Optimization of thermal neutron source based on 6 MeV Linear Accelerator using FLUKA simulation

Transcription

1 Chapter 6. Optimization of thermal neutron source based on 6 MeV Linear Accelerator using FLUKA simulation In this chapter an Accelerator based pulsed thermal neutron source has been designed. Initially, electron incident on e - target to generates bremsstrahlung radiation and further neutrons were produced through photo nuclear reaction in - n target. The collisions of these neutrons with the moderating material shifts neutron energy to thermal energy range. To perform this design Monte Carlo based FLUKA code was used. The design was optimized by varying different parameters of the target and moderating materials for each region. Beryllium was optimized as photonuclear target and reflector, while polyethylene and graphite was optimized as a moderator to reduce the neutron energy to thermal energy range. To verify the simulated results, a prototype experiment was carried out using 6 MeV linear accelerator. The results of experiment and simulation are found to be in good agreement with each other. 138

2 Chapter 6. Optimization of thermal neutron source Importance and Objective Neutrons are used in various applications but mainly neutron diffraction and scattering provides valuable tool for probing the structure of bulk materials. Neutron activation analysis is another best technique for analysis of such materials, because it has good sensitivity for a large number of elements, and is non destructive too. When neutrons interact with matter, it can induce nuclear reaction and corresponding emitted radiations can be of the form of prompt effect or delayed. The analysis of prompt neutrons and prompt photons resulting from fast neutron inelastic scattering and thermal neutron absorption with the elements, is useful for detecting and identifying fissile material [1]. For these prompt measurements, the neutron generators ability to emit pulsed neutron field presents a significant advantage. Measurement of delayed gamma is performed for the determination of inorganic impurities content in oil and products from its processing [2]. Neutron scattering has proved to be a valuable tool for studying the molecular structure and motion of molecules of interest to manufacturing and life processes. Accelerators and nuclear reactors produce low-speed neutrons with wavelength appropriate to see structures of the size of magnetic microstructures and DNA molecules. The wavelength of fast neutron is too short for investigating the matter and wavelength of 25 mev neutron is 1.8 Å, which is of the same order as typical interatomic distances and is quite suitable for diffraction experiments [3]. Neutrons can penetrate deeply into bulk materials and use their magnetic moment or strong interaction forces to preferentially scatter from magnetic domains or hydrogen atoms in long chain nucleosomes. Neutron facilities throughout the world generate neutrons by using nuclear reactors, radioisotopes and high energy particle accelerators as a primary source. The nuclear reactors are the highest neutron yield source, but size, complexity and cost have limited their use. Although, radioisotope based neutron sources are running continuously, but they can not be used in the applications that require pulsed neutrons. In addition, such sources also have low neutron

3 Chapter 6. Optimization of thermal neutron source flux and can be utilized for very specific applications. However, the particle accelerator based neutron sources are vary in size and diversity. Because of the compactness, easy handling, adjusted flux with the beam parameters, no radioactive waste and less shielding an electron accelerator based thermal neutron source has been designed. The present work deals with the designing of accelerator based pulsed thermal neutron source for scattering experiments in analysis of element in various bulk materials. When electron beam from an accelerator incident on high Z (e γ ) target it generates a cascade shower of bremsstrahlung radiations. Further, interaction of these radiations with suitable photo neutron (γ n ) target results in to the emission of fast neutrons. Shifting neutron energies from fast to thermal is possible by means of neutron interaction with set of moderating and reflecting materials. A large number of neutron collisions are required to get thermal neutrons. In the design of neutron source, different materials and their respective dimensions are determined using Monte Carlo based FLUKA code. Mostly, neutron flux decreases due to neutron capture, neutron escape from the geometry and inverse square law Φ(r) (1/r 2 ). Thus, when designing of various regions of such neutron source, the challenge is to slow down the neutron energies by maintaining the neutron economy and low gamma production from the respective e γ and γ n targets. A prototype experiment is simulated in FLUKA code and the integrated neutron flux is measured experimentally with activation technique. 6.2 Literature Survey The literature survey indicates that fair amount of work has been done in the field of thermal neutron generation using accelerators. The thermal neutron facilities developed for research purpose are compiled in Table 6.1.

4 Chapter 6. Optimization of thermal neutron source Table 6.1: Review of Thermal Neutron production. No. Author (J.name,(year), Simulation/ Type Result Vol,pp)[ref] Experiment 1 Picton D.J. J.Phys. D:Appl Phys. (1982) Various materials have 15, [4] been tested for moderator 2 Danon Y. NIMA (1995) MCNP simulation for designing 60 MeV e on Ta target measured neutron flux for old and new 352, [5] enhanced thermal neutron target target geometry of cold moderator design 3 Agosteo S. NIMA (2002) MCNP simulation, measurement 7 MeV Deuteron on thermal and epithermal neutron flux 476, [6] through activation tech. Beryllium target measured. Designed for BNCT 4 Auditore L. NIMB,(2005), MCNP simulation 5 MeV Linac based Neutron fluence calculated is n/s/cm 2 /ma 229,137 [7] from Be and 1.5 times from BeD2 5 Bartalucci Sergio REPORT No. 1961/PN,(2005) MCNP and FLUKA 1 GeV e Linac on thermal neutron energy spectra Inst. Nucl. Phys. [8] simulation tantalum target calculated and angular distribution studied 6 Bartalucci S. NIMA (2007) MCNP simulation for design 500 MeV Linac based Moderator design for time 575, [9] of flight measurement experiment 7 Favalli A. Rad. Prot. Dosm. (2007) Characterization of D T reaction based Thermal neutron facility has been tested 126,74, [10] Thermal neutron facility thermal neutron 8 Grzegorz T. Appl.Rad.Iso.,(2009), MCNP simulation D-T reaction based pulsed Thermal neutron flux calculated and measured. 67,1148 [11] thermal neutron source

5 Chapter 6. Optimization of thermal neutron source Thermal neutron beam production Accelerator based neutron sources produced fast neutrons with energies in the MeV range through reaction between the incident high energy electron and target material. Since the areas where neutrons used mostly are scattering, diffraction and to see structures of the size of magnetic microstructures and DNA molecules requires thermal neutrons. The initial fast neutrons must be slowed down before interacting with the object. This slowing down process is called moderation. Moderation of neutrons is accomplished by allowing them to collide with nuclei, thereby transferring some of their energy in the process Neutron Moderation Neutrons with energies less than 10 MeV are traveling at velocities less than 0.1 the speed of light and can be treated non-relativistically. Since accelerator source produces neutrons in the range 100 kev to 4 MeV, all theoretical development will be from a classical perspective. Thus we have the following the energy-velocity relation for non-relativistic neutrons E = 1 2 mv2 (6.1) where E is kinetic energy, m is neutron mass, and v is velocity. Equation 6.1 show that a change in velocity is also a change in energy, thus, the slowing down process is an energy transfer from the neutron to the medium. If a neutron with initial energy E and velocity v collides with an atom of mass A initially at rest, then, using conservation of energy and momentum, the ratio of the neutron energy after the collision, E, and the initial energy, E, is E E = A Acosθ (A + 1) 2 (6.2) where θ is the scattering angle in the center of mass. When θ = 0 (no scattering) this ratio is 1, and when θ = 180 (maximum scattering), i,e., a head-on collision,

6 Chapter 6. Optimization of thermal neutron source Equation 6.2 becomes [ ] E E θ = = [ ] 2 A 1 (6.3) A + 1 Equation 6.3 can be used to compare the efficiency of energy transfer between a neutron and nuclides with different mass. The light elements are better at slowing down neutrons due to the larger energy transfer per collision. As a general rule this is true however when deciding on a moderator material one must also be aware of the possibility of neutron absorption, which will remove the neutron entirely. Neutron scattering cross-sections are essentially independent of scattering angle from neutrons below 10 MeV. The distribution of energy transfer E /E for one collision is uniform over the range (E /E) θ=180 0 to 1.0. If we consider many neutrons with the same initial energy, each subsequent collision also has a uniform energy transfer distribution, however, the neutrons are now themselves distributed in energy, which broadens the spectrum after the first collision. This can be evaluate quantitatively by defining a parameter ξ to be the average value of ln(e/e ) after each collision, ξ = [ln E ] E avg = ln [ (A+1) 2 A +1+2Acosθ] dω 2 dω (6.4) where dω is the solid angle in the center of mass and the scattering is assumed to be isotropic. The moderator material must have a high average logarithmic energy loss (ξ) is given by integrating Equation 6.4, [12], ξ = 1 + (A 1)2 2A ln [ ] A 1 A + 1 (6.5) The moderating material should have a considerable scattering cross section, (Σ s ) and less cross section for absorption (Σ a ), such that less number of neutrons are lost due to absorption. No existing material possesses all these properties. However, it is possible to combine these parameters and define a moderating

7 Chapter 6. Optimization of thermal neutron source ratio, R m, by means of the expression: R m = ξ Σ s Σ a (6.6) The moderating ratio, R m, is a relative measure of the capacity of a moderator in spreading neutrons without absorbing a great number of them. It should be as large as possible so that a good moderating material can be met. Based on the moderating properties, materials were selected for the optimization study of the moderating/reflecting system of neutrons generated through photo nuclear reaction. 6.4 Conceptual Design of Pulsed Thermal Neutron Source A Tungsten (W) target having thickness 0.22 cm (range of the 6 MeV electron in W target) is mounted in path of electron beam for the production of bremsstrahlung radiations. LINAC is assembled with primary collimator to collimate the photon beam. The bremsstrahlung spectrum at the end of primary collimator is estimated using FLUKA. The bremsstrahlung spectrum is shown in Figure 6.1. The integrated bremsstrahlung fluence is (photon cm 2 )/e. At first electron source along with electron to gamma converter, primary collimator and shielding of photon mode LINAC is modeled in FLUKA. The material that first interacts with gamma, forms the first region. The function of this region is to generate neutrons. The materials which having photonuclear reaction threshold less than 6 MeV are tested for first region as a photo nuclear target. The photo neutron production threshold energy varies in general from 8-19 MeV for light nuclei (A < 40) and 6-8 MeV for heavy nuclei [13]. But, for deuterium and beryllium, threshold energy is MeV and MeV respectively [14]. The cross section of (γ, n) reaction with beryllium and deuterium from threshold energy to 20 MeV have been measured and validated by IAEA [15, 16]. Therefore, in case of 6 MeV incident electron, the target choice

8 Chapter 6. Optimization of thermal neutron source Bremsstrahlung Fluence ((photon-mev -1 -cm -2) /e _ ) 3.0e-3 2.5e-3 2.0e-3 1.5e-3 1.0e-3 5.0e-4 0.0e+0 Calculated on collimator exit Bremsstrahlung Energy (MeV) Figure 6.1: Bremsstrahlung spectrum for a 0.22 cm thick tungsten target, calculated at primary collimator exit face. is strictly limited to few light elements such as deuterium and beryllium for neutron production. The first region is positioned such that the collimated gamma interacts perpendicularly with it. The neutron fluence and yield are studied and these are depending on thickness of target. The neutrons produced in this way redirected towards second region. The function of this region is to convert neutrons to softer spectrum. In addition, materials for second region are checked for the possibility of neutron production through (n, 2n) reaction to maintain magnitude (Φ(r).r 2 ) or even increase it with thickness. Second region is placed in such a way that it surrounds the first region. The thermal (< 0.3 ev) neutron fluence, epi-thermal (0.3 ev to 100 kev) neutron fluence and fast (> 100 kev) neutron fluence and their percentage contribution in terms of thermal neutron content (TNC), epithermal neutron content (ENC), fast neutron content (FNC) are calculated for different dimensions of the second region. The TNC describes the number of thermal neutrons within neutron beam. Thermal neutron content (TNC) = Thermal neutron fluence Total neutron fluence 100 (6.7)

9 Chapter 6. Optimization of thermal neutron source In similar fashion ENC and FNC are calculated. Moreover, the factor that weighs up both the (N/N 0 ) and mean neutron energy defined as (N/N 0.E mean ) is also calculated [17]. The material for which the factor is highest, is found to be the best material for the second region. Once the second region composition and dimensions are optimized, third region is added in the geometry and respective neutron energy spectrum calculations are made in perpendicular direction to the incident photon beam. Bremsstrahlung fluence is maximum in forward direction and decreases sharply with angle [18], therefore, to minimize the gamma background the neutron beam is brought out perpendicular to the incident photon beam. The function of third region is to increase neutron fluence at the output window due to reflecting material. The material for this region should have high scattering cross section and low absorption cross section. The neutron beam brought out in perpendicular direction to the incident photon beam, is moderated in fourth region. The objective of the fourth region is to shift the energies of neutron to thermal energies. For this purpose low Z elements in the periodic system are tested. In the moderating material neutron looses energy until they reach an equivalent temperature equal to the environment. The thermal neutron fluence, its uniformity and neutron to gamma ratio are calculated at the exit window. The material and dimension of the fourth region changes until the neutron uniformity at the output window is greater than 90%. Once the total design is optimized, the shielding of source has been optimized for neutrons and gamma radiation. 6.5 Optimization of targets Region 1 (γ n target) Based on photo nuclear reaction threshold beryllium (Be), beryllium oxide (BeO), beryllium deuteride (BeD 2 ) and combination of Be and BeD 2 were

10 Chapter 6. Optimization of thermal neutron source simulated in FLUKA for the first region. Figure 6.2(a) and 6.2(b) shows the neutron yield and fluence as a function of thickness of cylinder for materials simulated for the first region. From Figure 6.2(a) it is observed that for all the materi- Neutron Yield (neutron/electron) 5e-06 4e-06 3e-06 2e-06 1e-06 Beryllium Beryllium Deuteride Beryllium Oxide Combine Be and BeD γ-n target thickness (cm) (a) Neutron Fluence ((neutron-cm -2 )/e _ ) 4e-08 3e-08 2e-08 1e-08 Beryllium Beryllium Deuteride Beryllium Oxide Combine Be and BeD γ-n target thickness (cm) (b) Figure 6.2: Variation in neutron yield and fluence as a function of target thicknesses for different materials.

11 Chapter 6. Optimization of thermal neutron source als as thickness increases, the neutron yield increases and beryllium found to be the highest neutron yield material as compared to other materials. Therefore, it was decided to use beryllium as a (γ, n) target for the first region. Figure 6.2(b) shows that for beryllium the neutron fluence increases till the thickness of 4 cm and further decreases with the increase in thickness because of the absorption of neutron in the material itself. Therefore, the thickness of the beryllium was taken 4 cm for the first region. The one more advantage of choosing beryllium for the first region is that it quickly (10 16 s) decays into stable He 4 atoms [19]. The neutron fluence, neutron yield (N 0 ) and mean energy of the neutron estimated in FLUKA for 4 cm thick beryllium cylinder is neutron cm 2 /e, neutron/e and 286 kev respectively. The neutron calculated in forward and orthogonal direction for 4 cm thick beryllium target is shown in Figure 6.3. It has been observed from the neutron energy spectra of beryllium that more than 85% of the neutrons has energy > 100 kev (i.e. fast neutrons). Neutron Fluence ((neutron-mev -1 -cm -2) /e _ ) 1e-07 1e-08 Forward direction Orthogonal direction Addition of forward and orthogonal 1e Neutron Energy (MeV) Figure 6.3: Neutron spectra calculated in forward and orthogonal direction from beryllium target.

12 Chapter 6. Optimization of thermal neutron source Region 2 (Filter) To increase the number of neutrons in region 2, the (n,2n) reaction threshold was checked for all stable elements. Out of these elements only beryllium and deuterium found to have threshold below 6 MeV and their threshold energies are and 3.33 MeV respectively. Neutron spectra for first region gives 3% and 1.5% of neutrons having energy more than 1.85 MeV and 3.33 MeV respectively. Therefore, the possibility of increasing neutrons is less through (n, 2n) reaction. To shift the neutron energy spectra, materials such as Beryllium (Be), Aluminum (Al), Alumina (Al 2 O 3 ), Uranium (U), Heavy water (D 2 O), Polyethylene (Pl) ((CH 2 ) n ) and Graphite (C) were simulated with different thicknesses as a second region. Figure 6.4 shows the total neutron fluence and thermal neutron fluence (E < 0.3eV) with filter thickness for different materials. Only materials such as polyethylene and beryllium are giving thermal neutrons and respective results are shown in Figure 6.4. For all the materials, total neutron fluence decreases with increase in thickness of the material because of absorption of neutrons in material. The absorption purely depends on the Z of Total Neutron Fluence ((neutron-cm -2 )/e _ ) 3.5e-08 3e e-08 2e e-08 1e-08 5e-09 Al(tot) D2O(tot) Gr(tot) Pl(tot) U(tot) Be(tot) Pl(ther) Be(ther) 1e-09 9e-10 8e-10 7e-10 6e-10 5e-10 4e-10 3e-10 2e-10 1e-10 Thermal Neutron Fluence ((neutron-cm -2 )/e _ ) Filter thickness (cm) Figure 6.4: Variation in total neutron fluence and thermal neutron fluence as a function of filter thickness for different materials.

13 Chapter 6. Optimization of thermal neutron source the material which varies for all the materials. It is observed from Figure 6.4 that for polyethylene neutron loss is slightly higher in comparison with other materials. Epi-thermal neutron content (%) Al D2O Gr Pl U Be Filter thickness (cm) (a) Figure 6.5: The variation in Epithermal Neutron Content (ENC and FNC) as a function of filter thicknesses for different materials. Moreover, polyethylene has advantage that it transfers more neutrons to the thermal energy range. The ENC and FNC is calculated for each case of thickness and materials. The variation in ENC and FNC with filter thickness for different materials are shown in Figure 6.5(a) and 6.5(b) respectively. It is observed from Figure 6.5(a) that the ENC found to be higher for polyethylene as compared to other materials, while the FNC found to be less for polyethylene as shown in Figure 6.5(b). It is observed from Figure 6.4 that the total neutron fluence for beryllium is 2 times higher than that of polyethylene, but these neutrons mostly contains the fast neutrons as seen from Figure 6.5(b). This results implies that polyethylene can act as a good filter material. For the confirmation, variation in factor (N/N 0.E mean ) with filter thickness is shown in Figure 6.6. N 0 is the neutron yield incidence on region 2, N is the neutron yield and E mean is the mean energy of neutrons coming out of the filter. This factor weighs up both N/N 0 ratio

14 Chapter 6. Optimization of thermal neutron source Fast neutron content (%) Al D2O Gr Pl U Be Filter thickness (cm) (b) Figure 6.5: The variation in Fast Neutron Content (ENC and FNC) as a function of filter thicknesses for different materials N/(N 0.E mean ) (ev -1 ) Al D2O Gr Pl U Be 1e-05 1e Filter thickness (cm) Figure 6.6: Variation in fraction of N/(N 0.E mean ) with filter thickness for different material.

15 Chapter 6. Optimization of thermal neutron source and mean neutron energy. If higher the magnitude of this factor, better the performance of the material. It is observed from the Figure 6.6 that the factor is higher for polyethylene as compared to other studied materials. The factor is increasing with thickness and saturates beyond 4 cm thickness of polyethylene, however, it is also observed in Figure 6.4 that the thermal neutron fluence is maximum for 4 cm thickness of polyethylene. Therefore, polyethylene of 4 cm thickness was optimized for second region. The total neutron fluence, thermal neutron fluence are neutron cm 2 /e, neutron cm 2 /e and the FNC, ENC and TNC for optimized target are 35.21%, 47.05% and 17.73% respectively Prototype experiment At this stage of the design of pulsed thermal neutron source, it was very important to compare the simulated results with some experimental results to confirm that results obtained so far are correct and following proper direction. An experimental setup of the prototype experiment in the present case is shown in Figure 6.7. Paraffin wax which seem to be an equivalent to the polyethylene material with respect to the neutron properties, was used as a moderating material for the measurement of thermal neutron flux. The thermal neutron flux was measured by the activation of Vanadium (V 51 ) with the following reaction n + 51 V 52 V + γ E γ = 1.43MeV, T 1/2 = min The LINAC was operated on photon mode with an initial electron beam parameters of energy 6 MeV, repetition rate 150 pps, pulse width 4.5 µsec, pulsed current 130 ma, average current 80 µa and tungsten was used as a an electron to gamma converter target having radius 0.3 cm and thickness of 0.22 cm. The bremsstrahlung radiations emitted from the e γ target were made to fall on the cylindrical beryllium target having thickness 4 cm, to generates neutrons through photo nuclear reaction (γ, n). In order to reduce the energy of fast neutrons, beryllium was covered with paraffin wax from all the sides. For the measurement of

16 Chapter 6. Optimization of thermal neutron source Iron Lead Wax Electron Collimator Bremsstrahlung Radiation Beryllium Thickness variation Neutron e- target Vanadium Figure 6.7: Experimental Setup for the measurement of thermal neutron flux. total and thermal neutron flux, vanadium and cadmium covered vanadium sample was mounted in the forward direction and irradiated for 15 minutes consecutively. Immediately after irradiation, the induced gamma activity was measured using HPGe detector for 10 minutes. Using this gamma activity, the neutron flux was calculated by the activation relation [20], which can be written as σφ = A λ β Nɛ (1 e λt 1 )e λt 2(1 e λt 3) (6.8) where φ is the incident neutron flux, σ is the cross section for (n, γ) reaction, A is the gamma activity i.e total number of counts, λ is the decay constant, β is the number of gamma quanta/disintegration, N is the number of atoms in the target; ɛ is the efficiency of the detector, t 1 is the irradiation time, t 2 is the cooling time i.e the time between end of irradiation and start of counting, t 3 is the counting time. This relation is written specifically for continuous energy spectra of neutrons. The procedure adopted for calculating (φ experimental ) using (σφ) experimental and (σφ) simulated is discuss in Chapter 5. Following the same procedure, the experiment was repeated for three

17 Chapter 6. Optimization of thermal neutron source Table 6.2: Simulated and experimental total and thermal neutron flux at different thickness of moderating material. Wax Simulated Neutron flux Experimental Neutron flux Percentage Thickness Total Thermal statistical Total Thermal quadrature of Thermal (cm) φ (n/cm 2 sec) error (%) φ (n/cm 2 sec) error (%) neutron (%) ± ± sets of samples. In this manner, the total and thermal neutron flux was measured at different paraffin thicknesses of 0 cm, 4 cm, 8 cm, 12 cm and 16 cm. In all the repeated experiment, the total neutron flux and thermal neutron flux was measured from gamma activity. The same setup as in the experimental condition was modeled in FLUKA for simulating the results at various paraffin thickness for the measurement of total and thermal neutron fluence and subsequently compared with experimental results. The experimental and simulated results of thermal and total neutron flux at different thickness of wax is shown Table 6.2. It is observed from the Table 6.2 that in both the cases total neutron flux is decreasing with increasing thickness of paraffin, while thermal neutron flux increases up to 4 cm thickness and further decreases with increase in the thickness up to 16 cm. However, overall the percentage contribution of thermal neutron (TNC) found to be increased with thickness of moderating material. The experimental errors were evaluated in quadrature and was found to be around 7% to 9%. It is clear from table that the experimental values are found to be in good agreement with the simulated values by FLUKA Region 3 (Reflector) Next step in the design of pulsed thermal neutron source is to optimize the material and dimensions for region 3. The purpose of the region 3 is to transfer more and more number of thermal neutrons to the output direction. The

18 Chapter 6. Optimization of thermal neutron source bremsstrahlung fluence decreases sharply with angle and it is found to be highest in the forward direction (0 ). Therefore, to reduce the gamma contamination, in thermal neutron beam, it was decided to consider neutron output in perpendicular direction (90 ) to the incident beam. The region 3 is positioned such that it surrounds the optimized geometry of region 1 and 2 with small opening for neutron output in perpendicular direction (90 ) to incident beam. The materials such as alumina, graphite, beryllium, lead and polyethylene were tested for region 3 with varying thicknesses. The effect of adding region 3 on neutron fluence and mean energy with thickness is shown in Figure 6.8(a) and 6.8(b). It is observed from Figure 6.8(a) that the neutron fluence from beryllium is almost 1.5 to 2 times higher than without reflector because it can serve as an additional booster for generating neutrons through (γ, n) reaction. The neutron fluence increases with reflector thickness and for beryllium it saturates beyond 6 cm thickness. Whereas, Figure 6.8(b) shows the mean energy of neutron which found to be decreased with increasing reflector thickness. The mean energy for polyethylene and beryllium less than 0.8 ev and which found to be lower than other materials. The variation in TNC and FNC as a function of reflector thickness is shown in Figure 6.9(a) and 6.9(b) respectively. It is seen from Figure 6.9(a) that the TNC increases with reflector thickness and get saturates beyond 8 cm, while FNC decreases with increase in reflector thickness as shown in Figure 6.9(b). But the percentage change in the fast neutron is less within 5% range for all studied materials, whereas, if reflector material changed from beryllium to polyethylene of same thickness, the percentage change in fast neutron is 1%. It is also observed that the beryllium provides 1.4 times more neutron fluence and less TNC as compared to polyethylene. Whereas, FNC and mean neutron energy remains the same for both the materials. It is therefore apparent advantage of optimizing 6 cm of beryllium surrounded with 10 cm of polyethylene. The beryllium in this case acts as a reflector and polyethylene acts as a moderator. In general, the effect of adding beryllium and polyethylene in region 3, was found to increase neutron fluence

19 Chapter 6. Optimization of thermal neutron source Neutron Fluence ((neutron-cm -2 )/e _ ) 1.2e e-08 1e-08 9e-09 8e-09 7e-09 Al2O3 Be Gr Pb Pl 6e Reflector thickness (cm) (a) Mean Neutron Energy (ev) Al2O3 Be Gr Pb Pl Reflector thickness (cm) (b) Figure 6.8: Variation in neutron fluence and mean neutron energy as a function of reflector thickness for different materials. by 60% because of the reflection of neutrons and neutron generated in beryllium. Moreover, the neutrons other than output direction also get thermalize in

20 Chapter 6. Optimization of thermal neutron source Thermal neutron content (%) Al2O3 Be Gr Pb Pl Reflector thickness (cm) (a) Al2O3 Be Gr Pb Pl Fast neutron content (%) Reflector thickness (cm) (b) Figure 6.9: Variation in TNC and FNC as a function of reflector thickness for different material. polyethylene such that the shielding can be made very easily. To shield the neutrons, polyethylene of thickness 30 cm was covered in all the direction except output canal. The produced thermal neutrons get absorbed by cadmium as it

21 Chapter 6. Optimization of thermal neutron source has very high absorption cross section with thermal neutrons. The cadmium of thickness 0.5 mm was used to absorbs the thermal neutrons Region 4 (Moderating Column) The neutron beam extracted at the perpendicular direction with respect to the photon beam is then moderated such that less neutron loss and more scattering occurs to shift the energy. This region mainly has an objective to shift the energy spectrum to thermal energy range and guide the uniform neutron at output canal. Materials tested for region 4 are alumina, polyethylene, graphite which mainly belongs to low Z elements of the periodic table. Results obtained using these materials are given in Figure 6.10 with varying the thickness of moderating column. It is found that the TNC increases with thickness. For polyethylene, Al2O3 Gr Pl Thermal neutron content(%) Moderating column thickness (cm) Figure 6.10: Variation in TNC as a function of moderator thickness for different material. the TNC increases fast with thickness as compared to other materials. Therefore, 14 cm thick polyethylene was optimized for region 4. The neutron beam profile was estimated at output for 1 1 mm bin and uniformity was measured. To obtain uniform beam, graphite was used in the thermal column. The dimension of

22 Chapter 6. Optimization of thermal neutron source graphite was adjusted until the beam uniformity found to be greater than 90%. The optimized design of the accelerator based thermal neutron source is shown in Figure For this optimized design, the neutron fluence obtained is around neutron cm 2 sec 1 with more than 80% of thermal neutrons and an acceptable neutron to gamma ratio is neutron cm 2 mr 1. The neutron spectra calculated on the exit plane of the source is shown in Figure Cadmium Polyethylene Polyethylene as Shielding Region 3 Iron Lead Beryllium Region 3 Electron e- target Collimator Bremsstrahlung Radiation Polyethylene Beryllium Region 1 Region 2 L E A D Graphite Moderating column Region 4 Neutron Output window Figure 6.11: Schematic diagram of the optimized accelerator based pulsed thermal neutron source (Not to the scale). 6.6 Conclusion In conclusion, a successful study has been carried out for the design of 6 MeV electron accelerator based pulsed thermal neutron source with the tungsten as e γ converter, beryllium as γ n converter in region 1, polyethylene as a filter in region 2, beryllium as reflector in region 3, polyethylene covered with

23 Chapter 6. Optimization of thermal neutron source At exit window Neutron Fluence ((neutron-mev -1 -cm -2) /e _ ) Neutron Energy (MeV) Figure 6.12: Neutron spectra calculated at exit plane of the 6 MeV Linear accelerator based thermal neutron source. cadmium as a neutron shield and graphite + polyethylene as a moderating column in region 4. The neuron fluence calculated for the optimized design is around neutron cm 2 sec 1 with an acceptable neutron to gamma ratio is neutron cm 2 mr 1. The design of this neutron source is therefore used for various applications such as neutron scattering, diffraction and to see structures of the size of magnetic microstructures and DNA molecules. Moreover, the measurement of neutron flux of prototype accelerator based pulsed neutron source for different thickness of wax as a moderator was carried out and respective experimental results show good agreement with the simulated results by FLUKA. 6.7 Future Scope An important and growing market for neutron generators is in analyzing bulk materials. Taking advantage of recently developed pulsed thermal neutron source can be used for the real-time analysis of materials such as cement and coal moving on conveyor belts. This source can be run on both fast and

24 Chapter 6. Optimization of thermal neutron source thermal-neutron activation analysis to measure the elemental content of the major constituents in the bulk material and use stoichiometric relationships to convert the elemental information to chemical assays. In the cement analysis, this information enables the optimal blending of raw materials before processing and the verification of chemical uniformity of the final product. In the coal analysis, on-line measurements have found particular use in reporting the thermal energy and sulfur content of coal and for determining the fraction of the coal that is not hydrocarbon and will remain as ash after combustion. Overall, this system has wide scope in the industrial applications.

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