A comparative study of the attenuation of reactor thermal neutrons in different types of concrete

~ Pergamon

Ann. Nucl. Energy Vol. 23, No. 14, pp. 1189-1195, 1996 Copyright © 1996 Elsevier Science Ltd 0306-4549(95)00122-0 Printed in Great Britain. All rights reserved 0306-4549196 $15.00 + 0.00

T E C H N I C A L NOTE A COMPARATIVE STUDY OF THE ATTENUATION OF REACTOR THERMAL NEUTRONS IN DIFFERENT TYPES OF CONCRETE

I. I. B A S H T E R I, A. E L - S A Y E D A B D O 2 a n d A. S. M A K A R I O U S 2 Department of Physics, Faculty of Science, Zagazig University, Zagazig, Egypt 2 Reactor and Neutron Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo, Egypt

(Received 2 November 1995) A b s t r a c t ~ h i s study was carried out to assess the distribution of thermal neutrons emitted directly from the core of the ET-RR-I reactor in ordinary concrete, ilmenite concrete and ilmenite-limonite concrete shields. Measurements were carried out by using a direct beam and a cadmium filtered beam of reactor neutrons. The neutron dose distributions were measured using Li2B4OT: Mn thermoluminescent dosimeters. The data obtained show that ilmenite concrete is better for slow and thermal neutron attenuation than both ordinary and ilmenite-limonite concretes. Also it was concluded that thermal neutrons emitted directly from the reactor core are highly absorbed within the first few centimeters of each type of concrete. The thickness of ilmenite concrete required to attenuate the doses of neutrons to a certain value along the beam axis for a direct reactor beam estimated to be about 75 and 57% of the shield thickness made from ordinary and ilmenite-limonite concretes, respectively. Empirical formulae were derived to calculate the neutron dose distributions in ordinary, ilmenite and ilmenitedimonite concrete shields both along and perpendicular to the beam axis for both the direct reactor neutrons and the reactor thermal neutrons. Copyright © 1996 Elsevier Science Ltd

1. I N T R O D U C T I O N

The a t t e n u a t i o n a n d distribution o f t h e r m a l a n d slow n e u t r o n s p r o d u c e d inside the concrete shield are a m o n g the m a i n r e q u i r e m e n t s for reactor shield design since this will in t u r n determine the m a i n source o f capture g a m m a rays. T h e t h e r m a l n e u t r o n distribution in the b o d y o f the shield is due to the t h e r m a l n e u t r o n s which enter the shield directly f r o m the reactor core a n d those which are p r o d u c e d by the slowing d o w n o f intermediate a n d fast n e u t r o n s ( L a m a r s h , 1975). Different types o f concretes are the m o s t c o m m o n l y used shielding materials. It is structurally satisfactory a n d relatively inexpensive w h e n c o m p a r e d with o t h e r shield materials. In addition, the aggregates used in a concrete c a n be selected, within limits, to achieve n e u t r o n a n d g a m m a a t t e n u a t i o n properties a p p r o p r i a t e for specific applications. One p r o b l e m in the use o f concrete is t h a t o f predicting the a t t e n u a t i o n properties of the solid material. In p a r t this is due to uncertainties in the c o m p o s i t i o n resulting from variations in the feed material a n d inhomogeneities in the p l a c e m e n t (Jaeger, 1975). E x p e r i m e n t a l studies o f t h e r m a l n e u t r o n dose distributions in ordinary, ilmenite a n d ilmenite-limonite concrete shields with densities o f 2.3, 4.6 a n d 2.9 g/cm 3, respectively, have recently been carried o u t ( M e g a h i d et al., 1981, 1989; K a n s o u h , 1988). In the present work, a c o m p a r a t i v e study o f reactor t h e r m a l n e u t r o n doses in ordinary, ilmenite a n d i l m e n i t e - l i m o n i t e concretes was carried o u t using d a t a from the a b o v e literature. Such studies were a i m e d to assess the effect o f reactor t h e r m a l n e u t r o n dose distribution in different concrete media. A semiempirical f o r m u l a has been derived for each type o f the concrete m e d i u m to calculate 1189

1190

Technical Note

the neutron doses along the reactor beam axis and those perpendicular to it for both the direct reactor neutron beam and the reactor thermal neutron beam.

2. EXPERIMENTAL DETAILS

Experimental measurements of reactor thermal neutron distributions for ordinary, ilmenite and ilmenite-limonite concrete shields have been carried out. The measurements were performed using one of the horizontal channels of the Egyptian thermal research reactor number 1 (ET-RR-1) with a 2 MW full power both along the beam direction (Z-direction) and perpendicular to the beam direction (R-direction). Three blocks of each concrete type with dimensions of 120 x 120 x 40 cm 3, were placed together to construct a shielding assembly with dimensions 120 x 120 x 120 cm 3. Each concrete block had two vertical holes, 20 cm apart, for placing the detector holders. A schematic diagram of the experimental apparatus is shown in Fig. 1. The compositions of the different concrete media are given elsewhere (Makarious et al., 1989). The neutron flux distributions were measured using dosimeters consisting of discs of Li2B4OT: Mn thermoluminescent together with LiF-7 thermoluminescent dosimeters for correction of the gamma response. Measurements have been carried out using a direct reactor neutron beam (i.e. a reactor neutron beam which has not passed through any filter) and then a reactor neutron beam which has passed through a cadmium filter. The difference between the two above neutron beams gives the reactor thermal neutrons. A Zn S(Ag) scintillator was used as a monitor to detect the reactor power fluctuations. Corrections due to uncertainties in the sample position and the statistical error in counting the TLD response have been taken into consideration. It was found that the measurement errors are within + 6% except at low intensities where the errors are within ___10%.

3. RESULTSAND DISCUSSIONS Figure 2 shows measured and calculated neutron dose distributions in ordinary, ilmenite and ilmenite-limonite concretes for a direct reactor beam both along the Z-direction and in the R-direction. Figure 3 shows the measured neutron dose distributions in ordinary, ilmenite and ilmenite-limonite concretes for a cadmium-filtered beam both in the Z- and R-directions. The reactor thermal neutron doses distributions in the Z- and R-directions were obtained by subtracting the neutron doses of the cadmium-filtered reactor beam from the corresponding values for the direct reactor beam as shown in Fig. 4. The dose attenuation curves in Figs 2 and 3 show that the neutron dose intensity decreases exponentially with the increase of the concrete thickness for values of R<20 cm and Z> 10 cm. From Figs 2 and 3 it can be seen that the slopes of the attenuation curves decrease when a cadmiuin-filtered beam was used. The relaxation lengths for the cadmium-filtered beam are higher than those for a direct reactor beam. The slopes of the attenuation curves decrease with increasing R values. This may be attributed to the fact that, at large values of R, the effect of accumulated scattered thermal neutrons produced from the thermalization of fast neutrons becomes more pronounced. From Figs 2 and 3 it is clear that the cadmium filter provides the main contribution in decreasing the neutron dose intensity within the first 50 cm of the concrete shield. The neutron dose distributions in ilmenite concrete decrease exponentially for Z>50 cm in both direct and cadmium-filtered reactor beams as shown in Figs 2 and 3. The depression in the attenuation of the neutron doses for Z<50 cm in ilmenite concrete may be due to the capture of reactor thermal neutrons in iron punchings added to the ilmenite concrete to improve its attenuation effectiveness to gamma rays. This depression was compensated by the new thermalized neutrons as a result of the inelastic scattering of fast neutrons in ilmenite concrete as shown in Fig. 4. The reactor thermal neutron dose distributions were shown in Fig. 4. This figure shows that the reactor thermal neutron dose intensity decreases exponentially with increasing concrete thickness for R = 0 and 10 cm at Z>30 cm. Thus, the reactor thermal neutrons provide main contribution to the direct reactor beam at Z<30 cm. The reactor thermal neutron doses decrease with the increase of the concrete thickness. In general the slope of the attenuation curves increase with the increase of the concrete density. We can conclude that ilmenite concrete (density---4.6g/cm 3) is a better neutron attenuator than both the ordinary and ilmenite-limonite concretes (densities =2.3 and 2.9 g/cm 3, respectively), especially at deep penetrations. The following empirical formulae were derived to calculate the direct reactor neutron dose distributions:

Technical Note 1

2

3

4

5

6

1191

7

8

9

-71

--~'Z

,,,.

I

"~i

I.

400

r ' - - I k - ' ',,~

cm

r-' ~'~~

-

"

l-Reactor core

2-Water cooling

3--Cast iron wall

4-Concrete wall surrounding the reactor

5-Experimental channel

6-Cast iron rotating gates 8-1nvestigated concrete blocks

7-Collimator

I 0--Canette bed

9-Bolt system 1 l-Regulators of the cassette height

Fig. 1, A schematic diagram of the experimental layout. Calculated

Measured

.

10 6

Ordinary Ilmenite nmenite-limonite

o X

105

",~

\@

10 4

.

\

,% \ \ ,.4

O

\ \ R (cm)

,e

~

_= o ~

.,~ 10 30 20

lO t 30

1001 0

I

I

lO

30

50

30

I

I

70

90

Z (cm) Fig. 2. Measured and calculated neutron doses in ordinary, ilmenite and ilmenite-limonite concretes along a direction parallel to the beam axis using a direct reactor neutron beam.

1192

Technical Note 10 6

,~

10 5

"~

104

e~ O

.~

Ios

io 2

0

I01

i0 o 0

I0

30

50

70

90

Z (cm) Fig. 3. Measured neutron doses in ordinary, ilmenite and ilmenite limonite concretes along a direction parallel to the beam axis using a reactor neutron beam filtered with cadmium.

lO-3RZ-O.1166R-O.O865Z]

(1)

D(il.) = 4.65 × 107 x exp[0.0506R ln(Z + 0.1) - 1.63x//Z - 0.253R]

(2)

D(il.-li.) =5.8 × 105 x e x p [ - 0 . 1 0 6 Z - 0 . 1 4 6 R x e x p ( - 0.0146Z)]

(3)

D(or.) = 7.67 x 105 x exp[1.068 x

where D(or.), D(il.) and D(il.-li.) are the direct reactor neutron doses (in thermoluminescent response) for ordinary, ilmenite and ilmenite-limonite concretes, respectively, and the units of Z and R are both cm. The statistical errors of equations (1), (2) and (3) were calculated and found to be within + 8 , + 7 and + 9 % , respectively. Calculated direct reactor neutron doses using the above formulae are compared with the corresponding measured doses in Fig. 2. The empirical formulae derived to calculate the reactor thermal neutron dose distributions are: DR.th.(Or.) = 6.02 X 105 X exp[5 x 10-4RZ - 0.114R -- 0.107Z]

(4)

DR th (il.) = 3.16 × 10 s × exp[ -- 0.152Z-- 0.16R x exp( -- 0.027Z)]

(5)

3RZ--O.14R--O.13Z]

(6)

DR.th.(il.--li.) = 3.2 × l0 s × exp[1.4 × 10

where DR th.(Or.), DR.th.(il.) and DR.th.(il.--li.) are the reactor thermal neutron doses (in thermoluminescent response) for ordinary, ilmenite and ilmenite-limonite concretes, respectively (R and Z are in cm). The statistical errors o f equations (4), (5) and (6) were calculated and found to be within + 8, + 8 and + 9%, respectively. Calculated reactor thermal neutron doses using the above formulae are compared with the corresponding measured doses in Fig. 4. G o o d agreement between measured and calculated dose values have been obtained as shown in Figs 2 and 4. Figure 5(a) represents the relation between the neutron doses for the three concrete types for both direct and cadmium filtered reactor beams where there is a difference of a factor of 10 in the dose intensity. The

Technical Note

1193

10 6

Measured

!

". m r~

1o 5

Calculated Ordinary llmenite llmenite-limonite

o \

N

% %

104

,

i0 3

%\xX~_\ \ \

\

°~ o

o

\

"?

g 102

\N

o

t %, \

u

t-

R (cm) "o

"'~',

101

"~'i

lOo I 0

I

I

10

30

I 50

30 20 I 7O

"\o

",. i0 N~

20

.o .10 .20 -30

9O

Z (em) Fig. 4. Measured and calculated neutron doses in ordinary, ilmenite and ilmenite limonite concretes along a direction parallel to the beam axis for reactor thermal neutrons.

cadmium tends to decrease the dose intensity especially at small penetration distances. Along the beam axis (i.e. at R =0) the cadmium filter causes a decrease of Z thicknesses at high intensity (10 TLD response) of about 10%0 of the thicknesses when the direct reactor beam is used while at low intensity (102 T L D response) the decrease is only about 2% for all concrete types used. Figure 5(b) represents the relation between the neutron doses for the three concrete types for both reactor thermal neutrons and the neutrons from the direct reactor beam again with a difference of a factor of 10 in the dose intensity. The figure shows that the reactor thermal neutrons provide the main contribution to neutrons from a direct beam at the first penetration distances of the concrete shield. The decrease in Z thicknesses along the beam axis due to the reactor thermal neutrons at 104 TLD response is about 24% of the thicknesses when the direct reactor beam is used while at 102 TLD response the decrease is about 20% for all types of concrete. It can be concluded that for a direct reactor beam, the thickness of ilmenite concrete required to attenuate the neutron dose intensity to a particular value along the reactor beam axis (i.e. at R =0) is respectively about 75 and 57% of the thickness required when the shield is made of ordinary and ilmenite-limonite concretes. Figure 6 represents the ratio between the reactor thermal neutron doses and the neutron doses using a direct reactor beam at R =0, 10, 20 and 30 cm. At R = 0 and Z = 10 cm the ratio decreases with the increase of the concrete density. The reactor thermal neutrons provide the main contribution to the neutron doses at this point in ordinary and ilmenite-limonite concretes. This contribution is about 59, 45 and 22% for ordinary, ilmenite-limonite and ilmenite concretes, respectively. For ordinary and ilmenite-limonite concretes, the ratio decreases with increasing concrete thickness along the beam axis. For ilmenite concrete the ratio increases with increasing concrete thickness and reaches its maximum at Z = 30 cm and R = 2 0 cm since the reactor thermal neutrons absorbed were compensated by the newly thermalized neutrons generated from the inelastic scattering of fast neutrons in the iron punching of ilmenite

Technical Note

1194

(a) - - -o- --o---

--×--

Ordinary Z (cm) Ilmenite Ilmenite-limonite ~

.~

~

o- ~

.o- lOOT--

- ""

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1

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./x~

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(b)

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80 J , direct beam

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10

~ x

~2

",'°-2

.~/,% 20

"

""~,~,3 *'2 0

10

20

30

R (cm)

Fig. 5. Dose curves differing by a factor of "10" in ordinary, ilmenite and ilmenite-limonite concretes for: (a) neutrons from direct and cadmium-filtered neutron beams, and (b) neutrons from direct neutron beam and reactor thermal neutrons. [1 = 102, 2 = 103 and 3 = 104 thermoluminescent response in arbitrary unit.]

c o n c r e t e s . T h e c o n t r i b u t i o n o f r e a c t o r t h e r m a l n e u t r o n s at Z = 90 c m a n d R = 0 c m is o n l y a b o u t 15, 7 a n d 4 % in o r d i n a r y , i l m e n i t e a n d i l m e n i t e - l i m o n i t e c o n c r e t e s , respectively. F o r o r d i n a r y c o n c r e t e at Z = 10 c m t h e r a t i o i n c r e a s e s w i t h i n c r e a s i n g v a l u e s o f R a n d r e a c h e s a m a x i m u m v a l u e (96%) a t R = 2 0 c m , t h e n d e c r e a s e s to 4 4 % a t R = 3 0 c m , s i n c e t h e a b s o r b e d r e a c t o r t h e r m a l n e u t r o n s in o r d i n a r y c o n c r e t e were n o t c o m p e n s a t e d b y t h e n e w l y t h e r m a l i z e d n e u t r o n s . F o r i l m e n i t e - l i m o n i t e c o n c r e t e t h e r e a c t o r t h e r m a l n e u t r o n s a r e m o r e p r o n o u n c e d a t Z = 10 c m as s h o w n in Fig. 6.

Acknowledgement--The authors are grateful to Professor R. M. Megahid, Reactor and Neutron Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo, Egypt, for his valuable comments and criticism. REFERENCES Jaeger R. G. (1975) Shielding Materials. Springer, Berlin. K a n s o u h W. A. (1988) M.Sc. Thesis, Faculty of Science, Zagazig Univ., Zagazig, Egypt. L a m a r s h J. R. (1975) Introduction to Nuclear Engineering. Addison-Wesley, U.S.A. Makarious A. S., EI-Kolaly M. A., Bashter I. I. and K a n s o u h W. A. (1989) Appl. Radiat. Isotopes 40, 257.

Technical Note ---o----o-----x-t~

o

ee

Ordinary Ilmenite Ilmenite-limonite

0.8 0.6

:~

1195

R = 30 cm

L~ 1 o - ' - - °.,,

0.4 0.2

~

0

o

1.0

i

0.8

\~,,.,o-.~,.

R -- 20 c m

0.6 0.4

o/ \

",o.\

0.2

I ~

0.8

~

0.6

~

0.4

~

0.2

~ O

I

-T~--#

o,, ~ - . o - - . ~

i

R = I0 em

o

•u

0.8

o

0.6 - e . .

.,o

I

R=Ocm

0.4 0.2

x.S~ ~o ......

of

10

30

50

70

90

I 110

Z (cm) Fig. 6. The ratio between the reactor thermal neutron dose and neutron dose using a direct reactor neutron beam in ordinary, ilmenite and ilmenite-limonite concretes for R = 0, 10, 20 and 30 cm.

Megahid R. M., Makarious A. S. and El-Kolaly M. A. (1981a) Ann. Nucl. Energy 8, 79. Megahid R. M., El-Kolaly M. A., Makarious A. S. and A b u E1-Nasr T. Z. (1981b) Int. J. Appl. Radiat. Isotopes 32, 507. Megahid R. M., EI-Essaly F. M. and E1-Akabawy U. A. (1989) Arab J. Nucl. Sci. Applic. 22, 20.