction factors for neutron resonance capture, self-shielding and multiscattering effects (∼3%). In comparisons with the previous measurements and the evaluated data, The
present results are seem to be good agreement with the previous measurements of Musgrove [15],Wisshak [16], Duamet [17], Macklin [18] and with the evaluated data of JENDL-3.3 and ENDF/BVI.8 within the experimental uncertainties. The comparisons results are shown in Figs.3-6.

0.6

0.08

Cross section (barn)

0.07
0.06
0.05
0.04
0.03
0.02

K.WISSHA K
K.GUB ERB . DUA M ET
LUO XIA O-B ING

0.5
Cross section (barn)

A .E.JOHNSRUD
A .R.DEL.M USGROVED.C.STUP EGIA
A .R.DEL.M USGROVE
J.H.GIB B ONS
G.G.ZA IKIN
R.P .A NA ND
This wo rk
ENDF/B -6.8
JENDL 3.3

This w ork
ENDF/B -6.8
JENDL 3.3

0.4

0.3

0.2
0.01

139

140

0.00
2.0E+048.0E+04

La(n,γ )152

La
1.4E+05

2.0E+05

Neutron energy (eV)Fig. 3 Neutron capture cross section of139
La in keV region

2006 Symposium on Nuclear Data

0.1
2.0E+04

Sm(n, γ )153Sm
8.0E+04

1.4E+05

2.0E+05

Neutron energy (eV)Fig. 4 Neutron capture cross section of152
Sm in keV region

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1.2

1.8

1.1

S.JA A G

1.0

R.L.M A CKLIN

1.4

Cross section (barn)

Cross section (barn)

1.6

M .LINDNER

This wo rk
ENDF/B -6.8

1.2
1.0
0.8

R.L.M A CKLIN
M .LINDNER
S.JA A G
A .K.CHA UB EY
A .G.DOVB ENKO
This wo rk
ENDF/B -6.8

0.9
0.8
0.7
0.6
0.5

0.6
191

0.4
2.0E+04

Ir(n,γ)

192

0.4

Ir

8.0E+04

1.4E+05

2.0E+050.3
2.0E+04

Neutron energy (eV)Fig. 5 Neutron capture cross section of191
Ir in keV region

193

Ir(n,γ )194Ir
8.0E+04

1.4E+05

2.0E+05

Neutron energy (eV)Fig. 6 Neutron capture cross section of
193
Ir in keV region

Table 4 The neutron capture cross sections of 139La, 152Sm and 191, 193Ir obtained in the present study
Average neutron energy[Energy range] (keV)<σa>La-139(mb)<σa>Sm-152(mb)<σa>Ir-191(mb)<σa>Ir-193(mb)

55 [51-59]

22.4 ± 1.2

345.5 ± 19.4

1016.5 ± 57.2

566.7 ± 32.6

144 [133-155]

12.01 ± 0.58

258.7 ± 14.5

514 ± 29.4

404.5 ± 22.8

5. Conclusion
The neutron capture cross section of 139La, 152Sm and 191, 193Ir at average incident neutron energiesof 55keV and 144keV have been measured by means of the activation method, using the filteredneutron beams at DNRI. The results, with uncertainties of 5-6.5%, were obtained relative to thestandard capture cross sections of 197Au. The energy ranges of the filtered neutron beams are 14 and15% for 55keV and 144keV beams, respectively. Although the slow neutron background existing in
the filtered neutron beam is quite low, the effects of strong resonance capture cross sections of thesamples and standards should be taken into account to improve the accuracy of the experimental
results.
Acknowledgments
We would like to express our thanks to the Vietnam Atomic Energy Commission (VAEC) and theDalat Nuclear Research Institute (DNRI) for their great encouragement and helpful in creating anadvantage research condition for this work. The present study was sponsored by the Fundamental
Research Program of VAEC under the subproject No. 409706. The advanced gamma-ray spectroscopy
system used in the experiments was supported by IAEA. We also wish to express our deep gratitude tothe members of the Nuclear Data Center of JAEA for their helpful reference data and materials.
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