Dose Rate and Shielding Scoping
Calculations for an IBN-12 Pu-Be Neutron
Source
Andrey Berlizov, Joseph Magill
JRC-ITU-TN-2007/78
The mission of ITU is to provide the scientific foundation for the protection of the European citizen against risks
associated with the handling and storage of highly radioactive material. ITU’s prime objectives are to serve as a
reference centre for basic actinide research, to contribute to an effective safety and safeguards system for the
nuclear fuel cycle, and to study technological and medical applications of radionuclides/actinides.
Report -No: JRC-ITU-TN-2007/78
Classification: Unclassified
Type of Report: Technical Report
Unit: Actinide Research / Nuclear Safeguards and Security
Action No: 53103 (FACIL)
Name
reviewed by the
project coordinator /
or action leader
(A. Berlizov)
approved by the
project leader
(J. Magill)
approved by the
head of unit
(K. Lützenkirchen)
released by the
director
(T. Fanghänel)
Date
Signature
European Commission
Joint Research Centre
Institute for Transuranium Elements
Contact information
Address: Dr. Andrey Berlizov, Institute for Transuranium Elements, Joint Research Centre,
Postafach 2340, 76125 Karlsruhe
E-mail: andriy.berlizov@ec.europa.eu
Tel.: +49 7247 951 586
Fax: +49 7247 961 99 586
http://itu.jrc.ec.europa.eu
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1
1. Introduction
The Institute for Transuranium Elements (ITU) was contacted in October 2007 by the Bundesamt für
Strahlenschutz (BfS) with a request to consider the chemical separation of the plutonium and beryllium in a IBN12 Pu-Be neutron source. These scoping calculations were performed to support decision making on the
possibility of performing such separation using facilities available in ITU. This report addresses primarily the
radiation safety and radiation protection issues, giving rough estimations for the dose rates, which would be
expected from the source, and shielding, which would be required for its safe handling.
2. Source properties
Pu-Be sources are fast neutron sources, where neutrons are produced mainly through the 9Be(,n)12C* reaction
with a relatively small contribution from the self-multiplication effect due to the neutron induced fission on Pu
isotopes and (n,2n) reactions on 9Be and other nuclides present in the source construction materials. An
experimental spectrum of the neutrons from a Pu-Be source is shown in Fig. 1 (Stewart, 1955). The intensity of
typical Pu-Be sources does not exceed 108 n/s. Alpha-beryllium neutron sources produce approximately 30
neutrons for every one million alpha particles
The fast neutrons from Pu-Be sources are always accompanied by less intense photon radiation originating
mainly from the decay of Pu isotopes and accumulated 241Am as well as from the de-excitation of 12C, yielding
4.44 MeV photons. The other less important sources of gamma-radiation include bremsstrahlung of secondary
electrons and various (n,x) activation reactions occurring in the source and its surrounding materials. A spectrum from a Pu-Be source collected using a conventional high-resolution gamma-spectrometer is shown in
Fig. 1 (Lakosi et al., 2005).
The neutron output of the Pu-Be sources and the amount of Pu inside them depends significantly on the isotopic
composition of Pu utilized for their production. Experimental studies show that Pu isotopics in Pu-Be source can
vary in broad ranges (Filß and Welse, 1993; Bagi et al., 2004).
The IBN type Pu-Be sources have been produced in the former Soviet Union and widely utilized in industrial and
military applications (acronym IBN stands for Source of Fast Neutrons). The construction of the sources and
their typical dimensions are shown in Fig.2 (data taken from the Booklet of the Radiation Sources Supplied by
the All-Union Enterprise "Isotope"). The sources include an active cylinder made of an inter-metallic compound
of plutonium and beryllium, which is tightly encased in a two-fold stainless steel capsule.
According to the data in Fig. 2, the IBN-12 type sources have external dimensions of 46 mm  46 mm and of
54 mm  64 mm for the active part and protective capsule respectively. Their typical strength is within (5 
1)107 n/s. The activity of Pu isotopes inside such sources does not exceed 1.31012 Bq and the -ray dose rate
at 1 m distance is less than 130 Sv/hr (the value has been converted from 9.510-10 ampere per kg indicated in
Fig. 2).
Fig. 3 shows the construction of containers (type KN-9m 107), which have been used for storing IBN-12 and
other neutron sources with similar strength. The major material used in the container shielding is paraffin
(estimated thickness is about 30 cm), which indicates that the major shielding effect for such sources is
achieved through the thermalisation of the fast neutrons.
Fig.4 shows a scanned copy of the Certificate of the IBN-12, which has been provided by the BfS. The values
indicated in the Certificate agree well with the typical characteristics of such sources: strength – 5.12107 n/s,
activity – 1.251012 Bq, external dimensions - 54 mm  64 mm. The date of the source production is August
1977. The neutron and photon dose field characteristics from the IBN-12 source, as measured and provided by
the BfS, are summarized in Table 1.
No specific data on the Pu content and isotopic composition in the IBN-12 source is available. According to the
information provided by the BfS (Beschreibung von geschlossen Pu-Be-Quellen schneller Neutronen) the typical
amount of 239Pu in the IBN-12 sources does not exceed 115.9 g. The isotopic composition averaged over ten
aged IBN type sources, as reported by Filß and Welse, 1993, was found to be: 0.6 wt.% - 238Pu, 75 wt.% - 239Pu,
24 wt.% - 240Pu, 0.5 wt.% - 241Pu (see Table 2).
3. Summary of the approach
Dose rates were evaluated basically by means of Monte Carlo calculations with application of the general
purpose Monte Carlo N-Particle Transport Code MCNP4c (Briesmeister, 1997). The calculations involve the
convolution of calculated neutron and photon flux energy distributions with the respective fluence-to-effective
dose conversion coefficients. The neutron and photon flux energy distributions were computed using analog (F2
tally) and/or non-analog (next-event-estimator F5 tally) approaches.
2
The anterior-posterior irradiation condition was assumed and the respective fluence-to-effective dose
conversion coefficients were utilized (Pelliccioni, 2000). The neutron doses were evaluated in tree different
spectral regions, from 0 to 0.55 eV, from 0.55 to 105 eV, and from 105 to 2106 eV, corresponding to thermal,
epithermal and fast neutrons respectively.
The photon doses were evaluated separately for 4.44 MeV photons, originating from the de-excitation of 12C,
and other photons, including bremsstrahlung of the secondary electrons and various (n,x) reactions in the
source and surrounding materials (capsule, shielding, air, measurement room etc.). The intensity of 4.44 MeV
photons was evaluated from the activity of the source, taking into account the experimental value 4.0610-5 (4.44
MeV photons per incident -particle) for the thick target -ray production yield (Heaton et al., 1997).
The MCNP model of the source included an active Pu-Be cylindrical part with dimensions 46 mm  46 mm
surrounded by 4 mm thick stainless steel walled capsule. A uniform distribution of the activity within the active
part of the source was assumed. The neutron emission rate 5.12107 n/s was taken according to the source's
Certificate. Energy distribution of the neutrons was modeled according to the experimental data shown in Fig.1.
The contribution of Pu isotopes and 241Am decay -rays to the total dose rate was evaluated deterministically
using the Dosimetry and Shielding module available in Nucleonica. The point-like model for the source was
considered. Additionally, the Nucleonica's Decay Engine was utilized in the evaluations of the current isotopic
composition of the Pu-Be source (Magill et al., 2007)
The dose and shielding calculations were performed for the following geometries: (A) source in free space, (B)
source positioned in the center of a measurement room (341.5 m3, assuming 30 cm thick concrete walls), (C)
source surrounded by a spherical shell (hollow radius - 50 cm, material layer thickness - 50 cm) made of water
and reinforced concrete. In all cases doses were evaluated at 1 m distance from the source. The geometry B
was considered to allow the comparison with the measured values reported by the BfS.
The following materials were involved in the calculations: water (H2O,  = 1.0 g/cm3), reinforced concrete (H 0.54 wt.%, O - 48.36 wt.%, Al - 4.43 wt.%, Si - 30.65 wt.%, Ca - 8.02 wt.%, Fe - 8.00 wt.%,  = 2.4 g/cm3), air (C
- 0.0126 wt.%, N - 76.5082 wt.%, O - 23.4792 wt.%,  = 1.29310-3 g/cm3), Pu-Be compound (238Pu - 0.235
wt.%, 239Pu - 35.23 wt.%, 240Pu - 11.28 wt.%, 241Pu - 0.235 wt.%, Be - 53 wt.%,  = 3.223 g/cm3). The density
and elemental composition of the Pu-Be compound were calculated from the source dimensions and anticipated
total Pu mass (154.5 g).
4. Results and discussion
To obtain an estimate for the current isotopic composition of the IBN-12 source, it was assumed that in 1993 it
was similar to the averaged isotopic composition found in the experimental study performed by Filß and Welse,
1993. Then, using Nucleonica, this composition was recalculated back to the fabrication date of the IBN-12
source (August 1977), and afterwards it was recalculated again to the present date. The summary of the
isotopic composition calculations is presented in Table 2 and in Fig. 5.
Partial and cumulative neutron and photon dose rates calculated for geometries A, B, and C are summarized in
Table 3. Spectral distributions of neutron and photon fluxes as simulated by MCNP are shown in Fig.6. One can
see that in the case of the bare source (geometry A) the calculated dose rates are underestimated in
comparison with the experimental values shown in Table 1 by a factor of 3 for neutrons and by a factor of 2.5 for
photons.
The inclusion of the measurement room into the simulation model (geometry B) improves significantly the
situation for neutrons, for which the resulting dose rate is almost doubled in comparison with the case of a bare
source, which is due to the efficient backscattering of the neutrons from the concrete walls (albedo). The
resulting discrepancy between the experimental and calculated neutron dose rates is about 1.8 times, which
seems to present rather a good agreement if one takes into account the approximations and simplifications
used in the calculations. At the same time the photon dose rate turns out to be almost uninfluenced by the
measurement conditions, so the discrepancy by about a factor of 2.5 is still observed between photon dose rate
in geometry B and the respective experimental value.
As a possible source of additional contribution to the measured photon dose rate, an activation of the source
construction elements (Fe, Ni, and Cr being the main components of the steel casing) by the fast neutrons has
been considered (Basenko et al, 2005). Considering a mass of 10 g for each component, the main activation
products were 54Mn (2.2x104 Bq), 51Cr (4.8x103 Bq), 55Fe (4.3x105 Bq), 57Co (3.7x105 Bq), and 58Co (2.4x105
Bq). Because of the relatively low activities, the activation products appear not to provide the explanation for the
measured photon dose rate.
As for the shielding calculations (geometry C), as one could expect, water provides the most efficient
thermalization of the source neutrons, which, in its turn, results in the higher reduction (by more than two
3
powers of magnitude) of the neutron dose rate for this type of shielding. The resulting neutron dose rate of 6
Sv/hr seems to be acceptable from the radiation protection point of view. At the same time, the water shielding
is not as good as the concrete one with respect to the attenuation of the photons. The photon dose rate is
reduced by the 50 cm water layer only by a factor of 3, while the dose attenuation in the case of the concrete
shielding with the same thickness is about a factor of 10. As one can see the major contribution to the photon
dose rate comes from the 4.44 MeV photons. This suggests that the addition of a high-Z high-density layer of
material behind the moderating medium is likely desirable to get better suppression of the photon contribution.
Tables 4-7 show details of the Nucleonica calculations of the Pu isotopes and 241Am decay radiation
contribution to the total gamma dose rate for various shielding materials and thicknesses. A total number of
2035 gamma and X-rays from a total of 42 parent and daughter nuclides have been taken into account in the
calculation. The dose rate varies from 689 µSv/h for no shielding (Table 4) to 15.5 µSv/h for shielding with 4 mm
Fe (to simulate the steel casing, Table 5). In both these cases, it can be seen that the main contributor to the
gamma dose rate is the 241Am line at 59.5 keV with a contribution of 529 µSv/h. The results are also shown for
50 cm water (Table 6) and 50 cm concrete (Table 7).
5. Conclusions
From the calculations performed the following two major conclusions can be drawn:

Reasonable agreement between the experimental and calculated neutron dose rate from the IBN-12 source
was obtained. The discrepancy between the measured and calculated photon dose rates was found to be
quite high, which perhaps requires further investigation. In this respect, the additional data on the isotopic
composition and mass of Pu inside the IBN-12 source could be of significant help. Experimental
measurement of the Pu isotopic composition in the source using the conventional NDA gammaspectrometry technique is therefore suggested.

The calculations performed demonstrate that the neutron and photon doses from the Pu-Be IBN-12 source
can be in principle reduced to the reasonable levels by using realistic amounts of traditional shielding
materials. The use of additional special shielding materials, e.g. borated high pressure polyethylene, can
probably further improve characteristics of the radiation shielding. Multilayer shielding is likely needed for
providing the efficient reduction of the resulting dose from the mixed neutron-photon radiation. Further
optimization of the shielding can be performed for the realistic configurations of the chemical processing
facility.
6. References
J.Bagi, T.C.Nguyen, L.Lakosi, Assessment of the Pu content of Pu-Be neutron sources. Nucl. Instrum. Methods
B 222 (2004) 242-248.
V.K.Basenko, A.N.Berlizov, I.A.Malyuk, V.V.Tryshyn, NAAPRO: a Code for Predicting Results and Performance
of Neutron Activation Analysis, Journ. Radioanal. Nucl. Chem., 263 (2005) 675-681.
J.F.Briesmeister. MCNP – a general Monte Carlo N-particle transport code. Los Alamos National Laboratory
Report, 1997, LA-12625-M.
P.Filß, H.-J-Welse, Bericht über die Messung von 10 Pu-Be Neutronenquellen am 14. und 15.0.1.1993 im BfS
Berlin, 1993, 5 Seiten.
R.K.Heaton, H.W.Lee, B.C.Robertson, E.B.Norman, K.T.Lesko, B.Sur, -particle induced -ray transitions in
light elements, Phys. Rev. C, 56(2) (1997) 922-937.
L.Lakosi, C.T.Nguyen, J.Bagi, Quantitative NDA of isotopic neutron sources, Appl. Radiat. Isot., 63 (2005) 681687.
J. Magill, J. Galy, R. Dreher, D. Hamilton, M. Tufan, C. Normand, A. Schwenk-Ferrero, H. W. Wiese,
NUCLEONICA: A Nuclear Science Portal, http://www.euronuclear.org/e-news/e-news-17/nucleonica.htm.
M.Pelliccioni, Overview of Fluence-to-Effective Dose and Fluence-to-Ambient Dose Equivalent Conversion
Coefficients for High Energy Radiation Calculated Using Fluka Code, Radiation Protection Dosimetry, 88(4)
(2000) 279–297.
L.Stewart, Neutron Spectrum and Absolute Yield of a Plutonium-Beryllium Source, Phys. Rev. 98(3) (1955)
740-743.
4
Table 1. Results of the dose rate measurements from the IBN-12 #027.
Radiation / Measurement conditions
Dose, mSv/hr
Neutron dose measurements:
At contact (about 0.3 m) to the source without shielding
1 m distance to the source without shielding
2 m distance to the source without shielding
30
1.8
0.4
Photon (gamma+X) dose measurements:
At contact (about 0.3 m) to the source without shielding
1 m distance to the source without shielding
2 m distance to the source without shielding
10
0.13
0.03
Table 2. Summary of the IBN-12 source isotopic composition calculations: Pu isotope composition measured in
January 1993 in the BfS. The corresponding masses are also given. The Nucleonica decay engine has been
used to calculate the original Pu composition 16 years earlier in 1977 and afterwards 30 years later in 2007.
Date
238
239
Pu
240
Pu
Pu
241
Pu
15 Jan 1993
0.6 (%)
75 (%)
24 (%)
0.5 (%)
15 Jan 1993
0.93 g
115.9 g
37.1 g
0.773 g
1977 composition
(7.8E12Bq)
 1.06 g
 115.9 g
 37.2 g
1.72 g
2007 composition
(3.0E12Bq)
 0.83
 115.9 g
 37.1 g
 0.39 g
241
Am
 1.29 g
Table 3. Results of the calculations of the partial and cumulative dose rates at 1 m distance from the Pu-Be
source.
Geometry
A
B
C (water)
C (concrete)
Neutron dose rate, Sv/hr:
Thermal (< 0.55 eV)
Epithermal (0.55 - 105 eV)
Fast (> 105 eV)
Total
0.0000036
0.58
626.26
626.84
9.00
27.77
987.47
1024.24
0.080
0.086
5.95
6.11
1.43
2.33
39.08
42.84
Photon dose rate, Sv/hr:
Pu + 241Am
9Be(,n)12C* 4.44 MeV
(n,x)-reactions and other
Total
15.5
36.49
1.30
53,29
15.5
38.05
4.56
58,11
3.7
12.66
2.92
19,28
0.0018
2.44
2.93
5,37
Cumulative dose rate, Sv/hr
680,1
1082,4
25,4
48,2
5
Table 4. Plutonium-Beryllium Source (Nov. 2007), no shielding. The main contributions to the photon dose rate
using Nucleonica (42 parent and daughters with 2035 gamma lines were included in the calculation).
Nuclide
Gamma Energy
(MeV)
Emiss. Probability P
(per disintegration)
Gamma Dose Rate
(μSv/h)
95 Am 241
0.0595
1.93E-2
529
95 Am 241
0.0263
1.30E-3
116
95 Am 241
0.0270
3.36E-4
28.3
95 Am 241
0.0332
6.53E-5
3.69
94 Pu 238
0.0435
6.85E-5
2.49
Total
689
Table 5. Plutonium-Beryllium Source (Nov. 2007), 4 mm Fe.
Nuclide
Gamma Energy
(MeV)
Emiss. Probability P
(per disintegration)
Gamma Dose Rate
(μSv/h)
95 Am 241
0.0595
1.93E-2
12.7
94 Pu 239
0.414
1.29E-6
0.228
94 Pu 239
0.375
1.36E-6
0.225
95 Am 241
0.103
1.12E-5
0.216
94 Pu 238
0.0999
1.28E-5
0.188
Total
15.5
Table 6. Plutonium-Beryllium Source (Nov. 2007), 50 cm water.
Nuclide
Gamma Energy
(MeV)
Emiss. Probability P
(per disintegration)
Gamma Dose Rate
(μSv/h)
95 Am 241
0.0595
1.93E-2
3.35
94 Pu 239
0.375
1.36E-6
2.92E-2
94 Pu 239
0.414
1.29E-6
2.92E-2
95 Am 241
0.103
1.12E-5
2.78E-2
94 Pu 238
0.0999
1.28E-5
2.49E-2
Total
3.7
Table 7. Plutonium-Beryllium Source (Nov. 2007), 50 cm concrete.
Nuclide
Gamma Energy
(MeV)
Emiss. Probability P
(per disintegration)
Gamma Dose Rate
(μSv/h)
94 Pu 239
0.414
1.29E-6
3.04E-4
94 Pu 239
0.375
1.36E-6
2.34E-4
95 Am 241
0.662
1.93E-7
1.87E-4
95 Am 241
0.723
5.21E-8
7.88E-5
Total
1.83E-3
6
Fig. 1. Experimental energy distributions of the neutrons (top diagram) and gamma-rays
(bottom diagram, black line) from a Pu-Be source.
7
Fig. 2. Drawing and typical characteristics of IBN-1  IBN-12 type Pu-Be neutron sources (in Russian).
8
Fig. 3. Drawings of the typical containers used for storing Pu-Be neutron sources (in Russian). The container for
the IBN-12 source is shown on the right bottom corner.
9
Fig. 4. Certificate of the Pu-Be fast neutron source IBN-12 #027 (in Russian)
10
Fig.5. Emission rate (emission probability per decay) of photons from the Pu-Be source (Nov. 2007). The
original source (dated 1977) contained only plutonium isotopes and beryllium. As a result of ageing, the source
today (Nov. 2007) contains Am-241 and other daughters of the plutonium isotopes. The calculations were made
with Nucleonica.
11
-4
10
Neutrons
Geometry A
Geometry C (water)
-5
Geometry B
Geometry C (concrete)
Fluence per source neutron, cm
-2
10
-6
10
-7
10
-8
10
-9
10
-10
10
-11
10
-9
10
-8
10
-7
10
-6
10
-5
-4
10
-3
10
10
E, MeV
10
-2
-1
10
0
10
1
10
2
10
-5
10
-6
Fluence per source neutron, cm
-2
10
-7
10
-8
10
-9
10
(n,x) photons
Geometry A
Geometry B
Geometry C (water)
Geometry C (concrete)
-10
10
-11
10
-12
10
0
1
2
3
4
5
6
7
8
9
10
11
12
E, MeV
-5
10
Fluence per source photon, cm
-2
9
-6
12
Be(,n) C* photons
Geometry A
Geometry B
Geometry C (water)
Geometry C (concrete)
10
-7
10
-8
10
0
1
2
3
4
5
E, MeV
Fig.6. Neutron and photon flux energy distributions at 1 m distance from the Pu-Be source.
12
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13
European Commission – Joint Research Centre – Institute for Transuranium Elements
Title: Dose Rate and Shielding Scoping Calculations for an IBN-12 Pu-Be Neutron Source
Author(s): Andrey Berlizov, Joseph Magill
2007 – 13 pp. – 21.0 x 29.7 cm
Abstract
The Institute for Transuranium Elements (ITU) was contacted in October 2007 by the Bundesamt für
Strahlenschutz (BfS) with a request to consider the chemical separation of the plutonium and beryllium in an
IBN-12 Pu-Be neutron source. These scoping calculations were performed to support decision making on the
possibility of performing such separation using facilities available in ITU. This report addresses primarily the
radiation safety and radiation protection issues, giving rough estimations for the dose rates, which would be
expected from the source, and shielding, which would be required for its safe handling.
14
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15