Progress in Nuclear Energy 93 (2016) 76e83
Contents lists available at ScienceDirect
Progress in Nuclear Energy
journal homepage: www.elsevier.com/locate/pnucene
Effect of heat treatment on neutron attenuation characteristics of high
density concretes (HDC)
Sivathanu Pillai C., Santhakumar A.R., Chandrasekaran S., Viswanathan S., Mathiyarasu R.,
Kumar J. Ashok, Preetha R.*, Venkatraman B.
Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102, India
a r t i c l e i n f o
a b s t r a c t
Article history:
Received 13 August 2015
Received in revised form
24 June 2016
Accepted 8 August 2016
In nuclear reactor and fuel cycle facilities, apart from ordinary concrete (2300 kg/m3) various special
types of concrete of density varying in range 3600e4600 kg/m3have been effectively employed to
achieve less shielding thickness. Radiation attenuation in concrete mainly depend on types of aggregate
used; water to cement ratio, elemental composition and moisture content apart from density of concrete.
In addition to this, operating temperature also plays a vital role in deciding its shielding properties due to
loss of water content and microcracking. The effect of heat on neutron attenuation properties of HDC
upon heat exposure is the scope of the study in this paper. Four types of concrete made of granite and
having hematite and steel shots as aggregate were prepared with density varying from 2300 to 4300 kg/
m3and exposed to 120 C for durations (14, 28 and 56 days). Neutron Attenuation Factor (NAF) was
obtained from experimental study and compared with results obtained with neutron transport calculations. The results of this study clearly indicate that the sustained and cyclic heat treatment at 120 C up
to 56 days reduced the neutron attenuation factor by a factor of two.
© 2016 Elsevier Ltd. All rights reserved.
Keywords:
High density concrete
Neutron attenuation
Heat treatment
Aggregates
1. Introduction
Concrete is one of the most widely used materials to shield
Gamma and Neutrons in nuclear installations. Depending on the
shielding requirement, various types of concrete with different
densities and different combinations are being used. Various reasons for using concrete are ease of fabrication, low cost for the
construction and maintenance (Kaplan, 1989). In order to provide
the highest attenuation of gamma and neutron radiation as well as
optimum shield layout design, a delicate balance must be achieved
between the proportion of high density aggregate and the ingredients, which contain hydrogen in a form of chemically bound
or adsorbed water.
In a nuclear reactor, the neutron flux is generally of the order of
1016 to 1018 n/cm2/s (Samarin, 2013). For efficient neutron shielding, concrete must contain some heavy elements, which are capable
to slow down fast neutrons, and a sufficient quantity of hydrogen to
slow down the intermediate and to absorb the slow neutrons. Ideal
shield materials which absorb both fast and slow neutrons at
* Corresponding author.
E-mail address: predinesh@igcar.gov.in (P. R.).
http://dx.doi.org/10.1016/j.pnucene.2016.08.003
0149-1970/© 2016 Elsevier Ltd. All rights reserved.
equally high rates do not exist. A shield material like concrete,
which brings down the energy of neutrons by elastic and inelastic
scattering along with absorption will be more effective.
Shield concrete undergoes heat exposures during the service life
either due to high operating temperature of the systems or due to
radiation. High Temperature Engineering Test Reactor (HTTR) is the
first High Temperature Gas-cooled Reactor in Japan, uses primary
upper shield, composed of concrete and carbon steel, to attenuate
neutrons and gamma rays generated in the core to satisfy dose rate
criterion for the operating floor (Sumita et al., 2000). The maximum
temperature inside of the primary upper shield during full-power
operation was estimated to be about 85 C. A Prototype Fast
Breeder Reactor (PFBR), sodium cooled, pool type of reactor, is
being built in Kalpakkam, India, uses high density concrete as part
of vault and roof shielding of main reactor vessel (Velusamy et al.,
2010). Roof slab made of steel, forms the top cover for main
vessel, which is filled with concrete for the purpose of nuclear radiation shielding. During normal operation, hot pool within the
reactor is at 550 C while roof slab is maintained at 120 C and
reactor vault is kept below 65 C.
Exposure of concrete to high temperature induces complex
changes in the moisture content as well as chemical composition of
the cement paste. Moreover, there exists a mismatch in the thermal
S.P. C. et al. / Progress in Nuclear Energy 93 (2016) 76e83
expansion between the cement paste and the aggregate. Therefore,
factors such as changes in chemical composition of concrete and
the extent of mismatch in thermal expansion lead to internal
stresses and microcracking in the concrete constituents (aggregate
and cement paste) (Naus, 2009). The increase in concrete temperature leads to a sequence of events e dehydration of cement paste
above 100 C, crack formation in the range of 400e1000 C,
decomposition of hydrated cement products like calcium hydroxide [Ca(OH)2] at 450 C, mineralogical changes in aggregates above.
500 C, disintegration of calcium silicate hydrates [CSHs] above
600 C and formation of glassy powder around 1200 C (Naus,
2009; Mohammed Haneefa et al., 2013; Lee et al., 2009; Gencel,
2012; Koksal et al., 2012).
Since shield concrete is exposed to varying temperature regime,
it is required to study the effect of temperature on radiation
attenuation characteristics of such concrete.
Hematite aggregate, which is locally available and steel shots
made of high carbon steel are utilized in concrete to achieve the
required density for gamma attenuation. A few have studied
(Gencel et al., 2010), effects of different concentrations of hematite
on physical and mechanical properties of concrete and have
concluded that the properties improved with hematite aggregate.
In a study (Gencel et al., 2011), five different concrete samples were
prepared with varying hematite proportion and evaluated for
neutron shielding. And it was found that there has been no effect
for hematite addition on neutron shielding properties and it was
concluded that the neutron shielding was dependent only on water
content.
Few studies were reported with different aggregates (dolomite,
serpentine) including hematite the samples (Kharita et al., 2008)
were subjected to temperatures from 20 to 800 C (considering fire
hazard) for 240 h and tested for mechanical and shielding properties. It was concluded that samples with hematite were the best
for shielding of gamma radiation whereas samples with serpentine
were the worst. For shielding of neutron radiation also hematite
samples were the best. This behavior may be the result of the high
content of iron in the samples and the presence of iron hydroxides.
The concrete composition that contains serpentine (which contains
considerable amount of hydrogen) comes as second best for
neutron shielding.
Divya et al (Divya Rani et al., 2013)., reported that ordinary
density concrete when subjected to sustained temperatures of 65,
75 and 90 C was not showing any change in density beyond 50
days.
There is paucity of information on shielding properties under
sustained cyclic temperatures, just above 100 C (boiling point of
water) as encountered in reactor environment. Thus, the effect of
elevated, sustained and cyclic temperatures on neutron attenuation
characteristics of concrete samples with granite aggregate, hematite aggregate and steel shots, having densities 2300 kg/m3to
4200 kg/m3has been studied in this work. The details of measurements carried out to study the neutron attenuation characteristics and the salient results obtained are discussed in this paper
along with theoretical validation.
2. Experimental
2.1. Materials
Hematite, a natural red rock that contains iron oxide, when pure
has the Mohs hardness between 5.5 and 6.5 and the specific gravity
between 4100 kg/m3and 4800 kg/m3. Hematite was prepared as
aggregate by crushing and grounding the ore in a laboratory mill,
then sorting it via sieves into two groups of coarse and fine aggregates. Four types of concrete mixtures were prepared for this
77
study using the Portland conforming to BIS IS 8112-2013 (IS 8112
-2013,Indian Standard 43 grade Portland Cement-Specification,
2013). Details of mix design used for concrete are presented in
Table 1. For a given water-cement ratio, for radiation shielding
concrete, cement content is generally quite high, greater than
350 kg/m3which helps to improve the neutron shielding characteristics of the concrete because of the high bound water content of
the paste (Gencel et al., 2010). Also, the increase in the cement
content will also increase the water content per unit volume of
concrete increasing the workability of concrete. Apart from
increasing cement content, admixture in the form of superplasticizers (naphthalene based) is added to achieve a cohesive
workable concrete mix which can be easily placed within reinforced concrete vault.
- Sample A - siliceous sand and crushed granite were used as fine
aggregate for granite aggregate concrete of density 2300 kg/m3
- Sample B -hematite aggregates were used for designing high
density concrete of 3600 kg/m3
- Sample C and D - In addition to the hematite aggregate, high
chrome high carbon steel shots of diameter 4 mm and 2 mm
were used for designing high density concrete of 3900 kg/
m3and 4200 kg/m3respectively.
2.2. Sample preparation
Absolute volume method was used to obtain denser concrete in
mix design of concrete samples. Three samples each,
0.15 0.15 0.15 m3 were prepared of each of the above
mentioned mixtures. The specimen size is arrived based on
capturing the bulk attenuation characteristics rather than the linear
attenuation as followed in the literature. The cubes were cast in
iron moulds and de-molded after 24 h. The cubes were then
conserved in water bath (22 C) for 28 days, then taken out and left
to dry for few days. The measured density of the samples is given in
Table 2.
2.3. Test procedures
2.3.1. Heat treatment
The samples were subjected to a heat treatment, at a constant
temperature of 120 C, in a thermal cyclic chamber (Fig. 1), size of
1.5 1.5 0.6 m3 with an operating temperature range of ambient
to þ250 C and rate of heating of 4 C/min. The samples were kept
in the chamber for a period of 14, 28 and 56 days (durations at
which usually concrete properties are tested). The time cycle of the
concrete samples starting from it curing, temperature exposure and
testing is shown in Fig. 2. After each stage of the heat treatment, the
samples were weighed and studied for its neutron attenuation
characteristics.
2.3.2. Neutron attenuation test
Neutron attenuation testing is carried out before and after each
campaign of heat treatment. Three samples for each density were
subjected to testing. Directly from thermal chambers the specimens
were weighed, before each radiation test and then placed in a
closed container such that there is no moisture absorption from the
atmosphere until radiations tests are conducted.
A 5 Ci 241Am-Be neutron source which has neutron emission
rate of 1.2 107 n/s was employed for neutron measurements. The
schematic diagram of experimental set up used is shown in Fig. 3.
The source was kept in paraffin cylinder and the concrete samples
were placed close the source position as indicated in Fig. 3. Neutron
dose rate was measured using He-3 based neutron detector. The
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S.P. C. et al. / Progress in Nuclear Energy 93 (2016) 76e83
Table 1
Details of aggregates used ordinary and Hematite concrete.
Sample ID
A
B
C
D
Cement kg/m3
Water kg/m3
420
420
420
420
155.4
185.0
176.4
176.4
Fine aggregate kg/m3
Coarse aggregate kg/
m3
20 mm
12 mm
580
884
740
555
580
884
740
555
772
1274
1105
1075
Steel shots kg/m3
4 mm
2 mm
e
e
416
831.63
e
e
416
675
Admixture kg/m3
2.94
5.04
5.04
5.04
Table 2
Measured density values of concrete samples.
S. no
Sample
Concrete make
Measured density (kg/m3)
1
2
3
4
A
B
C
D
Normal aggregate with river Sand
Hematite aggregates
Hematite aggregate with steel shots
Hematite aggregate with steel shots
2500
3650
4020
4290
±
±
±
±
10
12
14
11
distance between source and the detector was kept constant at
0.42 m for all measurements. The entire experimental setup was
situated at a height of 2 m above the ground in a large room of size
8 8 10 m3 in order to reduce the neutron ground reflection from
the room as much as possible.
Neutron dose rates were measured with and without concrete
samples. The difference between these two values was considered
for neutron attenuation. Multiple measurements were made and
the average values were taken. The experimental investigation of
neutron attenuation and neutron transport simulation study is
similar to the studies described in few research papers (Korkut
et al., 2012, 2013).
Fig. 1. Thermal cyclic chamber.
Fig. 2. Time cycle of curing, heating and testing of samples.
S.P. C. et al. / Progress in Nuclear Energy 93 (2016) 76e83
79
Fig. 3. Schematic diagram of experimental setup.
3. Results and discussions
3.1. Estimation of moisture loss in concrete
To estimate the rate of weight loss, the samples were withdrawn
from the thermal cyclic chamber at 1, 2,3,7,14,28 and 56 days. After
weighing, the samples were carefully placed in an air tight box to
prevent the re-entry of moisture in the specimens. Based on the
water cement ratio (w/c) (Table 1), initial theoretical water content
in each specimen and moisture loss was determined in terms of
weight loss and plotted as Fig. 5.
Water, which is the main source of hydrogen in concrete, can be
present in cement paste in a free, adsorbed and chemically combined state (Fig. 4 a and b). Physical adsorption involves only van
der Waals forces, and in a chemical adsorption, transfer of electrons
also takes place (Naus, 2009). Free water, which is not chemically
combined in the process of cement hydration, will be ultimately
lost, through the process of diffusion. The amount of chemically
bound water is influenced by the mineral composition of Portland
cement, age of cement paste, water-cement ratio of concrete and
the environment (temperature, humidity) to which this concrete is
exposed. Fig should be changed.
At temperatures up to 80 C the hydration products of ordinary
Portland cement essentially remain chemically unaltered and
changes in the properties can be attributed to physical effects (e.g.,
changes in van der Waals cohesive forces, porosity, and cracking)
(Naus, 2009). As the temperature to which the cement paste is
subjected increases, evaporable water is driven off until at a temperature of about 105 C all evaporable water will be lost, given a
sufficient exposure period. At temperatures above 105 C, the
strongly absorbed and chemically combined water (i.e., water of
hydration) are gradually lost from the cement paste hydrates. This
represents the dominant process affecting performance as temperatures increase above about 100 C. Dehydration of the hydration products is essentially zero up to about 400 C, increases most
rapidly around 535 C, and becomes complete at about 600 C and
Fig. 4. (a) Schematic description of concrete structure, and (b) Volumetric proportions of concrete components (Ichikawa and England, 2004).
80
S.P. C. et al. / Progress in Nuclear Energy 93 (2016) 76e83
Fig. 5. Moisture loss after heat exposure at 120 C.
thus impacts concrete performance at higher temperatures (Naus,
2009).
In the present investigation, loss of moisture content was high in
the first 7 days in all the samples (Fig. 5). This may be attributed to
the higher temperature (120 C) regime, which is beyond the
boiling point of water. Since the reduction in moisture content is
minimal after 14 days the free water would have evaporated
completely. Also, it is observed that the moisture loss in ordinary
density concrete (A) was relatively more than the heavy density
concrete.
3.2. Neutron attenuation factor (NAF)
In order to study the effect of heat treatment on the neutron
attenuation characteristics of concrete, NAF was estimated using
the relation,
NAF ¼
D0 Dt
D0
(1)
where D0 is the dose rate observed in the detector without concrete
block, Dt is the dose rate measured with concrete block. The variation in NAF with density, heat treatment and moisture content in
each concrete samples is discussed below.
3.2.1. Effect of concrete density on NAF
Fig. 6 shows the variation of NAF with density of concrete
samples and it is observed that the NAF increased with density
from 0.799 to 0.834. The same trend is observed after the heat
treatment. Even though, all samples were having the similar water
content, the increase in NAF with density would be attributed to
elements present in the HDC such as Fe, Si etc. The neutron spectrum from Am-Be source has thermal component of 1.44%,
0.1e1.0 MeV- 13.92%, 1-2 MeV-10.03% and 2e10 MeV -74.6%. Fast
neutron component present in Am-Be source spectrum got attenuation by inelastic scattering with elements such as Fe.
3.2.2. Effect of heat treatment on NAF
The variation in NAF with heat treatment of concrete samples is
plotted in Fig. 7. The heat treatment of 14 days and 28 days at 120 C
have shown measurable reductions in NAF value i.e., about 24.78%
and the subsequent heat treatment of 56 days did not reduce the
NAF value significantly (1.56%) for ordinary concrete. The reduction
percent of NAF for HDC sample with density of 3600 kg/m3for 14,
28 and 56 days of heat treatment was about 25.76%, 21.17% and
0.83% respectively, while that of HDC sample with density 3900 kg/
S.P. C. et al. / Progress in Nuclear Energy 93 (2016) 76e83
81
Fig. 6. Variation of NAF with concrete density.
m3 was about 24%, 17.62% and 1.73% at the same exposure time. It
was 20.86%, 16.97% and 1.09% reduction at the same heat treatment
for concrete with density of 4200 kg/m3. The reduction of NAF can
be attributed to the loss of free water in the concrete due to high
temperature regime.
It is also observed that NAF showed a decreasing trend for first
14 and 28 days of heat exposure and there is no variation for further
heat treatment. In addition to this, concrete sample D (HDC:
4200 kg/m3) showed the highest neutron attenuation compared to
sample A (HDC: 2300 kg/m3), though the final water content is
nearly same. This result can be attributed to the fact that not only
water molecules in concrete but also elements like Fe present in
Hematite aggregates play significant role in neutron attenuation.
by ICRP-21.
In simulation study, two concrete samples, corresponds to lower
and higher density, namely 2300 and 3600 kg/m3were studied.
Elemental composition data used in this study is given in Table 3,
obtained based on chemical analysis of the concrete samples.
Table 4 presents the results obtained with MC simulation for
each type of concrete exposed for 14 days. The simulated NAF
values for ordinary density concrete are closely matching within
±20%. In modeling the density is assumed to be uniform
throughout the sample block. But in actual condition, there could
be non-uniformity due to the heterogeneity in the distribution of
pores where free water gets trapped. This could be the reason for
the observed deviation NAF values obtained after 14 days heat
treatment. Thus, the theoretical simulation values validated the
measured values.
3.3. Simulation studies
Neutron transport in concrete medium is studied using Monte
Carlo code N-particle radiation transport code, MCNP (Briesmeister,
1989). Interaction of particles in material is considered statistically
and parameters such as position, energy deposition, direction, type
of collision are estimated from known probability distributions. The
outcome of the simulation is estimation of particle flux averaged
over many trials. The exact geometry of the experimental setup was
simulated as realistically as possible and 107 particles were tracked
to reduce the error in simulation. Point detector (F5 tally) was used
to compute the particle flux and flux-dose conversion factor given
4. Conclusions
The effect of heating on the neutron attenuation characteristics
of high density concrete samples was investigated in this paper for
a sustained and cyclic temperature regime of 120 C. NAF increased
with density of concrete samples from 0.799 to 0.834. When
samples were subjected to heat treatment the change in NAF values
is proportional to the duration of heat treatment. The investigation
also showed the loss of water content was only during the first 14
days of heat treatment in all the samples. NAF values showed a
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S.P. C. et al. / Progress in Nuclear Energy 93 (2016) 76e83
Fig. 7. Variation of NAF with heat treatment.
Table 3
Elemental composition data.
Element
Si
Ca
Al
Fe
Mg
S
Na
K
P
H
O
Wt % for ordinary concrete
(2300 kg/m3)
Wt % for high density
concrete (3600 kg/m3)
Before HT
After HT
Before HT
After HT
15.95
19.28
1.853
3.81
1.7
0.46
0.89
0.79
0.03
4.55
50.6
19.46
22.25
1.99
4.35
2.26
0.46
1.54
0.8
0.06
2.51
44.35
2.57
8.63
0.714
52.16
1.33
0.23
0.21
0.094
0.0021
3.74
31
2.3
8.84
0.572
53.42
1.42
0.24
0.15
0.084
0.002
2
31
decreasing trend for first 14 and 28 days of heat exposure and there
is no variation for further heat treatment. This may be due to the
expected loss of free water present in the concrete samples.
Treatment beyond 28 days, i.e, 56 days did not have much influence
on NAF. The magnitude of variation in the NAF factor is prominent
in lower density samples than in the higher density samples. The
experimental NAFs were validated with theoretical simulation. The
results of this study clearly indicate that the sustained and cyclic
heat treatment at 120 C up to 56 days, decreases NAF values by a
factor of two. Whenever, concrete is subjected to such temperature
this factor should be considered in shielding design calculations for
pool type of reactors where sustained and cyclic temperature is
more plausible.
Table 4
Neutron attenuation factor obtained using MC simulation.
Sample
Time of heat exposure (days)
Neutron attenuation factor (NAF)
Expt.
Concrete (2300 kg/m3)
3
HDC (3600 kg/m )
0
14
0
14
0.799
0.601
0.829
0.630
Deviation (%)
MC
±
±
±
±
0.001
0.012
0.011
0.014
0.831
0.714
0.844
0.727
±
±
±
±
0.015
0.013
0.011
0.012
3.85
15.83
1.78
13.34
S.P. C. et al. / Progress in Nuclear Energy 93 (2016) 76e83
Acknowledgement
The authors wish to thank, Director, Indira Gandhi Centre for
Atomic Research, without whose support, this work would not have
been possible.
References
Briesmeister, J.F., 1989. MCNP- Monte Carlo Neutron and Photon Transport Code.
ORNL. RSIC-CCC 200.
Divya Rani, S., Santhanam, Manu, Venkatachalapathy, V., Sivathanu Pillai, C.,
February 2013. Permissible temperature in fast breeder reactor vault concrete.
Indian Concr. J. 11e20.
Gencel, O., 2012. Effect of elevated temperatures on mechanical properties of high
strength concrete containing varying proportions of hematite. Fire Mater. 36
(3), 217e230.
Gencel, Osman, Brostoe, W., Ozel, C., Filiz, M., 2010. An investigation on the concrete
properties containing colemanite. Int. J. Phys. Sci. 5 (3), 216e225.
Gencel, Osman, Brostow, Witold, Ozel, Cengiz, Filiz, Mümin, 2010. Concretes containing hematite for use as shielding barriers. Mater. Sci. 16 (No. 3).
Gencel, Osman, Bozkurt, Ahmet, Kam, Erol, Korkut, Turgay, 2011. Determination and
calculation of gamma and neutron shielding characteristics of concretes containing different hematite proportions. Ann. Nucl. Energy 38, 2719e2723.
Ichikawa, Y., England, G.L., 2004. Prediction of moisture migration and pore pressure build-up in concrete at high temperatures. Nucl. Eng. Des. 228, 245e259.
IS 8112 -2013,Indian Standard 43 grade Portland Cement-Specification.
Kaplan, M.F., 1989. Concrete Radiation Shielding”, Concrete Design and Construction Series. Longman Scientific & Technical, New York, New York.
83
Kharita, M.H., Takeyeddin, M., Alnassar, M., Yousef, S., 2008. Development of special
radiation shielding concretes using natural local materials and evaluation of
their shielding characteristics. Prog. Nucl. Energy 50, 33e36.
Koksal, F., Gencel, O., Lobland, H.E.H., Brostow, W., 2012. Effect of high temperature
on mechanical and physical properties of lightweight cement based refractory
including expanded vermiculite. Mater. Res. Innovations 16 (1), 7e13.
Korkut, T., Karabulut, A., Budak, G., Aygun, B., Gencel, O., Hancerliogullari, A., 2012.
Investigation of neutron shielding properties depending on number of boron
atoms for colemanite, ulexite and tincal ores by experiments and FLUKA Monte
Carlo simulations. Appl. Radiat. Isotopes 70 (1), 341e345.
Korkut, T., Gencel, O., Kam, E., Brostow, W., 2013. X ray, gamma and neutron radiation tests on Epoxy-Ferrochromium slag composites. Int. J. Polym. Analysis
Charact. 18 (3), 224e231.
Lee, J., Xi, Y., William, K., Jung, Y., 2009. A multi scale model for modulus of elasticity
of concrete at high temperatures. Cem. Concr. Res. 39, 754e762.
Mohammed Haneefa, K., Santhanam, Manu, Parida, F.C., 2013. Review of concrete
performance at elevated temperature and hot sodium. Nucl. Eng. Des. 258,
76e88.
D.J. Naus,”A Compilation of Elevated Temperature Concrete Material Property Data
and Information for Use in Assessments of Nuclear Power Plant Reinforced
Concrete Structures”, NUREG/CR-7031,ORNL/TM-2009/175.
Samarin, Aleksander, 2013. Use of concrete as a biological shield from ionising radiation. Energy Environ. Eng. 1 (2), 90e97.
Sumita, Junya, Sawa, Kazuhiro, Takada, Eiji, Tada, Keiko, March 2000. An investigation of the effects of water content on the shielding performance of the
primary upper shield in the high temperature engineering test reactor (HTTR).
J. Nucl. Sci. Technol. (Suppl. 1), 73e77.
Velusamy, K., ChellapandI, P., Chetal, S.C., Raj, Baldev, April 2010. Overview of pool
hydraulic design of Indian prototype fast breeder reactor. Sadhana 35 (Part 2),
97e128 (Indian Academy of Sciences).