optimization of thermal neutron source based on 6 mev...

Chapter 6.

Optimization of thermal neutron source basedon 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 energymoderating 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

th l T if th i l t d ltthermal 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.g g

138

Chapter 6. Optimization of thermal neutron source .... 139

6.1 Importance and Objective

Neutrons are used in various applications but mainly neutron diffrac-

tion and scattering provides valuable tool for probing the structure of bulk mate-

rials. 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 ef-

fect 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 microstruc-

tures and DNA molecules. The wavelength of fast neutron is too short for inves-

tigating 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 mag-

netic domains or hydrogen atoms in long chain nucleosomes.

Neutron facilities throughout the world generate neutrons by using nu-

clear reactors, radioisotopes and high energy particle accelerators as a primary

source. The nuclear reactors are the highest neutron yield source, but size, com-

plexity 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


flux and can be utilized for very specific applications. However, the particle ac-

celerator based neutron sources are vary in size and diversity. Because of the

compactness, easy handling, adjusted flux with the beam parameters, no radioac-

tive 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 vari-

ous bulk materials. When electron beam from an accelerator incident on high Z

(e− γ ) target it generates a cascade shower of bremsstrahlung radiations. Fur-

ther, 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/r2). Thus, when designing of var-

ious 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.


Tabl

e6.

1:R

evie

wof

The

rmal

Neu

tron

prod

uctio

n.

No.

Aut

hor

(J.n

ame,

(yea

r),

Sim

ulat

ion/

Type

Res

ult

Vol,p

p)[r

ef]

Exp

erim

ent

1Pi

cton

D.J

.J.

Phys

.D:A

pplP

hys.

(198

2)V

ario

usm

ater

ials

have

15,2

369-

2400

[4]

been

test

edfo

rmod

erat

or2

Dan

onY.

NIM

A(1

995)

MC

NP

sim

ulat

ion

ford

esig

ning

60M

eVe−

onTa

targ

etm

easu

red

neut

ron

flux

foro

ldan

dne

w35

2,59

6-60

3[5

]en

hanc

edth

erm

alne

utro

nta

rget

targ

etge

omet

ryof

cold

mod

erat

orde

sign

3A

gost

eoS.

NIM

A(2

002)

MC

NP

sim

ulat

ion,

mea

sure

men

t7

MeV

Deu

tero

non

ther

mal

and

epith

erm

alne

utro

nflu

x47

6,10

6112

[6]

thro

ugh

activ

atio

nte

ch.

Ber

ylliu

mta

rget

mea

sure

d.D

esig

ned

forB

NC

T4

Aud

itore

L.

NIM

B,(2

005)

,M

CN

Psi

mul

atio

n5

MeV

Lin

acba

sed

Neu

tron

fluen

ceca

lcul

ated

is8.

5×1

07n/

s/cm

2 /mA

229,

137

[7]

from

Be

and

1.5

times

from

BeD

2

5B

arta

lucc

iSer

gio

RE

POR

TN

o.19

61/P

N,(2

005)

MC

NP

and

FLU

KA

1G

eVe−

Lin

acon

ther

mal

neut

ron

ener

gysp

ectr

aIn

st.N

ucl.

Phys

.[8]

sim

ulat

ion

tant

alum

targ

etca

lcul

ated

and

angu

lard

istr

ibut

ion

stud

ied

6B

arta

lucc

iS.

NIM

A(2

007)

MC

NP

sim

ulat

ion

ford

esig

n50

0M

eVL

inac

base

dM

oder

ator

desi

gnfo

rtim

e57

5,28

7-29

1[9

]of

fligh

tmea

sure

men

texp

erim

ent

7Fa

valli

A.

Rad

.Pro

t.D

osm

.(20

07)

Cha

ract

eriz

atio

nof

D–T

reac

tion

base

dT

herm

alne

utro

nfa

cilit

yha

sbe

ente

sted

126,

74,[

10]

The

rmal

neut

ron

faci

lity

ther

mal

neut

ron

8G

rzeg

orz

T.A

ppl.R

ad.Is

o.,(2

009)

,M

CN

Psi

mul

atio

nD

-Tre

actio

nba

sed

puls

edT

herm

alne

utro

nflu

xca

lcul

ated

and

mea

sure

d.67

,114

8[1

1]th

erm

alne

utro

nso

urce


6.3 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.

6.3.1 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 accel-

erator 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 =12

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=

A2 + 1 + 2Acosθ(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,


Equation 6.2 becomes

[E′

E

]θ= 1800

=

[A − 1A + 1

]2

(6.3)

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 slow-

ing 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 scatter-

ing angle from neutrons below 10 MeV. The distribution of energy transfer E′/E

for one collision is uniform over the range (E′/E)θ=1800 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

EE′

]avg

=

∫ln

[(A+1)2

A2+1+2Acosθ

]dΩ∫

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

2Aln

[A − 1A + 1

](6.5)

The moderating material should have a considerable scattering cross section, (Σs)

and less cross section for absorption (Σa), 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


ratio, Rm, by means of the expression:

Rm = ξΣs

Σa(6.6)

The moderating ratio, Rm, 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 pri-

mary collimator is estimated using FLUKA. The bremsstrahlung spectrum is

shown in Figure 6.1. The integrated bremsstrahlung fluence is 2.792 ×10−3

(photon−cm−2)/e−.

At first electron source along with electron to gamma converter, pri-

mary 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 2.226 MeV and 1.666 MeV

respectively [14]. The cross section of (γ, n) reaction with beryllium and deu-

terium 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


0.0e+0

5.0e-4

1.0e-3

1.5e-3

2.0e-3

2.5e-3

3.0e-3

0 1 2 3 4 5 6 7

Bre

mss

trah

lung F

luen

ce (

(photo

n-M

eV-1

-cm

-2) /e

_)

Bremsstrahlung Energy (MeV)

Calculated on collimator exit

Figure 6.1: Bremsstrahlung spectrum for a 0.22 cm thick tungsten target, calculated atprimary collimator exit face.

is strictly limited to few light elements such as deuterium and beryllium for neu-

tron 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).r2) 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)


In similar fashion ENC and FNC are calculated. Moreover, the factor that weighs

up both the (N/N0) and mean neutron energy defined as (N/N0.Emean) is also cal-

culated [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 ma-

terial. 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 ele-

ments 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

6.5.1 Region 1 (γ− n target)

Based on photo nuclear reaction threshold beryllium (Be), beryllium

oxide (BeO), beryllium deuteride (BeD2) and combination of Be and BeD2 were


simulated in FLUKA for the first region. Figure 6.2(a) and 6.2(b) shows the neu-

tron yield and fluence as a function of thickness of cylinder for materials simu-

lated for the first region. From Figure 6.2(a) it is observed that for all the materi-

0

1e-06

2e-06

3e-06

4e-06

5e-06

0 2 4 6 8 10 12 14 16 18 20

Neu

tron Y

ield

(neu

tron/e

lect

ron)

γ-n target thickness (cm)

Beryllium

Beryllium DeuterideBeryllium Oxide

Combine Be and BeD2

(a)

0

1e-08

2e-08

3e-08

4e-08

0 2 4 6 8 10 12 14 16 18 20

Neu

tron F

luen

ce (

(neu

tron-c

m-2

)/e_

)

γ-n target thickness (cm)

Beryllium

Beryllium DeuterideBeryllium Oxide

Combine Be and BeD2

(b)

Figure 6.2: Variation in neutron yield and fluence as a function of target thicknessesfor different materials.


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 (10−16s) decays into stable He4 atoms [19]. The

neutron fluence, neutron yield (N0) and mean energy of the neutron estimated

in FLUKA for 4 cm thick beryllium cylinder is 3.978 ×10−8 neutron−cm−2/e−,

2.133 ×10−6 neutron/e− and 286 keV respectively. The neutron calculated in

forward and orthogonal direction for 4 cm thick beryllium target is shown in Fig-

ure 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).

1e-09

1e-08

1e-07

0.001 0.01 0.1 1 10

Neu

tron F

luen

ce (

(neu

tron-M

eV-1

-cm

-2) /e

_)

Neutron Energy (MeV)

Forward direction

Orthogonal directionAddition of forward and orthogonal

Figure 6.3: Neutron spectra calculated in forward and orthogonal direction from beryl-lium target.


6.5.2 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 beryl-

lium and deuterium found to have threshold below 6 MeV and their threshold

energies are 1.851 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 (Al2O3), Uranium (U), Heavy water

(D2O), Polyethylene (Pl) ((CH2)n) and Graphite (C) were simulated with differ-

ent 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 ma-

terials. 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

0

5e-09

1e-08

1.5e-08

2e-08

2.5e-08

3e-08

3.5e-08

2 3 4 5 6 7 8 0

1e-10

2e-10

3e-10

4e-10

5e-10

6e-10

7e-10

8e-10

9e-10

1e-09

Tota

l N

eutr

on F

luen

ce (

(neu

tron-c

m-2

)/e_

)

Ther

mal

Neu

tron F

luen

ce (

(neu

tron-c

m-2

)/e_

)

Filter thickness (cm)

Al(tot)

D2O(tot)Gr(tot)Pl(tot)U(tot)

Be(tot)Pl(ther)

Be(ther)

Figure 6.4: Variation in total neutron fluence and thermal neutron fluence as a functionof filter thickness for different materials.


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 materi-

als.

10

20

30

40

50

60

70

80

90

2 3 4 5 6 7 8

Epi-

ther

mal

neu

tron c

onte

nt

(%)


Al

D2OGrPlU

Be

(a)

Figure 6.5: The variation in Epithermal Neutron Content (ENC and FNC) as a functionof 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 ob-

served 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 flu-

ence 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, varia-

tion in factor (N/N0.Emean) with filter thickness is shown in Figure 6.6. N0 is the

neutron yield incidence on region 2, N is the neutron yield and Emean is the mean

energy of neutrons coming out of the filter. This factor weighs up both N/N0 ratio


10

20

30

40

50

60

70

80

90

2 3 4 5 6 7 8

Fas

t neu

tron c

onte

nt

(%)


AlD2O

GrPlU

Be

(b)

Figure 6.5: The variation in Fast Neutron Content (ENC and FNC) as a function offilter thicknesses for different materials.

1e-06

1e-05

0.0001

0.001

0.01

0.1

1

10

2 3 4 5 6 7 8

N/(

N0.E

mea

n)

(eV

-1)


AlD2O

GrPlU

Be

Figure 6.6: Variation in fraction of N/(N0.Emean) with filter thickness for differentmaterial.


and mean neutron energy. If higher the magnitude of this factor, better the perfor-

mance 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 5.185 ×10−9 neutron−cm−2/e−, 9.194 ×10−10 neutron−cm−2/e− and

the FNC, ENC and TNC for optimized target are 35.21% , 47.05% and 17.73%

respectively.

6.5.3 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 experimen-

tal 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 mea-

surement of thermal neutron flux. The thermal neutron flux was measured by the

activation of Vanadium (V51) with the following reaction

n +51 V → 52V + γ Eγ = 1.43MeV, T1/2 = 3.743 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, beryl-

lium was covered with paraffin wax from all the sides. For the measurement of


e- target

Wax

Beryllium

Collimator

Electron

Iron

Lead

Thickness

variation

Vanadium

Bremsstrahlung

Radiation

Neutron

Figure 6.7: Experimental Setup for the measurement of thermal neutron flux.

total and thermal neutron flux, vanadium and cadmium covered vanadium sam-

ple was mounted in the forward direction and irradiated for 15 minutes consecu-

tively. 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

βNε (1 − e−λt1)e−λt2(1 − e−λt3)(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;

ε is the efficiency of the detector, t1 is the irradiation time, t2 is the cooling time

i.e the time between end of irradiation and start of counting, t3 is the counting

time. This relation is written specifically for continuous energy spectra of neu-

trons. The procedure adopted for calculating (φexperimental) using (σφ)experimental

and (σφ)simulated is discuss in Chapter 5.

Following the same procedure, the experiment was repeated for three


Table 6.2: Simulated and experimental total and thermal neutron flux at different thick-ness of moderating material.

Wax Simulated Neutron flux Experimental Neutron flux PercentageThickness Total Thermal statistical Total Thermal quadrature of Thermal

(cm) φ (n/cm2 − sec) error (%) φ (n/cm2 − sec) error (%) neutron (%)

×106 ×105 ± ×106 ×105 ±

0 1.861 3.739 1.26 1.761 3.609 6.65 20.484 1.152 5.755 2.77 1.073 5.625 7.03 52.428 0.226 1.524 3.54 0.219 1.404 7.32 63.96

12 0.0435 0.328 4.05 0.0439 0.330 7.78 75.2916 0.0103 0.0855 4.95 0.0099 0.0805 8.01 81.06

sets of samples. In this manner, the total and thermal neutron flux was mea-

sured 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 com-

pared 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, over-

all 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.

6.5.4 Region 3 (Reflector)

Next step in the design of pulsed thermal neutron source is to opti-

mize 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


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 sur-

rounds 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 with-

out 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 reflec-

tor 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 re-

flector 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 polyethy-

lene. 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


6e-09

7e-09

8e-09

9e-09

1e-08

1.1e-08

1.2e-08

0 2 4 6 8 10 12

Neu

tron F

luen

ce (

(neu

tron-c

m-2

)/e_

)

Reflector thickness (cm)

Al2O3

BeGrPbPl

(a)

0.6

0.65

0.7

0.75

0.8

0.85

0.9

0.95

1

1.05

1.1

0 2 4 6 8 10 12

Mea

n N

eutr

on E

ner

gy (

eV)


Al2O3

BeGrPbPl

(b)

Figure 6.8: Variation in neutron fluence and mean neutron energy as a function ofreflector thickness for different materials.

by ∼ 60% because of the reflection of neutrons and neutron generated in beryl-

lium. Moreover, the neutrons other than output direction also get thermalize in


22

23

24

25

26

27

28

29

30

31

0 2 4 6 8 10 12

Ther

mal

neu

tron c

onte

nt

(%)


Al2O3

BeGrPbPl

(a)

30

30.5

31

31.5

32

32.5

33

33.5

34

34.5

35

0 2 4 6 8 10 12

Fas

t neu

tron c

onte

nt

(%)


Al2O3

BeGrPbPl

(b)

Figure 6.9: Variation in TNC and FNC as a function of reflector thickness for differentmaterial.

polyethylene such that the shielding can be made very easily. To shield the neu-

trons, polyethylene of thickness 30 cm was covered in all the direction except

output canal. The produced thermal neutrons get absorbed by cadmium as it


has very high absorption cross section with thermal neutrons. The cadmium of

thickness 0.5 mm was used to absorbs the thermal neutrons.

6.5.5 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 scat-

tering 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 out-

put 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,

30

40

50

60

70

80

90

0 2 4 6 8 10 12 14 16 18 20

Ther

mal

neu

tron c

onte

nt(

%)

Moderating column thickness (cm)

Al2O3

GrPl

Figure 6.10: Variation in TNC as a function of moderator thickness for different ma-terial.

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 ob-

tain uniform beam, graphite was used in the thermal column. The dimension of


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 6.11. For this optimized design, the neutron fluence obtained

is around 3 ×106 neutron−cm−2−sec−1 with more than 80% of thermal neutrons

and an acceptable neutron to gamma ratio is 1 ×104 neutron−cm−2−mR−1. The

neutron spectra calculated on the exit plane of the source is shown in Figure 6.12.

Polyethylene Polyethylene as Shielding

Region 3

Moderating column

Region 4

Beryllium

Region 3

e- target

Polyethylene

Region 2

Beryllium

Region 1

Collimator

Electron

Iron

Lead

Graphite

Cadmium

Neutron

Output window

L

E

A

D

Bremsstrahlung

Radiation

Figure 6.11: Schematic diagram of the optimized accelerator based pulsed thermalneutron 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 tung-

sten 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


10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

Neu

tron F

luen

ce (

(neu

tron-M

eV-1

-cm

-2) /e

_)

Neutron Energy (MeV)

At exit window

Figure 6.12: Neutron spectra calculated at exit plane of the 6 MeV Linear acceleratorbased 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 3

×106 neutron−cm−2−sec−1 with an acceptable neutron to gamma ratio is 1 ×104

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 mea-

surement of neutron flux of prototype accelerator based pulsed neutron source

for different thickness of wax as a moderator was carried out and respective ex-

perimental results show good agreement with the simulated results by FLUKA.

6.7 Future Scope

An important and growing market for neutron generators is in analyz-

ing bulk materials. Taking advantage of recently developed pulsed thermal neu-

tron 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


thermal-neutron activation analysis to measure the elemental content of the ma-

jor constituents in the bulk material and use stoichiometric relationships to con-

vert 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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