**2.1 Sample preparation**

Before the mixing procedure, a part of mixing water at the percentage of water absorption capacity of expanded vermiculite aggregate by weight was added to vermiculite to make it fully saturated with water. Fig.1 shows the expanded vermiculite particles saturated with water.

Fig. 1. Expanded vermiculite fine aggregate saturated with water.

Then, the rest of the mixing water cement and silica fume or steel fiber were mixed together for 1 minute in a mixer, and finally, expanded vermiculite aggregate saturated with water

Neutron Shielding Properties of Some Vermiculite-Loaded New Samples 317

**Properties of samples including fiber steel (200C)** 

4F0 4 0 4F15 4 1.5 % 6F0 6 0 6F15 6 1.5 % 8F0 8 0 8F15 8 1.5 % **Properties of samples including silica fume (200C)** 

4S0 4 0 4S15 4 5 % 6S0 6 0 6S15 6 5 % 8S0 8 0 8S15 8 5 %

For neutron transmission measurement, we used a 241Am/Be neutron source and a Canberra portable neutron detector equipments. 241Am/Be source emits 4.5 MeV neutron particles. Physical form of 241Am/Be neutron source is compacted mixture of americium oxide with beryllium metal. Fast neutrons are produced by following nuclear reaction,

> ସܤଽ ݁ሺןǡ ݊ሻ ܥ ଵଶ

5.486 keV maximum energy alpha particles emitting from 241Am. Neutron energy value produced by this nuclear reaction is 4.5 MeV. Radiation characteristics of 241Am/Be neutron source are shown in Table.2 (Dose rate values have been obtained from The Health Physics

The NP-100B detector provides us to detect slow and fast neutrons. Tissue equivalent dose rates of the neutron field can be measured by it. The detectors contain a proportional counter which produces pulses resulting from neutron interactions within it. The probes contain components to moderate and attenuate neutrons. So that the net incident flux at the proportional counter is a thermal and low epithermal flux representative of the tissue equivalent dose rate and the neutron field. Because of neutrons have no charge; they can only be detected indirectly through nuclear reactions that create charged particles. The NP100B detector uses 10B as the conversion target. The charged particle – alpha or proton (respectively) created in the nuclear reaction ionizes the gas. Typical detector properties are shown in Table.3. Equivalent dose rate measurement results have read on RADACS

and Radiological Health Handbook, Scintra \_Inc., Revised Edition, 1992.).

program in system PC. Experimental design is shown in Fig.4.

Ratio Fiber Volume Fraction

Ratio Silica Fume Contents

Code of Sample Vermiculite /Cement

Code of Sample Vermiculite /Cement

Table 1. Codes and Properties of Samples

**2.2 Experimental design** 

was added to cement slurry and mixed for 3 minutes again, to get a homogenous structure. Fig.2 shows the fresh state of the mixture of lightweight mortar.

Fig. 2. Fresh state of lightweight mortar prepared with expanded vermiculite aggregate.

The prepared fresh mortar were cast in standard cube (with an edge of 150 mm) molds, in two layers, each layer being compacted by self-weight on the shaker for 10 s. All the specimens were kept in moulds for 24 h at room temperature of about 20oC, and then demoulded, and after demoulding all specimens were cured in water at 23 ±2 oC for 27 days. After 28 days curing, three plate specimens with a dimension of 150x100x20mm for neutron dose transmission measurements were obtained by cutting the cube specimens using a stone saw. Plate specimens obtained by the way was illustrated in Fig.3. We obtained 12 different samples. Codes and contents of samples were shown in Table.1.

Fig. 3. View of the samples


Table 1. Codes and Properties of Samples

#### **2.2 Experimental design**

316 Nuclear Reactors

was added to cement slurry and mixed for 3 minutes again, to get a homogenous structure.

Fig. 2. Fresh state of lightweight mortar prepared with expanded vermiculite aggregate.

The prepared fresh mortar were cast in standard cube (with an edge of 150 mm) molds, in two layers, each layer being compacted by self-weight on the shaker for 10 s. All the specimens were kept in moulds for 24 h at room temperature of about 20oC, and then demoulded, and after demoulding all specimens were cured in water at 23 ±2 oC for 27 days. After 28 days curing, three plate specimens with a dimension of 150x100x20mm for neutron dose transmission measurements were obtained by cutting the cube specimens using a stone saw. Plate specimens obtained by the way was illustrated in Fig.3. We obtained 12 different samples. Codes and contents of samples were shown in Table.1.

Fig.2 shows the fresh state of the mixture of lightweight mortar.

Fig. 3. View of the samples

For neutron transmission measurement, we used a 241Am/Be neutron source and a Canberra portable neutron detector equipments. 241Am/Be source emits 4.5 MeV neutron particles. Physical form of 241Am/Be neutron source is compacted mixture of americium oxide with beryllium metal. Fast neutrons are produced by following nuclear reaction,

#### ସܤଽ ݁ሺןǡ ݊ሻ ܥ ଵଶ

5.486 keV maximum energy alpha particles emitting from 241Am. Neutron energy value produced by this nuclear reaction is 4.5 MeV. Radiation characteristics of 241Am/Be neutron source are shown in Table.2 (Dose rate values have been obtained from The Health Physics and Radiological Health Handbook, Scintra \_Inc., Revised Edition, 1992.).

The NP-100B detector provides us to detect slow and fast neutrons. Tissue equivalent dose rates of the neutron field can be measured by it. The detectors contain a proportional counter which produces pulses resulting from neutron interactions within it. The probes contain components to moderate and attenuate neutrons. So that the net incident flux at the proportional counter is a thermal and low epithermal flux representative of the tissue equivalent dose rate and the neutron field. Because of neutrons have no charge; they can only be detected indirectly through nuclear reactions that create charged particles. The NP100B detector uses 10B as the conversion target. The charged particle – alpha or proton (respectively) created in the nuclear reaction ionizes the gas. Typical detector properties are shown in Table.3. Equivalent dose rate measurement results have read on RADACS program in system PC. Experimental design is shown in Fig.4.

Neutron Shielding Properties of Some Vermiculite-Loaded New Samples 319

system. And then we measured for each sample neutron equivalent dose rate while there is our sample between 241Am-Be source box and detector probe. The ratio of two values is

Nowadays concrete is often used in radiation shielding process. In several studies, some additive materials were added in concrete to increase its radiation shielding capacity. In this study, as an additive material, we have used vermiculite mineral with a good heat insulation material. Produced samples have three different vermiculite and cement ratio values. 4.5 MeV neutron dose transmission values (Fig.5) and attenuation lengths of samples (Table.4) were obtained. Attenuation length is just equal to the average distance a particle travels before being scattered or absorbed. It is a useful parameter for shielding calculations. Also we calculated experimental 4.5 MeV neutron total macroscopic cross sections (µ) using by dose transmission values. The various types of interactions of neutrons with matter are

<sup>0</sup> *totalx*

<sup>0</sup> ln

where I0 is known as beam intensity value, at a material thickness of x = 0. Equivalent dose rate has been used instead of beam intensity because of our equivalent dose rate measurements. Experimental 4.5 MeV neutron total macroscopic cross sections were shown

*I I x*

**4F0 4F15 4S0 4S15 6F0 6F15 6S0 6S15 8F0 8F15 8S0 8S15**

Sample

*total*

..... Σ =Σ +Σ +Σ + *total scatter capture fission* (1)

*XI Ie*−Σ <sup>=</sup> (2)

Σ = (3)

called dose transmission.

in Table.5.

**3. Results and discussion** 

combined into a total cross-section value:

The attenuation relation in the case of neutrons is thus:

Fig. 5. 4.5 MeV neutron dose transmissions for each samples

**0,82 0,84 0,86 0,88 0,90 0,92 0,94 0,96 0,98**

Dose Transmission

#### Fig. 4. Experimental Setup


\*http://www.stuarthunt.com/pdfs/Americium\_241Beryllium.pdf

Table 2. Radiation characteristics of 241Am-Be neutron source\*


\*http://www.canberra.com/pdf/Products/RMS\_pdf/NPSeries.pdf

Table 3. Typical Properties of Detector\*

We determined dose transmission values of vermiculite loaded samples. Firstly, we counted equivalent dose rate by fast neutrons while there is no sample between source and detector system. And then we measured for each sample neutron equivalent dose rate while there is our sample between 241Am-Be source box and detector probe. The ratio of two values is called dose transmission.

#### **3. Results and discussion**

318 Nuclear Reactors

**Physical Half-Life: 432.2 years Specific Activity: 127 GBq/g** 

Neutron - 4.5 MeV 2

**Specifications of Canberra NP100B Neutron Detector Detector Type** BF3 Proportional Counter **Detector Sensitivities** 0–100 mSv/h (0–10 Rem/h) **Energy Range** 0.025 eV – 15 MeV **Operating Temperature Range** –10 °C to +50 °C (+14 °F to +122 °F)

**Weight kg (lb)** 10 kg (22 lb)

**Detector Linearity** ±5% **Accuracy** ±10%

**(internally generated)** 1750–1950 V

We determined dose transmission values of vermiculite loaded samples. Firstly, we counted equivalent dose rate by fast neutrons while there is no sample between source and detector

**Housing** Moisture Proof Aluminum **Operating Humidity** 0–100% non-condensing

59.5 (35.9%) - -

5.486 (85.2%) - <sup>85</sup>

**Eeff Dose Rate** 

244 x 292 mm (9.6 x 11.5 in.)

**(**μ**Sv/h/GBq at 1m)** 

**(keV)** 

\*http://www.stuarthunt.com/pdfs/Americium\_241Beryllium.pdf Table 2. Radiation characteristics of 241Am-Be neutron source\*

Fig. 4. Experimental Setup

**Principle Emissions Emax** 

Gamma/X-Rays 13.9 (42.7%)

Alpha 5.443 (12.8%)

**Size (mm.) (Dia. x inch)** 

**High Voltage Supply** 

Table 3. Typical Properties of Detector\*

\*http://www.canberra.com/pdf/Products/RMS\_pdf/NPSeries.pdf

Nowadays concrete is often used in radiation shielding process. In several studies, some additive materials were added in concrete to increase its radiation shielding capacity. In this study, as an additive material, we have used vermiculite mineral with a good heat insulation material. Produced samples have three different vermiculite and cement ratio values. 4.5 MeV neutron dose transmission values (Fig.5) and attenuation lengths of samples (Table.4) were obtained. Attenuation length is just equal to the average distance a particle travels before being scattered or absorbed. It is a useful parameter for shielding calculations. Also we calculated experimental 4.5 MeV neutron total macroscopic cross sections (µ) using by dose transmission values. The various types of interactions of neutrons with matter are combined into a total cross-section value:

$$
\Sigma\_{\text{total}} = \Sigma\_{\text{scatter}} + \Sigma\_{\text{capture}} + \Sigma\_{\text{fission}} + \dots \tag{1}
$$

The attenuation relation in the case of neutrons is thus:

$$I\_X = I\_0 e^{-\Sigma\_{total} x} \tag{2}$$

$$\Sigma\_{\text{total}} = \frac{\ln\left(\frac{I\_0}{I}\right)}{\infty} \tag{3}$$

where I0 is known as beam intensity value, at a material thickness of x = 0. Equivalent dose rate has been used instead of beam intensity because of our equivalent dose rate measurements. Experimental 4.5 MeV neutron total macroscopic cross sections were shown in Table.5.

Fig. 5. 4.5 MeV neutron dose transmissions for each samples

Neutron Shielding Properties of Some Vermiculite-Loaded New Samples 321

1. Vermiculite mineral has high-level thermal insulation capacity. Concrete isn't decomposing with vermiculite addition. This mineral can be used as an additive for

2. According to the experimental results, neutron shielding property of concrete increase

3. To produce good materials which have high radiation shielding capacity and thermal insulation property, vermiculite and fiber steel may be doped in mortar. These

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materials can be used for neutronic and thermal applications.

materials", *Mater Design,* 32 (2011) 4354–4361.

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with increasing fiber steel and silica fume content.

**4. Conclussions** 

**5. References** 

203.

57.

radiation shielding process.

*Res* 39 (2005) 2643–2653.

*Dos*., 115(1-4) (2005) 58-61.

and Design 241 (2011) 2359–2363.




Table 5. 4.5 MeV neutron total macroscopic cross sections

As can be seen from Fig.5 and Table.4, dose transmission values and attenuation lengths decrease with increasing fiber steel and silica fume contents. This result indicates that neutron shielding capacity of samples is increased by silica and steel amount. According to the results, there is not a consistent relationship between vermiculite content and neutron shielding capacity of samples except of F15-samples. The sample named 8F15 is the best neutron attenuator in all specimens. The reason of this that, this sample has higher vermiculite and fiber steel content than others. The worst sample is 8S0 which has higher vermiculite but lower silica fume content. As a result, to increase neutron shielding capacity of sample, expanded vermiculite and fiber steel may be added in the mortar.
