I/I
In vivo Nitrogen Measurement
by use of (n,2n) method and PGNAA
by
Aiko Ishikawa
B.S., Applied Physics (1996)
Waseda University, Tokyo
Submitted to the Department of Nuclear Engineering
in partial fulfillment of the requirements for the degrees of
Master of Science in Nuclear Engineering
at the
MASSACHUSETTS INSTITUE OF TECHNOLOGY
June 1998
© 1998 Massachusetts Institute of Technology. All rights reserved.
Signature of author:
Nuclear Engineering Department
May 15, 1998
Certified by :
gro/ Jacqlyn C. Yanch, Thesis Supervisor
ofessor of Nuclear Engineering Department
N
1
Dr. Joseph J. Kehayias, Tesis Reaer
USDA Human Nutrition R search Center on Aging at Tufts University
Accepted by : _
----rf.fi rence M. Lidsky
Chairman, Department Comittiee on Graduate Students
Chairman,
Department
Scier",.
-
OlGe
In vivo Nitrogen Measurement
by use of (n,2n) method and PGNAA
by
Aiko Ishikawa
Submitted to the Department of Nuclear Engineering on May 15, 1998
in partial fulfillment of the requirements for the degrees of
Master of Science in Nuclear Engineering
Abstract. The technique of in vivo neutron activation analysis has been used for the
determination of various elements in the human body. There are two methods of
measuring nitrogen in vivo: (n,2n) method and prompt gamma neutron activation analysis
(PGNAA). The first method utilizes the fast neutron reaction '4 N(n,2n) 13N and the
second uses the thermal reaction 14N(n,y)1 N. Both methods are currently in use at
several different facilities for evaluation of the nutritional status of surgical patients or
metabolic study of normal human subjects. In order for the absolute determination of
nitrogen, it is necessary to obtain the uniform sensitivity of detection in the subject.
Experimental factors which influence the sensitivity are the activating neutron flux and
the absolute detector efficiency. The distribution of the two factors in the human body is
dependent on the spectral characteristics of the neutron source, the geometries of
irradiation and counting systems and the size of the subject. In this thesis investigations
of sensitivity and absorbed dose of the two nitrogen methods are made based on the
results of Monte Carlo simulation. Two experimental facilities utilizing the (n,2n)
method and the PGNAA method are simulated to assess the suitability and the feasibility
of in vivo measurement of nitrogen. The potential of the use of an electrostatic tandem
accelerator as a neutron source for nitrogen measurement is examined by simulating
hypothetical experimental set-ups. Furthermore, those results obtained by the simulation
are verified by performing accuracy measurements by the (n,2n) method. The counting
system designed for human thigh measurement is described. The (n,2n) method delivers
about half of the equivalent dose of the PGNAA method with same irradiation time. The
simulation results showed 10 - 20 times higher sensitivity for the (n,2n) method than the
PGNAA technique, depending on the patient position for the irradiation. The simulation
results for the accelerator sources presented higher sensitivity than the 238Pu-Be
radionuclide source conventionally used.
Thesis Supervisor: Jacquelyn C. Yanch
Title: Professor of Nuclear Engineering Department
Acknowledgments
I would like to thank Professor Jacquelyn C. Yanch for her advice for my thesis
and great patience with a number of mistakes I have made. I would also like to thank Dr.
Kehayias who taught me basic concepts needed for researching. I am grateful to
members of LABA group who have been a great help throughout my research work. It
has been my great honor to work with such competent and diligent scientists.
Table of Contents
1. INTRODUCTION
2. TBN MEASUREMENT
2.1 (N,2N) METHOD
2.2 PGNAA METHOD
8
12
12
17
3. NITROGEN MEASUREMENT USING A D-T NEUTRON
25
GENERATOR
3.1 OBJECTIVE
25
3.2 IRRADIATION SYSTEM
3.3 COUNTING SYSTEM FOR HUMAN THIGH MEASUREMENT
3.4 EXPERIMENTAL PROCEDURE
26
3.4.1 NITROGEN MEASUREMENT
3.4.2 MEASUREMENT OF DETECTOR EFFICIENCY OF THIGH COUNTING SYSTEM
3.4.3 INTERFERENCE OF OXYGEN
3.5 DATA ANALYSIS AND RESULTS
3.5.1
3.5.2
3.5.3
3.5.4
COUNTING RESULTS
OXYGEN INTERFERENCE
ESTIMATION OF DETECTOR EFFICIENCY AT 511 KEV
ESTIMATION OF NEUTRON OUTPUT
4. MONTE CARLO SIMULATION
4.1 MCNP CODE
4.2 FIGURES OF MERIT
4.2.1 DOSE
4.2.2 SENSITIVITY
4.3 NITROGEN MEASUREMENT SYSTEMS
4.3.1 D-T GENERATOR - (N,2N) METHOD
4.3.2 238Pu-BE RADIONUCLIDE - PGNAA METHOD
4.3.3 ACCELERATOR SOURCE 9BE(P,N) WITH LABA SET-UP - PGNAA METHOD
4.3.4 VARIOUS ACCELERATOR SOURCES WITH BNL SET-UP - PGNAA METHOD
27
29
29
31
31
32
32
33
33
35
36
36
37
37
41
47
47
53
62
68
4.4 RESULTS
70
4.4.1 DOSE
70
4.4.2 SENSITIVITY
4.5 SUMMARY
5. COMPARISON AND CONCLUSION
86
100
101
5.1 ACCURACY AND REPRODUCIBILITY
5.2 IRRADIATION AND COUNTING TIME
101
5.3 DOSE
106
5.3.1 INCIDENT DOSE
5.3.2 DOSE DISTRIBUTION
5.4 SENSITIVITY
5.4.1 SUPINE- AND PRONE-POSITION IRRADIATION FOR (N,2N) METHOD
5.4.2 ACCELERATOR SOURCES FOR PGNAA METHOD
5.4.3 (N,2N) METHOD AND PGNAA TECHNIQUE
102
106
106
108
109
109
110
5.5 CONCLUSION
5.6 FUTURE WORK
111
112
5.6.1 (N,2N) METHOD
113
5.6.2 PGNAA METHOD
114
6. REFERENCES
115
Table of Figures
12
Figure 2-1: (n,2n) method: Gamma-ray spectrum
Figure 2-2 : PGNAA method: Gamma-ray spectrum obtained from a normal
18
human body
20
Figure 2-3 : Schematic illustration of the PGNAA facility employed at BNL
21
Figure 2-4 : The design of BOMAB phantom
26
Figure 3-1 : (n,2n) method: irradiation set-up
27
Figure 3-2 : (n,2n) method: counting set-up
28
Figure 3-3 : (n,2n) method: counting set-up with a patient
Figure 3-4 : (n,2n) method: irradiation set-up for urea solution measurement 30
Figure 3-5: Nitrogen 511 keV y-spectrum obtained from a urea solution sample32
38
Figure 4-1 : The rectangular box phantom filled with a uniform solution
40
Figure 4-2 : Thigh model for MCNP simulation
50
Figure 4-3 : (n,2n) method: BCL irradiation system with the thigh model
Figure 4-4 : (n,2n) method: BCL irradiation system with the rectangular box
phantom
Figure 4-5 : (n,2n) method: BCL counting system with the rectangular box
52
phantom
Figure 4-6 : PGNAA method: BNL irradiation system with the thigh model for
57
calculation of dose distribution by MCNP
238
58
simulation
the
for
Figure 4-7 : Pu-Be neutron yield data points used
Figure 4-8 : PGNAA method: BNL irradiation system with the rectangular box
59
phantom for sensitivity calculations
Figure 4-9 : PGNAA method: BNL detector positioning simulated by MCNP 60
Figure 4-10 : PGNAA method: one of the BNL detectors simulated by MCNP 61
Figure 4-11 : A tandem electrostatic accelerator installed in MIT's LABA is
64
shown with the ion source
Figure 4-12 : Configuration of the LABA accelerator and experimental rooms 65
Figure 4-13 : PGNAA method: LABA irradiation system with the thigh model for
66
calculation of dose distribution by MCNP
box
rectangular
the
with
system
irradiation
LABA
Figure 4-14 : PGNAA method:
67
phantom for sencitivity calculations
69
sources
accelerator
various
for
Figure 4-15 : Neutron energy spectra
72
Figure 4-16 : Axial view of the thigh modelled by MCNP
74
model
thigh
in
distribution
Figure 4-17 : (n,2n) method: Total neutron dose
Figure 4-18 : (n,2n) method: Activating neutron dose distribution in thigh model75
76
Figure 4-19 : (n,2n) method: Photon dose distribution in thigh model
Figure 4-20 : PGNAA method with 238pu-Be source: Total neutron dose
distribution in thigh model
Figure 4-21 : PGNAA method with 238Pu-Be source: Activating neutron dose
distribution in thigh model
77
78
Figure 4-22 : PGNAA method with 238Pu-Be source: Photon dose distribution in
79
thigh model
Figure 4-23 : PGNAA method with 9 Be(p,n)
neutron dose distribution in thigh model
Figure 4-24 : PGNAA method with 9 Be(p,n)
neutron dose distribution in thigh model
Figure 4-25 : PGNAA method with 9 Be(p,n)
dose distribution in thigh model
Figure 4-26 : PGNAA method with 9 Be(p,n)
neutron dose distribution in thigh model
Figure 4-27 : PGNAA method with 9 Be(p,n)
neutron dose distribution in thigh model
Figure 4-28 : PGNAA method with 9 Be(p,n)
dose distribution in thiah model
source and BNL set-ups: Total
source and
source and
source and
source and
source and
80
BNL set-ups: Activating
81
BNL set-ups: Photon
82
LABA set-ups: Total
83
LABA set-ups: Thermal
84
LABA set-ups: Photon
85
Figure 4-29 : (n,2n) method: distribution of number of 14N(n,2n)13N reactions in
88
the phantom
4-3015: PGNAA method with
Figure
14
238Pu-Be
N(ny) N reactions in the phantom
source : distribution of number of
89
9Be(p,n)
9Be(p,n)
90
source and BNL set-ups:
Figure 4-31 : PGNAA method with
set-ups:
LABA
and
source
Figure 4-32 : PGNAA method with
1
14
91
in the phantom
reactions
5N
N(n,y)
of
number
of
distribution
93
Figure 4-33 : (n,2n) method: detector efficiency distribution
94
Figure 4-34 : PGNAA method: detector efficiency distribution
Figure 4-35 : (n,2n) method: distribution of composite sensitivity in the phantom
96
for prone position irradiation
Figure 4-36 : (n,2n) method: distribution of composite sensitivity in the phantom
97
for supine position irradiation
source: distribution of composite
Figure 4-37 : PGNAA method with 238Pu-Be
98
sensitivity in the phantom
9
distribution
system:
BNL
Figure 4-38 : PGNAA method with Be(p,n) source and
99
of composite sensitivity in the phantom
Figure 5-1 : Relationship between measurement time and figures of merit:
104
nitrogen counts and sensitivity
1. Introduction
Determination of nitrogen levels in the human body provides an estimation of
protein, a highly important and essential component of body composition [1-9]. Total
body protein (TBP) has a good correlation with total body nitrogen (TBN) with the
relationship TBP = 6.25 TBN [10]. Knowledge of TBP makes it possible to evaluate
nutritional status of patients with progressive disease or metabolic disorders. Body
nitrogen measurements are used to monitor their changing status and to assess the
nutritional requirements for treatments [1, 2, 5, 9, 11-16]. Such disorders include cancer,
protein malnutrition, renal failure, anorexia, cardiovascular disease, obesity,
hyperthyroidism, liver and kidney desease, and growth deficiencies [1, 4, 16-19].
Diseases associated with muscle wasting (e.g. cystic fibrosis) can also be studied by the
quantitative estimation of nitrogen [1, 4, 8, 17, 20-22]. Furthermore, estimation of total
nitrogen allows the study of the effects of age, sex, race, and body size on body protein
levels in normal subjects [16, 23].
The technique of in vivo neutron activation analysis (IVNAA) is currently used by
an increasing number of biomedical research centers as one of the most effective methods
for quantification of various elements in the human body. There are two nuclear methods
that can be employed for determination of TBN by the use of the IVNAA technique; one
is the '4N(n,2n) 13N method [5, 11, 24-26] and the other is prompt gamma neutron
activation analysis (PGNAA)[2, 6, 8, 14, 20, 22, 27, 28]. Both methods take advantage of
the fact that when a subject is irradiated with neutrons in a certain energy range, nitrogen
in the subject will be converted into a radioactive isotope. The produced isotopes will
decay at a characteristic rate and emit y-rays which are used to identify and determine the
concentrations of nitrogen present in the subject. Fast neutrons are utilized for both
methods, although the prompt gamma reaction 14 N(n,y)15 N occurs predominantly for
neutrons of thermal energy which result from thermalization of the incident fast neutrons
by the body tissues or by a separate moderator.
The major restrictions with the IVNAA techniques for the absolute measurement
of nitrogen in vivo are nonuniformity of neutron flux through the body and the
dependence of counting efficiency on the position of nitrogen nuclei in the subject [2, 5,
7, 12, 16, 19, 22, 25, 26, 28-35]. The application of either IVNAA technique, therefore,
had been limited to sequential, relative measurements until the late 1970's, when the
PGNAA method was modified to make estimations of absolute quantities of nitrogen
mass by Vartsky et al. [2, 12]. In this new method, total body hydrogen (TBH) is also
measured as an internal standard based on the fact that the ratio of N/H counts, unlike
TBN alone, is insensitive to body size. While sequential measurements of nitrogen are
useful for surgical patients following various therapeutic protocols to monitor the change
of body nitrogen level and to evaluate their response to the surgery, the absolute
determination makes it possible to compare individual patients and thus to have an even
broader range of clinical research application than we would have only by sequential
measurements [12, 15]. An overview of TBN facilities currently in use is presented in
Table 1-1 [8, 9, 16].
Facility
Leeds,
UK
SURRC,
Scotland
Birmingham,
UK
BNL,
US
Auckland,
New Zealand
Swansea,
UK
Edinbugh,
UK
Toronto,
Canada
Sydney,
Australia
---
Neutron source
References
[11, 23, 36-38]
PGNAA
14 MeV
D-T generator
14 MeV
D-T generator
10 MeV p-Li
cyclotron
238Pu-Be
PGNAA
238Pu-Be
[2, 12, 13, 18, 35,
41, 42]
[14]
PGNAA
252Cf
[28, 34]
PGNAA
252Cf
[7, 22]
PGNAA
252Cf
[3, 6, 20, 43]
PGNAA
252Cf
[8, 44]
Method
- (n,2n) or PGNAA
(n,2n)
(n,2n)
PGNAA
[25, 26, 39]
[1, 27, 40]
Table 1-1 : Overview of TBN measurement facilities currently in use
There is another nuclear technique for TBN measurement other than the IVNAA,
which is nuclear resonance absorption (NRA) [9]. This method, based on the gamma
resonance absorption reaction 14N(y,p)13C, was first developed as a means of explosives
detection in materials transported in the airport environment [45]. In this method, 9.17
MeV y-rays are resonantly, and uniquely, absorbed by nitrogen in the patient and thus
nitrogen-specific resonant radiography can be achieved by imaging the transmission
profile of the photons having 9.17 MeV energies. The feasibility of the use of the NRA
method is currently under investigation. It is expected that this technique will offer the
advantages of low dose, good activation uniformity, and simplification of the absolute
measurement procedure.
A major objective of the work described in this thesis is to assess the suitability
and feasibility of the two IVNAA techniques: the (n,2n) method and the PGNAA method,
for TBN measurement. Monte Carlo calculation is used to simulate the two experimental
facilities which are currently in operation and to compare sensitivity and the absorbed
dose of each system. The potential of the use of an accelerator as a neutron source for
nitrogen measurement will be also investigated by simulating hypothetical set-ups.
Experimental verification is examined by performing nitrogen measurements by the
(n,2n) method and the work done to improve detector efficiency for
measurement of the human thigh will be also described.
2. TBN measurement
2.1 (n,2n) method
In the 1'4N(n,2n) 13N method, the body is irradiated with 14 MeV neutrons
produced by a deuterium-tritium (D-T) generator [11, 24, 25, 29] or with those produced
by a cyclotron [5]. This fast neutron reaction (cross section = 5.7 mb) has a high threshold
energy of 11.3 MeV and the created radioisotope 13N decays to 13C by positron emission
having a half-time of 9.96 min. The nitrogen mass in the human body can be estimated by
the detection of the 511 keV annihilation quanta which are produced in the medium as a
consequence of the positron decay of 13N. Figure 2-1 shows the y-ray spectrum obtained
from a patient counted for 6 minutes following 14 MeV neutron irradiation [29].
5
GAMMA- RAY SPECTRUM
OF PATIENT' WIL
6 min POST IRRADIATION - q = 9 x 104 n cm-2 sec - I
I
10
I
I
I
I
' (ANNIHILATION PEAK)
0.51 MeV
U2i
I-
3CI
:3
4
Z No
(.64 MeV)
"A 10.78M.V)
1.37 MOV
38CI
2.16 MeV
2.75
MV )
S103
10
20
30
40
50
60
70
80
90
I00
CHANNELS (33 keV/CHANNEL)
Figure 2-1 : (n,2n) method: Gamma-ray spectrum
obtained from a patient counted for 6 minutes
following 14 MeV neutron irradiation [29].
110
There are two major problems with this method which have been discussed since
the early 1970's when the (n,2n) method was introduced as the first technique for
nitrogen measurement in vivo. One problem comes from the fact there are no
characteristic y-rays emitted by this reaction and the 511 keV photon peak represents not
only nitrogen counts but also other positron emitters produced by fast neutrons in the
human body, such as, oxygen, phosphorus, potassium, and chlorine [5, 11, 25, 26, 29,
31]. Furthermore, the y-rays emitted in nuclear reactions can undergo Compton scattering
or pair production which will lead to additional contributions under the 511 keV peak.
Table 2-1 shows nuclear reactions that might give rise to interference with the nitrogen
measurement. Some of the interference can be reduced by making adjustments with the
counting time according to the differences of their half-lives. For example, the irradiated
subject can be counted long enough so that 30P, having a shorter half-life than '3 N,
14N(n,2n)' 3N
Cross section
[mb]
5.7
Half-life
[min]
9.96
Threshold
[MeV]
11.3
Principle
emissions
P+
O
160(p,C) 3 N
19-58
9.96
5.5
p+
P
31P(n,2n) 30 P
16.04
2.5
12.7
Cl
31P(n, c)2Al
35Cl(n,2n) 34mC1
121.2
2.82
2.3
32.0
2.0
12.9
37 Cl(n,y) 38 C1
433
1.6
530
102
1100
37.2
7.7
900
38.0
8.9
thermal
13.4
thermal
12
thermal
Element
Reaction
N
K
Na
Zn
Ca
39K(n,2n) 38K
23Na(n,y) 24Na
64Zn(n,2n)63Zn
48 Ca(n,y) 49 Ca
Table 2-1 : Reactions which might interfere with nitrogen
measurement by (n,2n) method
+
y (1.78)
p+, y (1.17, 2.12,
3.3 MeV)
y (1.64, 2.17)
p , y (2.17 MeV)
y (1.37, 2.75)
P+
y (3.08)
contributes too little to the total counts. In that case, of course, the compensation which
comes from nuclides having longer half-lives needs to be taken into account. A number
of investigators have estimated the extent of interference in measuring nitrogen using
phantoms containing each of the elemens listed in Table 2-1 and no significant
interference from those products except oxygen was observed [11, 26, 29, 31, 32]. The
interference from oxygen is induced by a knock-on proton from hydrogen in the subject.
Since the product from this reaction is same as that from the nitrogen reaction, there is no
way to distinguish these two elements. Contributions from oxygen can be estimated by
comparing 511 keV photons counted from a phantom containing nitrogen and a nitrogenfree sample. 12 - 21 % of oxygen interference is obtained by some workers [11, 26, 31,
32].
Another complication that makes it difficult to use the (n,2n) reaction as a method
for absolute measurement of nitrogen in vivo is the lack of uniformity of sensitivity
within extended subjects such as the human body [29-31, 33]. Vartsky et al. defined
composite sensitivity as the number of counts detected from the body / unit mass of
element under investigation (N) / unit incident dose delivered to the body [2]. There are
three factors that influence the spatial variation of sensitivity through the body: 1)
attenuation of the activating neutron flux in the body, 2) self-absorption of photons within
the body, and 3) detector efficiency. The distribution of fast neutrons within the subject is
dependent on the type of neutron source and the irradiation set-up which characterize the
energies of incident neutrons and determine the absorbed dose which is one of the
components of composite sensitivity. The type of detectors and the geometry of the
counting facility are other factors which affect the sensitivity because they determine the
detector efficiency of the system. Also, the sensitivity is significantly affected by the size
of the subject because of the influence on the escape probability of photons from the
subject and the penetration of the activating neutrons. The sensitivity, therefore, is quite
dependent on the whole experimental set-up employed. Estimations of the sensitivity
have been made by some workers using 14 MeV neutron generators for bilateral
irradiation and counting . Vartsky et al. reported uniformity of the sensitivity of 48 %
within a 25 cm-thick box phantom, from a fast neutron fluence experiment with a 100 cm
target-to-skin distance (TSD) [31 ]. The effects of photon attenuation and detector
efficiency were roughly estimated according to geometrical assumptions. The variation of
the sensitivity was expressed as the difference of maximum and minimum values from
their mean. Other experimental results of sensitivity were obtained by two research
groups, the Scottish Universities Research and Reactor Centre (SURRC, East Kilbride,
Scotland) and the University of Leeds (UK). A 31 % uniformity for a phantom of 23.5 cm
was reported by Leeds [11, 33, 36] while SURRC obtained a variation of 24 - 54 % for 15
- 30 cm-thick phantoms with approximately three times longer TSD and different
counting systems [26, 39, 46].
Despite the limitations inherent to the (n,2n) method, two above facilities have
established the technique for absolute nitrogen measurement [11, 25, 26, 36]. In the
SURRC bilateral irradiation was achieved by placing two D-T neutron generators above
and below the patient. The Leeds system had one D-T generator which irradiated the
patient from either side by rotating the patient couch through 1800. Both irradiation
systems allowed maintaining the patient in the supine position, which is needed for
measurement of critically ill patients. Calibration procedure taken for both techniques
used three anthropomorphic phantoms of different sizes which represent various habitus
of the human body [25]. The method is based on the assumption that the distribution of
nitrogen and interfering elements in the human body can be considered to be averaged
over the whole subject and thus the relative counts of those elements are similar to those
of the phantom. Williams et al. (SURRC) found that the reproducibility and absolute
accuracy were dependent on the size of the subject, with results in the ranges of 1.4 - 3.2
% and 3.0 - 4.3 %, respectively [25]. These results were obtained by using three
phantoms with a dose equivalent of 10 mSv. The Leeds group performed nitrogen
measurements using two phantoms by the irradiation from both side of the subject and
reported reproducibility of 1.3 - 1.6 % and accuracy of 1.2 - 2.4 % with a dose equivalent
of 5 mSv, which was 10 times higher than that used for patient studies.
2.2 PGNAA method
As another effective technique for body nitrogen measurement, the PGNAA
method was introduced at Birmingham in the UK [27] soon after the emergence of the
(n,2n) method. Unlike the (n,2n) method, 10.83 MeV y-rays produced from the neutron
capture reaction 1aN(n, y)15N* (cross section = 0.075 b) are uniquely characteristic of
nitrogen and thus can be readily identified by spectroscopic analysis. The specific y-rays
are emitted in a very short time (life time = 10- 5 sec) by the de-excitation of 1N* to the
ground state with about a 15 % branching ratio. The thermal neutrons that contribute to
this reaction are produced by thermalization of fast neutrons in the medium of the subject
or by the use of a moderator through which incident neutrons pass before interacting with
the subject.
The limitation of this method comes from the high background counts in the
nitrogen peak of 10.83 MeV y-rays. The y-ray spectrum obtained from a normal human
body for a 36 minute measurement is shown in Figure 2-2 [14]. High background is due
to the fact that counting and irradiation are performed simultaneously for prompt gamma
reactions and photons created from other elements existing in the whole system
contribute to the total counts [2, 6]. Because of the high characteristic energy, however,
large NaI(TI) crystals are needed to ensure high detector efficiency. The background is
also affected by the size and shape of the subject and thus so is the nitrogen signal.
Three types of neutron sources have been utilized for the PGNAA method: a
cyclotron [1, 27, 40], a 23 8Pu-Be radionuclide [2, 14, 20], and a 2 52Cf spontaneous fission
a 10
a.
S102
A
2
4
6
8
Energy (MeV)
10
12
Figure 2-2 : PGNAA method: Gamma-ray
spectrum obtained from a normal human body
for 36 minute measurement [14].
source [6, 22, 28, 47]. Despite the needs of replacement, the two radionuclide sources
have gained more acceptance than the cyclotron because of the lower cost for
construction, the accessibility from the clinical environment, the simplicity of the
operation, and the lower absorbed dose [17, 20, 48, 49]. A
252Cf neutron
source provides
lower nitrogen background than a 23 8Pu-Be radionuclide because the background in the
nitrogen region is mainly contributed from random summing ofy-rays in the energy range
of 4 - 7 MeV [43]. A 2 8Pu-Be source decays with photons of 4.4 MeV which may be
coincidently detected and contribute to the nitrogen region (9.5 - 11.1 MeV), while no yrays in the energy range of 4 - 7 MeV are emitted from a 252Cf source. It was also
reported that a 252Cf source provides a 40 %larger thermal neutron flux per incident dose
than a 238Pu-Be source with comparable uniformity [44, 50]. Furthermore, the
transportation regulations of a 252 Cf source are relatively less stringent than a 238Pu-Be
radionuclide [43]. The short half-life, 2.65 years, of a 252 Cf source, however, entails the
need of more frequent replacement of the neutron source than if a 238Pu-Be source is used
[7].
A special feature of this technique which made it possible to determine the
absolute amount of nitrogen in the human body is the use of total body hydrogen (TBH)
as the internal standard, which was proposed in the mid 1970's by Vartsky et al. at the
Brookhaven National Laboratory (BNL) [2, 12]. This technique takes advantage of the
fact that the ratio of nitrogen to hydrogen counts is less dependent on body size than
nitrogen counts alone. Hydrogen in the human subject can be counted simultaneously
with the counting of nitrogen by the prompt gamma technique using the thermal neutron
reaction 1H(n, y) 2H (cross section = 0.33 b). Characteristic y-rays of 2.23 MeV are
produced with 100 % yield per neutron capture. Non-uniformity of the thermal neutron
flux in the body can be reduced by normalizing the nitrogen signal relative to the
hydrogen signal. The ratio of the N/H counts observed in the subject is then compared
with that obtained from a similar-sized anthropomorphic phantom with known amounts
of these elements. This method is employed in practice for absolute measurement of
nitrogen in a number of clinical facilities[6-8, 14, 34, 35]. Calibration procedures are
basically same, although every facility has its own way of measuring nitrogen and
hydrogen background, deriving TBN and TBH using equations developed, and correcting
data to alter the difference between the standard phantom and the subject, according to
the facility construction materials and the phantoms they used. In this chapter, therefore,
the calibration technique utilized at BNL [35, 42] is described.
Figure 2-3 schematically shows the PGNAA facility currently used at BNL. The
details of the composition, location and dimension of the collimator assembly and
detectors will be provided in chapter 4. To experimentally determine absolute nitrogen
concentration in the body, the patient is irradiated over five 20 cm-long sections along the
length of the body, starting from the shoulder, for 200 sec for each section. This
procedure is repeated for the prone and supine position to achieve a bilateral irradiation
and counting and consequently the nitrogen gross counts of the patient are obtained.
detector
premoderator (D 2 0)
collimator
238 Pu-Be
source
Figure 2-3 : Schematic illustration of the PGNAA facility employed at BNL
Patient is irradiated over five 20 cm-long sections along the length of the
body starting from the shoulder, for 200 sec for each section. This
procedure is repeated for the prone and supine position to achieve bilateral
irradiation and counting.
Measurement of the nitrogen background in the energy range of 9.5 - 11.1 MeV
that corresponds to the nitrogen peak is performed using a BOttle Mannequin ABsorber
(BOMAB) phantom shown in Figure 2-4. Three BOMAB phantoms of different sizes
which are filled with nitrogen-free tissue-equivalent solution are available. One of the
three BOMAB phantoms which has a similar size to that of the subject is measured so
that the nitrogen net counts of the patient can be obtained simply by subtracting the
counts observed in the phantom from the nitrogen gross counts observed in the patient.
The net hydrogen counts are obtained by trapezoidal subtraction in the hydrogen region
O0
O0
Fat*
0
Figure 2-4: The design of BOMAB phantom
Fat layers surround the sections of thorax,
lumber and thighs [35, 51].
(2.06 - 2.47 MeV) and subsequent subtraction of interference from hydrogen in the
facility from that value. The interference is assumed to be a fixed percentage of the
hydrogen true counts, because significant dependence of the body size on the interference
was not observed from the measurements of D2 0-filled BOMAB phantoms performed by
the BNL research group. The N/H counts ratio obtained from the measurement are
compared with that obtained from the same BOMAB phantom in which nitrogen-free
tissue-equivalent solution is replaced by solution containing known amounts of these
elements. In addition, measurement of BOMAB phantoms with fat layers of various
thickness (Figure 2-4) showed that the ratio of N/H counts decrease with increasing fat
layer thickness. The method of estimating the thickness of subcutaneous adipose tissue
has not been reported yet. Corrections between the two values are made for the size and
for the thickness of fat. The BOMAB phantom is irradiated and counted with the same
procedure as the patient, i.e., 200 sec for each five sections from the shoulder for supine
and prone positions. The corrected N/H ratios are averaged using the volume of each
section as a weighting factor.
The absolute amount of body nitrogen is obtained from the relationship
TBN = C x <N/H> x TBH
where
C:
A proportionality constant determined from irradiation of the BOMAB
phantom that was assigned to the patient according to the size
<N/H>: An average size- and fat-corrected N/H value over five sections
TBH is determined from the relationship
TBH = 0.11 TBW + 0.12 TBF + 0.07 TBP
where
TBW:
total body water (obtained by dilution of tritiated water technique [52, 53])
TBF:
total body fat = wt - (TBW + TBP + BMA)
wt:
body weight
BMA: body mineral ash
= TBCa (obtained by delayed gamma activation analysis) / 0.34
TBP:
total body protein = 6.25 TBN
TBW can be estimated by administration of a small quantity of 3 H2 0 to the human
subject and the radioactivity of tritium is counted for blood sample of known volume
using a liquid scintillator. The correlation factor of the nitrogen-protein relationship, 6.25,
comes from the general assumption that nitrogen composes 16 % of protein as a rough
approximation [10].
A reproducibility of 2.1 % was reported for measurement of the N/H count ratio
by Stamatelatos et al. by repeating measurements of an anthropomorphic (Remcal)
phantom ten times within one week [35]. Other facilities which are presently utilizing the
PGNAA method for TBN measurement demonstrated reproducibility of 1.5 - 4.1 % using
an anthropomorphic or a rectangular box phantom [6, 8, 14, 34, 35].
This chapter described the technical features and limitations of two different
neutron activation analysis methods for in vivo measurement of nitrogen. The calibration
procedure for the PGNAA methods employed by the BNL research group was also
explained. In the following chapter, experimental procedures and results of the nitrogen
measurement performed using the (n,2n) method will be described and comparison of the
two techniques will be made based on the computer simulation in chapter 4.
3. Nitrogen measurement using a D-T neutron
generator
3.1 Objective
In order to assess the feasibility of the use of a 14 MeV neutron generator installed
at the Body Composition Laboratory (BCL, United States Department of Agriculture,
Human Nutrition Research Center on Aging, Tufts University, Boston) [54-57] for
nitrogen measurement in vivo, measurements of a nitrogen sample are performed.
Detector efficiency of the counting system, which was newly designed for human thigh
measurements, is also measured. In addition to the evaluation of the system, the results of
measurement of the nitrogen sample will be used for comparison with those obtained
from Monte Carlo simulation. This verification of the simulation procedure is needed
because the simulation will be also used for estimation of the sensitivity, detector
efficiency and dose of both the (n,2n) method and the PGNAA method. Furthermore,
oxygen interference which has been considered to be the major problem of the (n,2n)
method [11, 26, 31, 32] will be investigated by experiments using the 14 MeV neutron
generator.
3.2 Irradiation system
An irradiation system using a small sealed D-T neutron source at BCL has been
developed for the in vivo measurement of body oxygen, hydrogen, and carbon [54-57]
and now the feasibility of the application for nitrogen measurement in vivo is under
investigation. The main composition of this neutron source is a 13 cm-long MF Physics
A-320 vacuum tube in which a deuterium accelerator is sealed. The ion beam is pulsed at
the repetition rate of 4 - 10 kHz, where 103 - 104 neutrons per pulse are delivered. A
scanning facility was also developed for the detection of O, H, and C though the system
will not be used for nitrogen measurement. Figure 3-1 shows the irradiation system
schematically.
patient
Aluminum bed
Lead
Steel
D-T neutron generator
Figure 3-1 :(n,2n) method: irradiation set-up
The subject on the bed is irradiated by14 MeV
neutrons produced from the D-T generator.
The tube is located 64 cm beneath the scanning bed made of 3.2 mm-thick aluminum and
2
incident neutrons are collimated by a steel shield defining a 40.64 x 30.80 cm
rectangular beam window. A 5.08 cm-thick lead comprises the upper part of the
collimator, which reduces the the intensity of photons produced in the surrounding
materials by neutron capture and inelastic scattering.
3.3 Counting system for human thigh measurement
HV1
amplifier and MCA
HV2
z
NaI(TI) detector 1
x
detector 2
y
Figure 3-2 :(n,2n) method: counting set-up
Each Nal(TI) detector in plastic pieces isconnected
to the amplifier, the multi-channel analyzer and the
high voltage supply. An aluminum sheet covers the
detectors and the subject is placed on it.
A whole body counting room, which is separate from the neutron generator, is
used for counting 511 keV photons created by irradiation of the subject. They are
countedby two NaI(TI) detectors (11 x 11 x 43 cm 3) which are positioned in a
polyethylene holder and covered with a thin (< 1 mm) aluminum sheet. Figure 3-2
illustrates the whole counting system which was designed specially for measurement of
nitrogen in the human upper thighs which contain relatively large amounts of nitrogen
because of their large proportion of muscle. A patient sitting on the counting set-up is
shown in Figure 3-3.
Two NaI(T1) detectors
Bed
SIDE
Figure 3-3 : (n,2n) method: counting set-up with a patient
The Nal(TI) detectors are placed on the bed in the whole
body counting room.
FRONT
3.4 Experimental procedure
3.4.1 Nitrogen measurement
3.4.1.1 Preparation of sample
Two polyethylene bottles (29.7 x 15.2 x 22.0 cm 3 for each) were filled with
urea/water ([NH 2]2 CO) solution containing, in total, 893.4 g of nitrogen and 17.29 liter of
water. The nitrogen amount is almost twice the amount of nitrogen in the upper thighs of
reference man [58](TBN in the reference man is 1800 g. We made a rough assumption
that the amount of nitrogen in both upper thighs is almost 1/4 of that in a whole body.
3.4.1.2 Irradiation
The urea solution bottles were placed on the aluminum bed under which the small
sealed Zetatron tube is located. Figure 3-4 shows the axial view of the (n,2n) irradiation
set-up with the sample. The voltage was set 65 kV, and the sample were irradiated for 20
min. The samples were positioned 64 cm away from the 14 MeV neutron source and the
thickness of the sample along the direction of irradiation was 15.2 cm.
Two urea solution bottles
Figure 3-4 : (n,2n) method: irradiation set-up for urea solution measurement
A 14 MeV neutron generator is located beneath the aluminum bed (TSD = 64
cm). Two bottoles of urea/water solution were located on the aluminum bed
for nitogen measurement.
3.4.1.3 Counting
The whole body counting room was used for detection of 511 keV photons
emitted from the 13N decays. Calibration of the counting system illustrated in Figure 3-2
was performed by placing
137 Cs
and 60Co standard samples on a carton spacer which was
located above the NaI(TI) detectors. The nitrogen background was measured before
counting the urea sample.
After the irradiation of the urea sample, the bottles were transferred to the
counting system. Transfer time was 46 sec. The sample was counted for 20 min.
3.4.2 Measurement of detector efficiency of thigh counting system
As a sample which provides estimation of the detector efficiency for the urea
solution sample used for our experiments, we used
0.5218 ± 0.0136 ptCi.
137Cs
137Cs
solution which has an activity of
was chosen because the characteristic energy of y-rays
detected is 662 keV, which is close to 511 keV of annihilation quanta emitted from
nitrogen, and thus the error that comes from self-absorption in the sample will be
minimized. The
137Cs
solution is filled in a bottle identical to ones of urea solution. The
sample is located above two NaI(TI) detectors and counted for 10.5 min.
3.4.3 Interference of oxygen
Two identical polyethylene bottles were also filled with distilled and deionized
water (18.86 liter in total) to simulate nitrogen-free samples. Irradiation and counting
were performed in exactly same way as with the urea solution experiment except the
transferring time was 39 sec instead of 46 sec for urea solution.
3.5 Data analysis and results
3.5.1 Counting results
Gamma rays produced by the positron emission were counted in the range of
photon energy of 432 to 560 keV. Figure 3-5 shows the y-spectrum obtained from urea
solution samples. The nitrogen net counts were calculated by subtraction of the oxygen
counts. The elements investigated and counts results are listed in Table 3-1.
FULL SCALE:
8 - 2844 XeU (4 XeU/Channel),
8 - 32 Counts/Minute
RUN654
DETECTOR A SPECTRUM
RUN
STATIC
MINUTE
28
AIHO
BOTTLES
RUN TITLE: TUO UREA
GROSS CNTS/MIN: 2482.95
GROSS COUNTS: 49659
**xx USER-DEFINED UINDOU ( 432 - 568 XeU ) ****
GROSS CHTS/MIN INUINDOU: 576.45
GROSS COUNTS INUINDOU: 11529
NET CNTS/MIN INUINDOU: 219.225
4384.5
NET COUNTS INUINDOU:
Figure 3-5 : Nitrogen 511 keV y-spectrum obtained from a urea solution sample
H20
Counting
time [min]
20
20
20
' 37 Cs + H 2 0
10.5
H20
10.5
Element
Compound
(2 bottles)
Nitrogen
(2 bottles)
Oxygen
Background (2 bottles)
[NH 2 ]2 CO + H2 0
H20
137Cs
(1 bottle)
Background (1 bottle)
Gross c )unts
11529 _S107
6540 __81
6362 __80
607229 _ 779
3754 _61
Net counts
5582 ± 130
178 ± 114
603655 + 782
Table 3-1 : Counts measured from each sample
The nitrogen count and its interference were
obtained from the energy window of 432 to 560
keV.The nitrogen net counts was corrected by
transferring time and oxygen fraction in the
sample. The 137 Cs was counted in the range of
540 to 760 keV.
3.5.2 Oxygen interference
According to the meaning of the interference mentioned by Leach et al. [31 ], the
interference of oxygen was calculated as the fraction of counts contributed from oxygen
alone out of the counts produced from 14N(n,2n)13N alone. Both nitrogen and oxygen
were normalized to elemental composition of reference man [58]. As a result, the oxygen
interference obtained from phantoms filled with nitrogen-containing and nitrogen-free
solution was 4.07 ± 2.60 %. This result is much smaller than - 20 % reported by other
investigators using rectangular shaped phantoms.
3.5.3 Estimation of detector efficiency at 511 keV
The detector efficiency at 662 keV, S 6 62 for two bottles of samples was assumed
to be same as that for one bottle and calculated to be 5.98 ± 0.16 % (error due to counting
statistics).
From this result, the estimation of the detector efficiency at 511 keV was made based on
the following assumptions.
* Photons emitted from the sample impinged at right on the detectors.
*
The path lengths of all photons in the sample were a half of the thickness of the
sample (b = 7.6 cm) in average.
* The attenuation in the urea and
137
Cs solution is the same as in water.
* Attenuation in the air was negligible.
8511
'662
[(1-e
1-
a- Ureasol. 1511
NL -1
e'')Na -
1- e
Cssol. 662
where,
a: the thickness of NaI(T1) along z-axis = 10.5 cm
b: a half thickness of the sample = 7.6 cm
From the assumptions made above, the detector efficiency at 511 keV for the two bottles
of samples was estimated to be 5.74 ± 0.15 % (error due to counting statistics). The ratio
of the detector efficiencies at 511 keV and 662 keV estimated on the assumptions
mentioned above was confirmed by Monte Carlo simulation which provided detector
efficiency at 511 keV of 5.50 + 0.16 % (error due to counting statistics).
3.5.4 Estimation of neutron output
The neutron output of the D-T generator was calculated using the detector
efficiency obtained in the last section and the results of nitrogen measurement.
Assumptions:
*
The D-T generator was a isotropic point source of 14 MeV neutrons.
*
The flux of activating neutrons (11.3 MeV - 14 MeV) averaged over the sample is
proportional to the 14 MeV neutron output of the D-T generator.
The ratio of the activating neutron flux in the sample to the total neutron output was
calculated from Monte Carlo simulation. According to the simulation results, the
activating neutrons composes 11.8 % of total neutrons yielded in the sample. As a result,
the neutron output was estimated to be 4.9891 x 107 n/sec (4.6 % error due to counting
statistics).
4. Monte Carlo simulation
Computer simulations of nitrogen experiments using a Monte Carlo code are
performed for evaluation of the capabilities of the (n,2n) method and the PGNAA
method. Sensitivity and dose of each method are investigated by simulating nitrogen
measurement facilities currently in use. In the simulation several experimental problems
inherent to the two methods can be eliminated so that the statistical error of the results are
minimized. Various types of phantoms are designed for the simulation in order to
adequately assess each system. The simulation results will be used for the comparison and
the assessment of the suitability of the two methods for in vivo measurement of nitrogen.
4.1 MCNP code
Monte Carlo N-particle (MCNP) is a general-purpose Monte Carlo transport code
which has the capability to model the behavior of neutrons, photons, and electrons
through statistical sampling processes based on extensive nuclear data collected from
several sources [59]. MCNP allows the user to build three-dimensional configurations of
experimental geometries and to calculate the neutron, photon and/or electron fluence
averaged over a surface or a volume region ('cell') specified by the user based on highly
dense, point-wise cross-section data. MCNP also has the capability of calculating the
number of particles crossing the surface and the energy deposition throughout the cell.
The MCNP simulation code has been under development at Los Alamos National
Laboratory in Los Alamos, New Mexico since World War II. The use of MCNP for
simulating the behavior of neutrons (and photons) has been validated by a number of
investigators [44, 60-62]. Version 4b is used for the work of this thesis [59].
4.2 Figures of merit
4.2.1 Dose
The dose delivered to the body should be limited to the extent that the possible
risks are negligible compared to enormous benefits caused from the radiation exposure.
The lower absorbed dose is more desirable on the basis of the concept of as low as
reasonably achievable (ALARA).
The total dose equivalent reported by the PGNAA facilities is in the range of 0.14
- 0.75 mSv using the quality factor of 10 for the PGNAA method with a 238Pu-Be or a
252Cf neutron source for a 14 - 40 minute-scan [6-8, 14, 34, 42]. On the other hand, a
dose of 0.5 mSv is reported for the (n,2n) method using a 14 MeV neutron generator for
an 80 second irradiation time [11]. The dose normalized to unit irradiation time is 20 - 90
times higher for the (n,2n) method than the PGNAA technique. The dose equivalent for a
400 sec-bilateral irradiation at the PGNAA facility of BNL most recently upgraded was
estimated to be 0.8 mSv using the radiation weighting factors currently recommended by
ICRP [63], measured at the bed level by Eberline ESB2 and RO-2 monitors for neutron
and y-rays, respectively [42].
4.2.1.1 Equivalent dose
In order to examine the eqivalent dose of nitrogen measurement systems with
MCNP calculations, a 40 x 30 x 15 cm 3 box phantom is used. The phantom is filled with
a uniform solution composed of 25 % adipose tissue and 75 % muscle which roughly
approximates the composition of the human thighs. Table 4-1 shows the density and
box thigh phantom
density
[g/cm 3 ]
1.025
H
C
[%]
[%]
10.5
25.675
N
[/
2.725
O
Na
P
S
C1
K
[%]
[%]
[%]
[%]
[%]
[%]
60.2
0.1
0.15
0.25
0.1
0.3
Table 4-1 : Density and composition of the box thigh phantom
The percentages are expressed as weight fractions [64].
......
............................
7J
Figure 4-1 : The rectangular box phantom filled with a uniform solution
composed of 25 % adipose tissue and 75 % muscle. A 2 cm-diameter
and 1 cm-thick cylinder cell is investigated for the maximum skin dose
calculation.
composition of the box thigh phantom used in the simulation. The maximum skin dose is
investigated by calculating the dose delivered to a 2 cm-diameter and 1 cm-thick cell,
which is included in the box phantom and located at the center of the neutron beam
window on the bed level. Figure 4-1 shows the configuration used for the calculation of
the maximum equivalent dose. Table 4-2 shows the weighting factors recommended by
ICRP [63], which were used for the calculation of equivalent dose.
Type and energy range
Neutrons, energy < 10 keV
10 keV to 100 keV
100 keV to 2 MeV
2 MeV to 20 MeV
> 20 MeV
Photons, all energies
Weighting factor
5
10
20
10
5
1
Table 4-2 : The weighting factors recommended by ICRP [63]
4.2.1.2 Dose distribution in human thighs
The neutron and photon dose calculation in MCNP is attained by using tabulated
data of kerma factors for each radiation as fluence-to-kerma conversion factors [62, 65].
The use of the fluence-to-kerma factors for neutron/photon dose calculation is valid
because the dimensions of the cells under investigation are larger than the ranges of
secondary charged particles (- 1 cm) and thus the condition of the secondary charged
particle equilibrium is established [60]. As mentioned earlier, the calculation of neutron
and/or photon fluence is based on the cross-section data provided in MCNP. In the
simulation, neutron cross section data of natural elements are chosen from MCNP
libraries except hydrogen, nitrogen and oxygen, for which the data of 1H, ' 4N and
160
are
and crystalline
used because of the lack of natural element data. The chemical binding
to be
effects of the thigh solution on the behavior of low energy neutrons are assumed
data
same as that of water which can be corrected by use of thermal neutron cross-section
supplied in MCNP.
While total dose delivered from a whole measurement is used as a crucial measure
in the
to assess the suitability of the system for human study, the distribution of the dose
human body is also important. Knowledge of the spatial dose variation is essential
is
because the uniformity of dose contribution from activating neutrons in the subject
adipose
tissue
50 cm
muscle
11 cm
13 cm
femur
18 cm
20 cm
Figure 4-2: Thigh model for MCNP simulation
Three tissues; adipose tissue, muscle, and
femur comprise the model.
Tissue
adipose tissue (at)
muscle
femur
density
H
C
[g/cm 3 ]
[%]
[%
0.95
1.05
1.33
11.4
10.2
7.0
59.8
14.3
34.5
0
Na
%]
[%]
[%]
0.7
3.4
2.8
27.8
71.0
36.8
N
0.1
0.1
0.1
Mg
[%]
0
0
0.1
P
[%]
0
0.2
5.5
S
Cl
K
Ca
[%]
[%]
[%]
[%]
0.1
0.3
0.2
Table 4-3 : Density and composition of adipose tissue, muscle, and femur
which comprise the thigh model. The percentages are expressed as weight
fractions [64].
0.1
0.1
0.1
0
0.4
0
0
0
12.9
needed for absolute measurement of nitrogen in the body. To simulate the human thigh
measurement mentioned in chapter 2, a model of human upper thighs was designed.
Figure 4-2 illustrates the model consisting of three tissues; adipose tissue, muscle, and
femur [64, 65]. Density and composition of each tissue is listed in Table 4-3.
4.2.2 Sensitivity
The uniformity of sensitivity to nitrogen detection in the subject is important for
the absolute measurement of nitrogen and the investigation of this critical factor should
be made to evaluate different neutron sources and the experimental set-ups. It should be
noted again that the sensitivity is influenced by three factors: activating neutron flux, selfattenuation of y-rays detected, and detector efficiency. For sensitivity calculations a
rectangular box phantom model will be used. The phantom is filled with the thigh
solution composed of elements shown in Table 4-1. A 40 x 40 x 15 (depth) cm 3 phantom
is divided into smaller cells so that the spatial variation of sensitivity within the phantom
can be examined. The region under investigation of sensitivity variation is limited to the
horizontal size corresponding to the neutron beam window of the nitrogen measurement
system.
4.2.2.1 (n,2n) method
The composite sensitivity can be obtained as follows.
At the time irradiation stops, the activity of ' 3N; A 13 [#/min] is:
-
A 13 = R 13 x(1-e
Ato
(1)
where R13: (n,2n) reaction rate [#/min]
X: decay constant of 3N = ln2/T 1 /2 [min-l]; T1/ 2 = 9.96 [min]
to: irradiation time = 20 [min]
When the end of the irradiation is considered to be t = 0, the activity of 13N after the
irradiation is:
A 1 3 (t) = R 13 x (1 - e-Ato) x e - t
(2)
In the simulation, the transferring time i.e., the time between the irradiation and the
counting, is assumed to be 46 sec which was taken in conducting nitrogen experiments at
the BCL facility (chapter 3). The number of disintegrations during the counting time 20
min (= t 2 - ti); D13 [#] is:
D13
= fA3(t)dt
= R13 (1 -
e
)(e2t1 - e - 2)
(3)
where tj: time when the counting starts = 46 [sec] = 0.7667 [min]
t2 : time when the counting ends = 46 [sec] + 20 [min] = 20.7667 [min]
Net counts of 511 keV annihilation quanta; Cn2n [#] is:
Cn2n = 6511 x e1
3
x D13
(4)
where
8511:
absolute detector efficiency [%]
el 3: probability that a 511 keV photon is emitted per disintegration of 13N = 200 [%]
Composite sensitivity; Sn2n [#/g/Sv] can be obtained such that:
Cn2n
Sn2 n -_n
mNn2n x dn n
2
x
(t 2 -t)
'511 x e1 3 x R 13(I - eAto )(e-t - e-t2
(5)
mNn2n x dn2n x to x A
where
mNr2n: amount of nitrogen for one cell of thigh solution box phantom [g]
den: incident skin dose rate [Sv/min]
The incident dose can be estimated from the results of the dose calculation.
Obviously unknown values in estimating the sensitivity are only the (n,2n) reaction rate
and the absolute detector efficiency, which will be calculated by MCNP simulation.
Since 14N(n,2n) 13 N is a delayed gamma reaction, two MCNP runs are needed in
order to obtain the sensitivity of one cell for the (n,2n) method. The first run provides the
occurring rate of (n,2n) reactions in each cell (= RI 3 [#/min]), and the second run
calculates the absolute detector efficiency supposing the cell to be a uniform source of
photons of 511 keV(= Es11 [%]). Incorporation of these two products into the above
equations results in the composite sensitivity for a unilateral nitrogen measurement at a
given position in the phantom.
In nitrogen measurement using the (n,2n) method, as shown in Figure 3-3, a
patient will be asked to sit above the detectors so that the irradiated parts of the thighs
will be detected from below. If the irradiation is conducted in a prone position, decreasing
detector efficiency with the thickness of thighs will be compensated with increasing fast
neutron flux delivered to the subject. The discrepancy in sensitivity measurement
resulting from irradiation in different positions will be also examined.
4.2.2.2 PGNAA method
Because of the prompt gamma reaction of nitrogen, the sensitivity measurement
for the PGNAA method could be performed in one MCNP run by building both
irradiation and counting systems. In this case, only the cell of interest would contain
nitrogen with the rest of the phantom filled with a nitrogen-free solution. Such
simulations, however, require too much computer time to obtain results with acceptable
fluctuations due to the low cross section of 14 N(n,y) 1 N reaction and the relatively low
detector efficieircy of the BNL counting system. Thus sensitivity of the PGNAA method
will be obtained by separating the measurement into two steps as was done with the (n2n)
method, i.e., irradiation and counting. Since there is no need to take into account the loss
of the nitrogen counts during the irradiation and transferring time,
R15 x t' = D15
(6)
for the PGNAA, where the notation '15' indicates
15N,
and t' is the measurement time.
Using eq. (5), composite sensitivity for the PGNAA can be obtained as following:
CPG
mNPG x dpG x t'
-
810.83 x e 1 5 x
mNr
R1 5
(7)
xdPG
The notations of 'PG' and '10.83' indicate the PGNAA method and 10.83 MeV which is
the energy to be detected, respectively. e15 = 15 %;the branch ratio of the disintegration
to the ground state. As with the (n,2n) method, the reaction rate of 14 N(n,y)'N (= R 15
[#/min]) and the absolute setecter efficiency (= 10.8 3 [%]) obtained by MCNP simulations
leas to the composite sensitivity of the PGNAA technique.
4.2.2.3 Reaction rate
MCNP calculates the total number of a requested type of reaction which occurs in
all elements existing in a cell. The calculation is made in a way that the average neutron
fluence over the cell is multiplied by the total number of atoms and the microscopic cross
section of the reaction specified. The rate of 14N(n,2n)' 3N reactions occurring in the
phantom can be obtained by specifying (n,2n) reactions because nitrogen is the only one
nuclide for which the (n,2n) cross section data are available among all nuclides in the
phantom. The estimation of nitrogen counts made by this method can be free from
interference problems associated with the (n,2n) nitrogen measurement.
On the other hand, it is not possible to calculate the occurring rate of 14N(n,y)'N
reactions in the same way because MCNP provides the total number of the (n,y) reactions
occurring from all other elements existing in the phantom. Instead, tabulated results of the
(n,y) activity in each nuclide given after the simulation will be used to estimate the
number of neutron absorption reactions from nitrogen for the PGNAA method. The high
background contributed from the
238Pu-Be
source will be excluded in the simulation by
separating the measurement into two stages.
4.2.2.4 Absolute detector efficiency
The absolute detector efficiency can be estimated by use of a "pulse height tally"
provided by MCNP. This tally scores the number of photons which deposit energies
within the range specified by the user, by all the tracks passing through a cell. The
absolute detector efficiency of each cell in the thigh phantom is obtained by defining an
isotropic photon source uniformly distributed in the cell. The energy of the photons is 511
keV for the (n,2n) method and 10.83 MeV for the PGNAA method. Therefore, the
detector efficiency results obtained by the simulation are only determined by two factors:
the counting geometry and the energy of the photon source. The photon energy range is
set at 432 - 560 keV and 9.5 - 11.1 MeV for the (n,2n) method and the PGNAA method,
respectively.
4.3 Nitrogen measurement systems
Measurements of dose and nitrogen counts using four different experimental
systems are simulated by the MCNP code. The four types of nitrogen measurement
system are listed in Table 4-4. The details of the source characteristics and the geometric
configurations defined in the simulations of one (n,2n) system and three PGNAA systems
will be described in this section. As a neutron source for the PGNAA method, the
feasibility of the use of an accelerator recently built in the Laboratory for Accelerator
Beam Applications (LABA) at the Massachusetts Institute of Technology will be
investigated.
Method
(n,2n)
Neutron source
D-T generator
PGNAA
2Pu-Be
PGNAA
7Li(p,n),
Be(p,n),
9Be(d,n)
PGNAA
Li(p,n), Be(p,n),
9Be(d,n)
Source of input data
Experimental set-up currently used at BCL
Experimental data
Experimental set-up currently used at BNL,
Reports in literature
Experimental set-up practicable at LABA,
Neutron spectrum experimentally measured by LABA
Experimental set-up currently used at BNL,
Neutron spectrum experimentally measured by LABA
Table 4-4 : Four types of nitrogen measurement systems simulated by MCNP
4.3.1 D-T generator - (n,2n) method
The method of nitrogen measurement employing the 14N(n,2n)13N reaction is
described in chapter 3. The simulated irradiation system with the thigh model is shown in
Figure 4-3. In the simulation the neutron source is defined as a 14 MeV isotropic point
source and the neutron output of 4.9891E7 n/sec is used. The target-to-skin distance
(TSD) is 64 cm. It is also assumed that the material under the irradiation system consisted
of 100 % concrete. The composition of the materials simulated by MCNP is listed in
Table 4-5. Figure 4-4 shows the irradiation set-up with a box thigh phantom used for
sensitivity calculation. The region investigated is 40 (width) x 30 (length) x 15 (depth)
cm 3 which corresponds to the size of the neutron beam window. The phantom is divided
into 36 cells of 10 (width) x 10 (length) x 5 (depth) cm 3 for calculations of sensitivity
distribution.
The configuration of the counting system with the box thigh phantom is shown in
Figure 4-5. Two rectangular-shaped NaI(TI) counters are placed in parallel in the
polyethylene holder so that the human thighs are measured from below. The composition
of NaI(TI) crystals is sodium and iodide at the atomic ratio of 1:1 and the thallium
material
density
3
[g/cm ]
lead
11.4
steel
7.855
aluminum
bed
concrete
2.6989
2.3
compositions
size
[%, in weight]
[cm]
outer: 99.06 (x), 61.28 (y), 5.08 (z)
Pb 100
inner: 40.64 (x), 30.80 (y), 5.08 (z)
outer: 99.06 (x), 61.28 (y), 50.8 (z)
Fe 70, Cr 19, Ni 11
inner: 19.69, 30.16, 40.64 (x)
30.80 (y), 5.08,10.16, 35.56 (z)
86.36 (x) x 193.04 (y) x 0.32 (z)
Al 100
200 (x) x 200 (y) x 50 (z)
(approximately)
O 52.9, Si 33.7, Ca 4.4,
Al 3.4, Na 1.6, Fe 1.4,
K 1.3, 'H 1, Mg 0.2, C 0.1
Table 4-5 : Density, size, and composition of each component of the collimator assembly
simulated for the BCL (n,2n) system
activator is neglected in the simulation. Two detectors are separated by 2 cm. The parts of
the photomultipliers are assumed to be 100 % polyethylene as is the composition of the
polyethylene holder. The thin (< 1 mm) aluminum sheet covering the detectors is
neglected in the simulation. Table 4-6 lists the density, size, and compositions of
materials used in the simulation.
material
density
[g/cm3 ]
sodium-iodide
crystal
photomultiplier
3.67
polyethylene
size
[cm]
11 x 11 x 43
compositions
[atomic fraction, except
borated polyethylene]
Na:I = 1:1
1
6 (dia), 13 (long)
C:H = 1:2
1
55 (x), 52 (y), 13 (z)
C:H = 1:2
Table 4-6 : Density, size, and compositions of each component of the detector and the
shielding materials simulated for the BCL (n,2n) system
o0
Sz
s
...
.
9:
X
oro
nbed,
o
on
A-*
o
on
I
*;-----; r--% ~
; -'
~~:1~~~:~:~
~-''X-
o
I
i'
ou,
1
ad
,rete;
so0
-60
-40
-20
0
20
40
60
80
-80
-60
-40
-20
0
20
(b) lateral view
(a)axial view
Figure 4-3 : (n,2n) method: BCL irradiation system with the thigh model
for calculation of dose distribution by MCNP
40
60
aluminum bed
lead
concrete.'
T7
A
-80
-60
-40
-20
(a) axial view
0
20
40
60
80
-80
-60
-40
-20
20
0
(b)lateral view
Figure 4-4: (n,2n) method: BCL irradiation system with the rectangular box phantom
filled with thigh solution for sensitivity calculation by MCNP
4
z t
'
polyethylene holder '
Nal detectors
40
-40
-20
0
20
40
(a) axial view
.(b) lateral view
Figure 4-5 : (n,2n) method: BCL counting system with the rectangular box phantom
filled with thigh solution for sensitivity calculations by MCNP
4.3.2 238Pu-Be radionuclide - PGNAA method
The simulation of nitrogen measurement by the PGNAA technique employed at
BNL is carried out using information supplied in published BNL papers [2, 35, 42]. The
schematic configurations of the neutron irradiation set-up with the thigh model are shown
in Figure 4-6. Because no detailed neutron spectrum data of the
238 Pu-Be
radionuclide
are provided [66], that of a 239Pu-Be neutron source is used for the energy range above 0.5
MeV [67] because of the similarity of the energies of ct-particles emitted from
239Pu.
2 38
Pu and
The influence of this modification on the simulation results can be estimated
negligible because the range of the neutron energy 0 - 11 MeV and the average neutron
energy 4.5 MeV reported in Vartsky's paper [2] are very close to those used for the
simulation (0 - 10.5 MeVand 4.6487 MeV, respectively). The neutron yield data used for
commission the simulation are shown in Figure 4-7. The neutron source is assumed to be
isotropic in the simulation and the source strength is estimated to be 2.08 x 108 n/sec [2].
The
239Pu-Be
point source is modeled as being surrounded by a graphite reflector
with a pyramid-shaped opening inside. The neutron source is positioned at the apex of the
pyramid and TSD is 72 cm. The size of the rectangular beam aperture of the graphite is
28 (width) x 11.2 (length) cm 2 . In the simulation the graphite is assumed to consist of 100
% 12C for which corrections are made for the chemical binding effects on thermal neutron
scattering. The graphite block is surrounded by bismuth, lithium-doped resin, and borated
lead, in that order, and the entire assembly is capped with a thick bismuth block providing
the beam aperture of 40 (width) x 16 (length) cm 2 . Density, size, and compositions of
graphite
density
[g/cm3]
2.1
bismuth
9.8
material
compositions
size
[cm]
[%, in weight, except D20]
C
100
72
(z)
30
(y)
x
outer: 30 (x) x
inner: 11.2 (x) x 28 (y) x 42 (z)
Bi 100
5 (thickness), 34 (z)
(side)
lithium-doped
1
28 (thickness), 53 (z)
C 79.6, 1H 13.0, 6Li 1.1,
7Li
resin
0.9, F 3.1, 160 2.3
borated lead
11.4
12 (thickness), 12 (z)
Pb 95.0, '°B 5.0
bismuth
9.8
120 (x) x 120 (y) x 18 (z)
Bi 100
68 (x) x 68 (y) x 2.5 (z)
7H 1
2 H:' 6 0
( upp er )
heavy water
1.1034
1
= 1:1
(atomic fraction)
Table 4-7 : Density, size, and composition of each component of the collimator assembly
simulated for the BNL PGNAA system
each component are listed in Table 4-7. Bi is an excellent shielding material because of
its high density and low yield of (n,y) reactions. The Li-doped resin shields neutrons and
minimizes the interference for hydrogen measurement because of the high neutron
capture cross-section of lithium. Although no consideration for the hydrogen
measurement is needed for the work of this thesis, this composition is also emulated
because the component may affect the dose measurement. The borated lead is used to
shield thermal neutrons and y-rays also because of the high cross-section of the thermal
neutron reaction lB(n,ca) and the high density of the lead. The fast and epithermal
neutrons collimated by the assembly are premoderated by heavy water before interacting
in the phantom. The thickness is 2.5 cm, which is chosen by the BNL researchers as a
compromise between the yields of nitrogen counts and the uniformity of composite
sensitivity based on the experimental results. The chemical binding and crystalline effects
of D20 are corrected as was done for light water. The size of the neutron beam window is
48 (width) x 19.2 (length) cm 2 . Figure 4-8 shows the irradiation system with the
rectangular box phantom filled with thigh solution simulated for sensitivity calculations.
The region measured for sensitivity calculations is 40 (width) x 20 (length) x 15 (depth)
cm 3 and the variation of the sensitivity is investigated by dividing the phantom into 48
cells of 10 (width) x 5 (length) x 5 (depth) cm 3.
material
density
[g/cm 3]
sodium-iodide
crystal
photomultiplier
3.67
bismuth
9.8
borated carbide
(side)
borated carbide
(front)
boric acid
borated
polyethylene
15.2 x 15.2 x 15.2
compositions
[atomic fraction, except
borated polyethylene]
Na:I= 1:1
5.2 (dia), 30.8 (long)
C:H = 1:2
Bi = 1
2.52
behind: 2 (thick), 30.8 (long)
front: 4 (thick), 18.7 (long)
5 (thick), 30.8 (long)
2.52
15.2 x 15.2 x 3.5 (thick)
1.435
behind: 7 (thick), 30.8 (long)
front: 5 (thick), 18.7 (long)
28 (dia), 2.5 (thick)
1
1.067
size
[cm]
lB:1B:C
= 0.85 : 3.15 : 1
°TB:B:C
= 0.85 : 3.15 : 1
10B: "B: 1H: 16 0 = 0.2135
0.7865 : 3 : 3
l'B 0.99,11B 4.01, 'H 13.571,
C 81.429 [%, in weight]
Table 4-8 : Density, size, and compositions of each component of the detector and the
shielding materials simulated for the BNL PGNAA system
In the PGNAA method, the characteristic y-rays (10.83 MeV) emitted promptly by
the neutron capture reaction in nitrogen are detected by two 15.2 cm-cube NaI(TI) crystal
counters. Figure 4-9 shows the detectors positioned at 600 to the z-axis and at 200 at the
x-axis in the simulation. The center of the detector window is 55 cm above the bed level.
The location of the detectors is optimized by BNL researchers so as to minimize the
nitrogen background [35] and the nonuniformity of composite sensitivity in the subject
[42] for the absolute measurement of TBN. The detectors are shielded with several
materials to reduce the nitrogen background and the hydrogen interference for the BNL
calibration procedure. The simulated configuration of one of the two detectors is shown
in Figure 4-10. Bismuth, which has a high density, shields the sides of the detectors from
y-rays. Boron carbide surrounds the photomultiplier and covers the front of the detector.
The outer layer of the detector consists of boric acid and a borated polyethylene slab is
added on the front surface. '0 B contained in the last three materials, as Li was used in the
collimator, reduces the number of H(n,y)D reactions which lead to interference in the
hydrogen measurement. Table 4-8 lists the density, size, and composition of materials
used for detector simulation.
I
.
.
I
.
.
.
.
I
.
z
.
z
bx
--
borated lead
D 20
y
I i
0~
Li-doped resin
I-60
-60
-40
-20
0
(a) axial view
20
40
60
80
-80
-60
-40
-20
0
20
40
60
-40
-20
0
20
40
60
(b) lateral view
Figure 4-6 : PGNAA method: BNL irradiation system with the thigh model for calculation of dose distribution by MCNP
A 238Pu-Be neutron source was positioned at the apex of the pyramid-shaped opening shown at the center of the
collimator (TSD = 72 cm).
80
logarithmic scale
100 -
A
10 4
.
01
1.
.1
.
I
E
1.00E-07
I
I
1.00E-06 1.00E-06
II
1.00E-04
I
1.00E-03 1.00E-02
1.00E-01
1.00E+00 1.00E+01
1.00E+02
Neutron Energy [MeV]
linear scale
12Y
10.
*
.
8-
*
+
2* " 64
**
2-
2
4
6
8
10
Neutron Energy [MeV]
Figure 4-7 : 238 Pu-Be neutron yield data points used for the simulation
Experimental data obtained using a 238Pu-Be were used for the energy
range below 0.5 MeV[66], shown in the log scale (upper figure), and
the data of a 239Pu-Be was used for the range above 0.5 MeV[67],
shown in the linear scale (lower figure).
1
4-..
1
I
I
._
I
,
.
I
~
I
.'
.1
I-
z
I ;s:i i
I
IJ
I
,'
borated lead
D20
I
Li-doped resin'
i
-80
-60
-40
-20
0
(a)axial view
20
40
60
-80
-60
-40
-20
0
20
(b) lateral view
Figure 4-8 : PGNAA method: BNL irradiation system with the rectangular box phantom for sensitivity calculations
A 238pu-Be neutron source is positioned at the apex of the pyramid-shaped opening shown at the center of the
collimator.
40
60
60
7
i0
Y:,
inn
0
L 1--0
0
- 50
so
lO1
-inn
(b) axial view
(a) top view
Figure 4-9 : PGNAA method: BNL detector positioning simulated by MCNP
The shielded detectors are positioned at 600 to the z-axis and at 200 at the x-axis.
60
.
1
I
.
. . .
A . . . . . .
...
boric acid
I
- bismuth
Nal detector
-40
-20
0
20
40
-40
-20
Figure 4-10 : PGNAA method: one of the BNL detectors simulated by MCNP
A 15.2 cm-cube Nal crystal is shielded with several shielding materials.
0
20
40
4.3.3 Accelerator source 9Be(p,n) with LABA set-up - PGNAA method
Figure 4-11 shows the picture of a 4.1 MeV tandem electrostatic accelerator
which has been developed and recently installed in MIT's LABA [68, 69]. The
accelerator was designed to produce high-currents of charged particles and thus high
intensity neutron beams for boron neutron capture therapy (BNCT) research. It will be
possible to generate proton or deuteron beam currents up to 4 mA. Furthermore, the
recent installation of a switching magnet allows the set-up of multiple experimental
equipment simultaneously using five separate beam lines. Figure 4-12 illustrates the
configuration of the LABA accelerator and the five endstations utilized for different types
of experiments. Currently three reactions are under investigation as most likely effective
neutron sources for BNCT: 7Li(p,n), 9Be(p,n) and 9Be(d,n). The neutron energy spectra
produced from protons of various energies on a thick target were obtained by the LABA
research group in collaboration with the University of Ohio [70, 71].
MCNP simulation is used to investigate the 9Be(p,n) reaction as a possible
neutron source for nitrogen measurement employing the PGNAA method. The beryllium
target is bombarded by protons accelerated up to 4 MeV through the aluminum beam
tube. The neutron energy spectra experimentally measured are used to define the source
in the simulation. The source is defined as a surface source with 2.307 cm-diameter. The
Be target and Al tube are surrounded by heavy water and a thick graphite reflector. The
produced neutrons are premoderated by the 15 cm-thick
O)0.
This configuration has been
developed for the purpose of creating an epithermal beam which is needed for BNCT
research. Figure 4-13 shows the moderator/reflector assembly with the thigh model. A
patient is assumed to be irradiated in a standing position. The thigh model is located 16
cm away from the Be target. The characteristics of the geometry and the materials used in
the simulation are listed in Table 4-9. The configuration with the box thigh solution
model shown in Figure 4-14 is used for the measurement of the distribution of the
number of 14N(n,2n) 13N reactions in the phantom.
material
aluminum tube
graphite
reflector
heavy water
density
[g/cm 3]
2.6989
2.1
1.1034
size
[cm]
29.65 (long)
(od),
2.007 (id), 2.307
compositions
[atomic fraction]
Al 1
around Al tube: 20.307 (thick), 18 (long)
around D2 0: 18 (thick), 15 (long)
9.23 (dia), 15 (long)
12C
1
H:
6
= 1:1
Table 4-9 : Density, size, and composition of each component of the detector and the shielding
materials simulated for the LABA PGNAA system
...
....
...
....
.. ...
.. ...
.....
...
.
. .
;*:;
_
)r
::-:'
Figure 4-11 : A tandem electrostatic accelerator installed in MIT's LABA is shown with the ion source
next room and
Protons or deuterons are accelerated through the beam tube extended to the
9
bombarded on the thick Be or Li target [68]. The feasibility of the use of the Be(p,n) as a neutron
source for nitrogen measurement is investigated.
64
4
Figure 4-12 : Configuration of the LABA accelerator and experimental rooms
The installation of a switching magnet provideded the five separate beam
lines [68].
i
|,
1
::r:^
I,,.
....
i~~
; s-
.
~~~:;ai-~~~
~1
.I...
! ;-;-;
:~i
i
I
I
I
I
y
;
~8~::~
-;;
,,,,,
fl
_
: i--;
o-
:Z ;*-ri:"
o
|
--
"- -
~
-- 40 -- 0
,
-"
~~
-2
- r-
--
-- -- ---
i: ;i
-
i
;Isl_
.3~
":-~;
"
~~
o
I
.
: ;::
I
o
!
....:
i:
i
-4
- ..: --**: "
0
(a) axial view
10
20
30
-30
-20
-10
0
10
20
(b) lateral view
Figure 4-13 : PGNAA method: LABA irradiation system with the thigh model for calculation of dose distribution by MCNP
The Be target is bombarded with 4 MeV protons accelerated through the aluminum tube surrounded by the graphite
reflector. The produced neutrons are moderated by 15 cm-thick D20.
30
40
-40
-30
-20
-10
0
10
20
30
40
-
(a) axial view
-30
-20
-10
0
10
20
(b) lateral view
Figure 4-14 : PGNAA method: LABA irradiation system with the rectangular box phantom for sencitivity calculations
67
30
40
4.3.4 Various accelerator sources with BNL set-up - PGNAA method
The 238pu-Be neutron source used at the BNL facility is replaced by various
accelerator sources in the MCNP simulation. The reactions and the proton/deuteron
yield and
energies investigated in the simulation are listed in Table 4-10. The neutron
energy range for each neutron source measured at LABA is also listed. Experimentally
determined neutron energy spectra for those sources are shown in Figure 4-15. The
Reaction
7Li(p,n)
9Be(p,n)
9Be(p,n)
9Be(d,n)
9Be(d,n)
Proton energy
[MeV]
2.5
3.0
4.0
1.5
2.6
Neutron yield
[n/min/pA]
4.86E07
1.44E10
6.00E10
n/a
n/a
Neutron energy range
[MeV]
0- 1.0
0-1.0
0 - 2.0
0-6.3
0-7.3
Table 4-10 : LABA accelerator neutron reactions used for the MCNP simulation
composition and the positioning of the surrounding materials including the counting
of the BNL
system are defined exactly in the same way as those used in the simulation
the 238Pu-Be
system. By investigating this hypothetical system, the comparison between
and the accelerator sources can be made, regardless of the effects of the experimental
assemblies for nitrogen measurement, to assess the suitability as neutron sources for the
PGNAA method.
Furthermore, proton currents of the charged particles which will give the same
238
which
level of incident dose as the Pu-Be are calculated to estimate the source strength
BNL.
would make the accelerator neutron sources as effective as the PGNAA system at
0
1
2
3
4
5
Neutron Energy [MeV]
Figure 4-15 : Neutron energy spectra for various accelerator sources
6
7
8
4.4 Results
4.4.1 Dose
4.4.1.1 Incident skin dose
The incident skin dose was calculated at the center of the neutron beam window
of the irradiation systems for the (n,2n) and PGNAA techniques. Calculated results of
equivalent dose per unit irradiation time for BCL and BNL systems are presented in
Table 4-11. Neutron yields of the D-T generator and the
238Pu-Be
radionuclide source
used for the simulation are also shown.
Neutron yield
[n/sec]
Equivalent dose
per unit time
[mSv/sec]
(error bar)
BCL - (n,2n) method,
4.9891E7
1.6490E-3
D-T generator
BNL- PGNAA
2.0300E8
(3.01 %)
3.0464E-3
Method and facility
method,
2 38Pu-Be
(4.47 %)
Table 4-11 : Neutron yield and equivalent dose per unit irradiation time
for two nitrogen systems: BCL ((n,2n)) and BNL (PGNAA).
The proton currents for the LABA accelerator source, 4 MeV 9Be(p,n), which
would give same incident skin dose as the BNL
238Pu-Be
source are listed in the Table 4-
12. The calculations were made for the collimator/reflector assemblies of BNL and
LABA. Table 4-13 lists the results of proton currents for other types of (p,n) reactions.
The BNL collimating assembly was used for all different neutron sources and the
simulation of the 2.5 cm-thick D20 moderator was neglected for this case.
Neutron source
4.0 MeV p-Be
w/ BNL system
4.0 MeV p-Be
w/ LABA system
Neutron yield
[n/min/pA]
6.00E10
Proton current
[pA]
0.2703 ± 0.0178
6.00E10
0.0982 ± 0.0069
Table 4-12 : Proton current used for a 4 MeV 9Be(p,n) reaction, which will give
same dose as the 238pu-Be for the PGNAA irradiation systems.
4.0 MeV p-Be
Neutron yield
[n/min/pA]
6.00E10
Proton current
[pA]
0.1937 + 0.0111
3.0 MeV p-Be
1.44E10
1.3823 ± 0.0831
2.5 MeV p-Li
4.86E 10
0.2088 + 0.0110
Neutron source
Table 4-13 : Proton current which will give same dose as the 238 Pu-Be for the
PGNAA irradiation systems. No D20 moderator.
4.4.1.2 Dose distribution
Figure 4-1 - 4-3 shows the dose distributions, which are contributed from
neutrons, fast neutrons (11.3 - 14 MeV), and photons, respectively, along the thigh length.
The five plots shown in the figure correspond to the dose delivered to adipose tissue 1,
adipose tissue 3, muscle 1, muscle 3, and femur, indicated in Figure 4-15.
In the same way, the dose distributions of neutrons, thermal neutrons, and photons
are shown in Figure 4-4 - 4-12 for three different irradiation systems utilizing the
PGNAA method: the BNL PGNAA facility with the 238pu-Be source, the LABA facility
with the accelerator source, and the combined system of the BNL facility and the
accelerator source. As the accelerator source, the 9Be(p,n) reaction was used for both
experimental set-ups of the BNL and the LABA. The proton beam accelerated up to 4
MeV at a current of 0.2703 pA was used to provide the same incident maximum dose as
the
238Pu-Be
source. Table 4-14 lists the each radiation dose averaged over the thigh
length for five sections in the thighs.
adipose tissue 3
muscle 3
muscle 3
femur
femur -
adipose tissue 1
: neutron beam
Figure 4-16 : Axial view of the thigh modelled by MCNP
Doses given in adipose tissue 1,3, muscle 1,3, and
femur were plotted along the length of thigh.
Measurement
system
D-T generator
(BCL)
23 8Pu-Be
(BNL)
4 MeV, .0982pA
9Be(p,n) (LABA)
4 MeV, .2703pA
9Be(p,n) (BNL)
Type
of
radiation
total
neutron
activating
neutron
(11.3-14)
total
photon
total
neutron
activating
neutron
(0- 1E-6)
total
photon
total
neutron
activating
neutron
(0- 1E-6)
total
photon
total
neutron
activating
neutron
(0- 1E-6)
total
adipose
tissue 1
[ptGy/min]
1.7E+1
muscle 1
femur
muscle 3
[ tGy/min]
1.2E+1
[pGy/min]
2.6E+1
[ptGy/min]
6.4E+0
adipose
tissue 3
[ptGy/min]
6.1E+0
9.4E+O
6.8E+0
1.6E+1
4.OE+O
4.OE+O
2.OE+O
2.OE+O
7.2E+O
1.2E+O
9.1E-1
3.lE+1
2.1E+1
4.1E+l
9.5E+O
8.4E+0O
5.1E-3
2.9E-2
7.6E-2
1.OE-2
6.0E-4
2.OE+O
2.2E+O
8.1E+O
1.2E+O
8.5E-1
1.8E+O
8.7E-1
7.5E-1
6.2E-2
1.5E-2
7.8E-2
3.OE-1
4.2E-1
3.3E-2
1.4E-3
1.7E+O
1.8E+O
4.7E+0O
5.8E-1
3.6E-1
2.1E+1
1.1E+l
1.3E+1
2.1E+0O
1.5E+0
1.2E-2
6.4E-2
1.6E-1
2.OE-2
1.1E-3
3.3E+0
3.8E+0
1.4E+1
2.OE+0
1.2E+O
photon
Table 4-14 : Activating neutron dose averaged over the thigh length for five
sections in the thighs: adipose tissue 1,3, muscle 1,3, and femur. Relative error
ranges from 0.07 to 2.10 % which represents statistical precision of simulated data.
(n,2n) method - D-T generator at BCL
Neutron Dose
1 I I 1 I 1 I ( I I I I 1 ) I I I 1 I 1 I 1 I I ( I 1 I 1 I I I I L~( I 1 1 I I I I I I
4]b
mcnp
05/02/98 17:2 8:37
.~.---i
.,
iI----i- ---~---i-
tally
- I-
N-
114
n
10000 00
nps
bin normed
mctal - mc/ddth
------1 ------4....-----.C------'''
4-------I-------------------------...
-- Z
f
cell
1
d
flag/dir
1
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cosine
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energy
3 t
t
time
1
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-
.
-----------
--
--
.
a.t. 3
___-----+'-:--___-
muscle 1
muscle 3
femur
2
4
y-axis
6
8
1
: 3-8 inside beam window
Figure 4-17 : (n,2n) method: Total neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1,
adipose tissue 3, muscle 1, muscle 3, and femur.
(n,2n) method - D-T generator at BCL
11.3 - 14 MeV Neutron Dose
.
..
.
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05/02/98 17:28:37
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1
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0
metal = mc/ddth
f
cell
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flag/dir
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in
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'
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4
2
y-axis
1
I
I
6
. . . . . . . . .
8
10
3-8 inside beam window
Figure 4-18: (n,2n) method: Activating neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1,
adipose tissue 3, muscle 1, muscle 3, and femur.
femur
(n,2n) method - D-T generator at BCL
Photon Dose
I..
. .. .
t.
i . . . . . . . . .
. .
.
.. 0innp
mcnp
. . .
4b
05/02/98 17:28:37
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---
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metal
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------
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flag/dir
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------
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414
a.t.
1
a .t . 3
-------I
---
muscle 1
~-----------------
2
femur
.-----------
.
4
6
--- muscle 3
8
10
y-axis : 3-8 inside beam window
Figure 4-19: (n,2n) method: Photon dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1, adipose
tissue 3, muscle 1, muscle 3, and femur.
PGNAA - 239Pu-Be source with BNL setup
Neutron Dose
.
.
. ..
.
. .
.
.
.
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..
4D
01/28/98 17:58:59
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tally
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i
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%, \
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muscle 1
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€
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.
. . .
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i .
l
l . l .l l .
.
l
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] l l
l l
y-axis : 4-7 inside beam window
Figure 4-20 : PGNAA method with 238Pu-Be source:
Total neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1,
adipose tissue 3, muscle 1, muscle 3, and femur.
femur
PGNAA - 239Pu-Be source with BNL setup
Thermal Neutron Dose
.
. . 1I . .
.., ,
. .
,
,
, ,. i I
.,
.
.
. . . . . . . . . . ., . . . .,
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,/\ /
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""
- "-
. . ."
,
a.t.
3
muscle
muscle
femur
4
6
8
10
y-axis : 4-7 inside beam window
Figure 4-21 : PGNAA method with 238Pu-Be source:
Activating neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1,
adipose tissue 3, muscle 1, muscle 3, and femur.
PGNAA - 239Pu-Be source with BNL setup
I
I I
I I I I I
Photon Dose
.. . . I
I
, .
I .I
.,
.
I I
I
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4b
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01/28/98 17:58:59
414
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1
metal - mc/pdth
iN
/
/
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flag/dir
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e
energy
t
time
a.t. 1
a.t.
3
muscle
muscle
femur
2
4
6
8
y-axis : 4-7 inside beam window
Figure 4-22: PGNAA method with 238Pu-Be source:
Photon dose distribution in thigh model
Five plots correspond to the dose given adipose
tissue 1, adipose tissue 3, muscle 1, muscle 3, and
femur.
10
PGNAA - Be(p,n) with BNL setup
Neutron Dose
S
....
....
....
...
.
.
..
4b
mcnp
01/21/98 08:27:31
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o
114
n
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= mc/padth
so
./
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energy
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time
1
t
,,
-------
a.t. 3
3
a.t.
---------------.
muscle 3
------ 2
6
4
y-axis : 4-7 inside beam window
8
10
Figure 4-23 :PGNAA method with 9Be(p,n) source and BNL set-ups:
Total neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1, adipose
tissue 3, muscle 1, muscle 3, and femur.
fem xr
t
PGNAA - Be (p, n)
with BNL setup
Thermal Neutron Dose
*e
.
S
.
.
. , . . , ,
,
~Ii
.
, |
,
.
4b
mcnp
I
01/21/98 08:27:31
114
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n
/
/\
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mc/lpadth
mctal
f
cell
d
flag/dir
u
user
s
segment
m
-_------------
/-/
mult
c
cosine
e
energy
t
time
\
a.t. 1
a.t. 3
muscle
muscle
femur
2
4
6
8
10
y-axis : 4-7 inside beam window
Figure 4-24: PGNAA method with 9 Be(p,n) source and BNL set-ups:
Activating neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1, adipose tissue
3, muscle 1, muscle 3, and femur.
81
PGNAA - Be(p,n) with BNL setup
Photon Dose
-
.
i
,
.
.
.
.
,
I
...
.
.
.
..I
.
.
,..
I
.
.
.
.
.
.
.
I
4b
mcnp
01/21/98 08:27:31
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h-----
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/
mctal = mc/padth
".
f
cell
d
flag/dir
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user
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/'
\
/\
,/~
/ '"
'.
/\
m
/\
/',
/'
_,l,/S
c
cosine
e
energy
t
time
a.t. 1
",-,
c--I;----
--
mult
.,,
a.t. 3
muscle I
muscle 2
----------femur
4
6
8
10
y-axis : 4-7 inside beam window
Figure 4-25: PGNAA method with 9Be(p,n) source and BNL set-ups:
Photon dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1, adipose tissue
3, muscle 1, muscle 3, and femur.
82
PGNAA - Be(p,n) with LABA setup
Neutron Dose
.
*
.
,
II
, . . . .i..
i
. . . . . . . . .
. . . .
. . .
mcnp
4b
03/30/98 19:00:30
tally
114
n
nps
20000000
bin normed
cmetal
;
o
ae
S-------
-
mc/adth2
*
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cell
d
flag/dir
1
u
user
1
a
segment
1
m
mult
1
c
cosine
1
e
energy
3
time
1
t
a.t. 1
-------
a.t. 3
muscle 1
muscle 3
--------2
4
8
6
10
y-axis : 4-7 inside beam window
Figure 4-26 : PGNAA method with 9Be(p,n) source and LABA set-ups:
Total neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1, adipose tissue
3, muscle 1, muscle 3, and femur.
83
-femur
t
PGNAA - Be(p,n) with LABA setup
Thermal Neutron Dose
4b
mcnp
03/30/98 19:00:30
114
tally
n
20000000
nps
/
/
bin normed
/
/
mctal = mc/adth2
I
------- ----
/
/
/
/
I
I
f
cell
d
flag/dir
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segment
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mult
c
cosine
e
energy
t
time
/
/
/
I
/
--
a.t. 1
------
\\%
a.t. 3
muscle
- %-\ %
,
.
I
.
,
.
.
.
.
-.-I
.-.--.--
-
- -.
- - .
- .-
.
. --. -- . ' .
.
.
.
.
.
.4
y-axis : 4-7 inside beam window
9
Figure 4-27 : PGNAA method with Be(p,n) source and LABA set-ups:
Thermal neutron dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1, adipose tissue
3, muscle 1, muscle 3, and femur.
84
muscle
femur
PGNAA - Be(p,n) with LABA setup
Photon Dose
....
*
4b
mcnp
..
.
03/30/98 19:00:30
414
tally
P
20000000
aps
bin normed
"
. _/
mectal = mc/adth2
\
/
/
/
/
/
f
cell
*
d
flag/dir
1
u
user
1
s
segment
1
mult
1
c
cosine
1
e
energy
1
t
time
1
S/
S
.-
/
a.t. 3
--------
a t. 3
muscle 1
------------- -2
8
6
10
y-axis : 4-7 inside beam window
Figure 4-28 : PGNAA method with 9Be(p,n) source and LABA set-ups:
Photon dose distribution in thigh model
Five plots correspond to the dose given adipose tissue 1,
adipose tissue 3, muscle 1, muscle 3, and femur.
85
muscle 3
femur
4.4.2 Sensitivity
As mentioned in 4.2.2, composite sensitivity is proportional to the product of the
reaction rate and the absolute detector efficiency. The variation of those two factors will
be also investigated.
4.4.2.1 Number of reactions
Reaction
Neutron source
D-T generator,
(BCL)
238 Pu-Be
(BNL)
4 MeV Be(p,n)
(LABA)
4 MeV Be(p,n)
(BNL)
14
N(n,2n) 3N
14 N(n,Y)s15N
14N(n,y) 15N
14N(n,y)15N
Average number of
reactions [#/g/mSv]
(error bar)
9.7418E1
(0.28 %)
1.3527E5
(0.49 %)
9.5823E5
(0.66 %)
2.1248E5
(0.53 %)
Uniformity
[%]
(# of data points)
41.40 ± 0.78
(36)
54.66 ± 1.21
(48)
103.58 + 1.56
(48)
70.78 ± 1.21
(48)
Table 4-15 : Average number of reactions and their uniformity in the thigh box
phantom measured for irradiation systems used at BCL, BNL and LABA and the
combination of the accelerator source and the BNL collimating system.
One of the factors which will affect the composite sensitivity is the occurring rate
of reaction: 14N(n,2n)13N or 14N(n,y)15N. The reaction rates which were normalized to the
amount of nitrogen and the incident dose are listed in Table 4-15. The irradiation time for
the (n,2n) method was assumed to be 20 minutes. The uniformity is expressed as the
coefficient of variation (standard deviation / mean value) of 48 data points in the
phantom. The variation of the reaction rate in the phantom for each system is shown in
Figure 4-13 - 4-16.
Neutron source
Average number of
reactions
[#/g/mSv] (error bar)
1.1755E5
(0.50 %)
1.8166E5
(0.52 %)
3.1216E5
Uniformity
14N(n,y) 15N
238Pu-Be
4.0 MeV p-Be
3.0 MeV p-Be
[%]
(48 data points)
57.84 ± 1.22
66 .02 ± 1.20
71 .35 ± 1.20
(0.53 %)
1.5 MeV d-Be
2.6 MeV d-Be
2.5 MeV p-Li
1.3502E5
(0.52 %)
1.6720E5
(0.51 %)
2.3025E5
(0.52 %)
66 .80 + 1.24
63 .86 + 1.20
66 .14 ± 1.19
Table 4-16 : Average number of 14 N(n,y)15N reactions and their
uniformity in the thigh box phantom measured for the irradiation
system used at BNL and the combination of the accelerator
source and the BNL collimating system. No D20 moderater.
Also the average numbers of 14N(n,y) 15N reactions in the phantom calculated
using the BNL collimating facilities are shown in Table 4-16. Simulated neutron sources
are a 238Pu-Be source and various types of accelerator sources. For comparison, all
simulations were performed without the 2.5 cm-D 2 0 moderator.
Number of (n,2n) reactions using a D-T generator (BCL)
E
150 C 100S50
E
20
z
-lO
-20
y-axis [cm]
x-axis [cm]
150 -
S100
50
E
1020
y-axis [cm]
o
E
-20
8850
20
x-axis [cm]
10
10
20
"00
-20
y-axis [cm]
x-axis [cm]
Figure 4-29 : (n,2n) method: distribution of number of 14N(n,2n)13N reactions in the phantom
The phantom was divided into three 5 cm-thick layers along the depth. From the upper to the
lower, the investigated region in the phantom gets closer to the D-T neutron generator source
used at BCL. The horizontal region examined in the simulation is 40 cm (x) x 30 cm (y). The
phantom was divided into 36 cells in total and the number of (n,2n) reactions was calculated
for each cell.
88
Number of (n,g) reactions using a 238Pu-Be source (BNL)
x 10 s
............
0
0
E10
z
-10
-10
-20
y-axis [cm]
C,)
uc2
0
x-axis [cm]
x 10s
''
E
L-0
"
0
zv.oc
a)
oc)
3
y-axis [cm]
C,,
cn2
x-axis [cm]
x 10 s
E
M
zz
C
u,2
o0
5.
Ea
01
E
-,
5
0
.........
.....
.
z8
-5
-5
-10
-20
y-axis [cm]
-10
1
x-axis [cm]
Figure 4-30 : PGNAA method with 23 8Pu-Be source:
distribution of number of 14 N(n,y) 15N reactions in the phantom
The phantom was divided into three 5 cm-thick layers along
the depth. From the upper to the lower, the investigated
region in the phantom gets closer to the 238Pu-Be neutron
source used at BNL. The horizontal region examined in the
simulation is 40 cm (x) x 20 cm (y). The phantom was
divided into 48 cells in total and the number of (n,y) reactions
was calculated for each cell.
89
10..
x1
x 10s Number of (n,g) reactions using a 4 MeV p-Be accelerator source (BNL)
>,
(14
Do
5
E
-5-
z
-1o -20
y-axis [cm]
x-axis [cm]
x 105
C',
E
"
o}
C
cc2
o10
E
zM
y-axis [cm]
x-axis [cm]
x 10 s
C
4
0
.0
E
z
-5
y-axis [cm]
-10
-20
-10
0
10
x-axis [cm]
Figure 4-31 : PGNAA method with 9 Be(p,n) source and BNL set-ups:
distribution of number of 14N(n,y)1sN reactions in the phantom
The phantom was divided into three 5 cm-thick layers along the depth.
From the upper to the lower, the investigated region in the phantom
gets closer to the neutron source. The horizontal region examined in
the simulation is40 cm (x) x 20 cm (y). The phantom was divided into
48 cells in total and the number of (n,y) reactions was calculated for
each cell.
xNumber of (n,g) reactions using a 4 MeV p-Be accelerator source (LABA)
x 10
(I,O3
U)
28
...............
.
...
........
..............: ....
:........
:......
0
.00
E
Z
-10
-10
-20
y-axis [cm]
Cn
E
02
't
x-axis [cm]
1" - i,
106...
cc
................
.
10
._10CD
E
z
y-axis [cm]
x-axis [cm]
Co
3.
7...
......
U)
. .
as
......
......
-10
x
0:
."10
E
z
10
0
-5
-10
-20
y-axis [cm]
[c0
x-axis [cm]
Figure 4-32: PGNAA method with 9Be(p,n) source and LABA set-ups:
distribution of number of 14 N(n,y)15 N reactions in the phantom
The phantom was divided into three 5 cm-thick layers along the depth. From the
upper to the lower, the investigated region in the phantom gets closer to the
neutron source. The horizontal region examined in the simulation is 40 cm (x) x
20 cm (y). The phantom was divided into 48 cells in total and the number of (ny)
reactions was calculated for each cell.
91
4.4.2.2. Detector efficiency
Figure 4-17, 18 show the distribution of the absolute detector efficiency within
the thigh phantom. These results were obtained by simulating counting systems of two
methods: those employed at BCL and BNL. Because the detectors are positioned below
the sample for the BCL counting system, the detector efficiency decreases with increasing
depth. The BNL system also shows higher detector efficiencies for the cells closer to the
detectors. Table 4-17 presents the average detector efficiency and its uniformity in the
thigh phantom for the counting geometries used for the (n,2n) and the PGNAA methods.
Facility
BCL - (n,2n) method
BNL - PGNAA method
Average detector
efficiency
[%]
9.07 ± 0.033
0.58 ± 0.002
Uniformity
[%]
(# of data points)
69.45 ± 0.74
(36)
20.57 + 1.73
(48)
Table 4-17 : Average detector efficiency and their uniformity in the thigh phantom for
nitrogen measurement counting systems used at BCL ((n,2n)) and BNL (PGNAA).
Thigh counting system at BCL - Detector efficiency: (n,2n) reaction
0.20.15
0o.o05
01- .
0
0
y-axis [cm]
0.2
10
20
-10
z-20
x-axis [cm]
S...................
y-axis (cm]
x-axis [cm]
-0.15-
0.05
. ..
0I
-20
Z
y-axis [cm)
-10
0
10
-20
x-axis [cm]
Figure 4-33.: (n,2n) method: detector efficiency distribution
From the upper to the lower, the investigated region inthe
phantom gets closer to the Nal(TI) detectors used at BCL.
The horizontal region examined in the simulation is 40 cm (x)
x 30 cm (y). The phantom was divided into 36 cells in total
and the detector efficiency for 511 keV y-rays was calculated
for each cell.
Nitrogen counting system at BNL - Detector efficiency : (n,g) reaction
C
0o
C0
10
...
.
20
10
x-10
0.01
-20
y-axis [cm]
~~~~~..... -
-
x-axis [cm]
-- i ' '
0.005
0
-10
C
B
4
-20
y-axis [cm)
x-axis [cm]
0.01
0.005
0
l
C
0
z
y-axis [cm]
-10
-20
x-axis [cm]
Figure 4-34: PGNAA method: detector efficiency distribution
From the upper to the lower figure, the investigated region in
the phantom gets closer to the neutron source. Two Nal(TI)
detectors are diagonally located in the plus y-region. The
horizontal region examined in the simulation is 40 cm (x)x20
cm (y). The phantom was divided into 36 cells in total and the
detector efficiency for 10.83 MeV y-rays was calculated for
each cell.
4.4.2.3. Sensitivity
The composite sensitivity was calculated as explained in 4.2.2. Those results are
shown in Table 4-18. The uniformity of sensitivity for the bilateral measurement is also
shown. To achieve the bilateral measurements for the (n,2n) method, the patient was
assumed to be irradiated by two identical D-T neutron generators and detected by
identical counting systems from both sides of the patient, although this measurement
method is not practical for the current system. The irradiation time and counting time
were both assumed to be 20 minutes. The distributions of sensitivity in the phantom for
unilateral measurements are presented in Figure 4-19 - 4-22.
Method and
neutron source
(n,2n) :
D-T generator
- prone irradiation
(n,2n) :
D-T generator
- supine irradiation
PGNAA :
2 38
Pu-Be
PGNAA :
4 MeV p-Be
Uniformity for
bilateral irradiation
[%]
(# of data points)
29.46 ± 3.27
(36)
Average composite
sensitivity for
unilateral irradiation
[#/g/mSv]
5.0355
(0.74 %)
Uniformity for
unilateral irradiation
[%]
(# of data points)
33.92 ± 2.47
(36)
8.4779
(0.86 %)
97.12 ± 1.82
(36)
107.25
(0.81 %)
42.34 ± 2.31
(48)
27.83 ± 4.17
(48)
191.34
(0.92 %)
56.15 ± 2.21
(48)
27.65 + 3.74
(48)
Table 4-18 : Average composite sensitivity and their uniformity
in the thigh box phantom measured for the nitrogen
measurement systems
Sensitivity distribution : (n,2n) method - prone irradiation
3
c-
' ,,
ii
I:
P-'
-10
.20
10
0
--
2
-A -
3
;
x-axis [cm]
Sensitivity distribution : (n,2n) method - prone irradiation
/9
L-31
'4-
-2
-31
A
-5
-15
5
15
y-axis [cm]
Figure 4-35 : (n,2n) method: distribution of composite
sensitivity in the phantom for prone position irradiation
Investigated regions get closer to the D-T generator as
'1' -> '3'.
'
Sensitivity distribution : (n,2n) method - supine irradiation
-A - -2
- 2
---
----
~---~
' '~-~~r-~~-'-
-10
-20
0
10
x-axis [cm]
Sensitivity distribution : (n,2n) method - supine irradiation
25
U)
EE20,
20
is-
15
o
E
10-
o
--
..-- ""
' '"""
""
" -Imm
0
-15
-5
5
15
y-axis [cm]
Figure 4-36 : (n,2n) method: distribution of composite
sensitivity in the phantom for supine position irradiation
Investigated regions get closer to the D-T generator as
'1' -+ '3'.
-
2
-- A -
3
Sensitivity distribution using a 238Pu-Be source : PGNAA
I
r
160 140120 100 -
U-r
-I-
- -
-
-
II
-
-m -
3
2
80 60 4020 S1
0
-20
x-axis [cm]
Sensitivity distribution using a 238Pu-Be source : PGNAA
-
180
160
140
120
100
-- -A-
80
60
40
20
0
-10
-5
0
5
10
y-axis [cm]
Figure 4-37 : PGNAA method with 238Pu-Be source:
distribution of composite sensitivity in the phantom
Investigated regions get closer to the source as '1' -+ '3'.
2
3
3
Sensitivity distribution using a 4 MeV p-Be accelerator source : PGNAA
350
300
250
I-'
200
U-'-
I-
150
--
2
--
3
-
2
---
3
100
0
x-axis [cm]
Sensitivity distribution using a 4 MeV p-Be accelerator source : PGNAA
350
360
- --
I
"
300-
--
250
"
200-
S /
150100
4r
50-
-10
-5
IIs
0
5
10
y-axis [cm]
Figure 4-38 : PGNAA method with 9Be(p,n) source and BNL system:
distribution of composite sensitivity in the phantom
Investigated regions get closer to the source as '1' -> '3'.
4.5 Summary
In this section MCNP simulations of nitrogen measurements were conducted
using various types of experimental systems. In the simulation the maximum skin dose
delivered to the subject and the sensitivity to the nitrogen detection were calculated for in
vivo measurement systems of the (n,2n) and the PGNAA methods. The measurement
geometries and characteristics for all simulations were described and the results of dose
and nitrogen sensitivity were presented. The investigation was made not only for the
nitrogen measurement facilities in operation, but also for hypothetical accelerator
systems, in order to examine the potential of using different types of neutron sources than
those presently used.
100
5. Comparison and conclusion
This chapter discusses the simulation results presented in chapter 4 and
demonstrates how experimental parameters determined by investigators, such as
irradiation and counting time, affect figures of merit of the two nitrogen measurement
systems: the BCL (n,2n) system and the BNL PGNAA facility. Possible improvements
which can be made on those systems are also described for future work.
5.1 Accuracy and reproducibility
Table 5-1 presents a comparison between the nitrogen net counts of both methods
which were calculated by MCNP simulation using a thigh box phantom (40 x 20 x 15
cm 3 ) in which 1.4414E25 nitrogen nuclei were contained. An irradiation time of 6
minutes and a counting time of 20 minutes were assumed for the (n,2n) method while a
200 sec-measurement was assumed for the PGNAA method.
Method and facility
BCL - (n,2n) method,
D-T generator
BNL - PGNAA method,
238Pu-Be
Neutron yield
[n/sec]
4.9891E7
2.0300E8
Nitrogen counts
[#] (error bar)
5.1801E3
(3.14 %)
1.9494E4
(0.64 %)
Table 5-1 : Nitrogen net counts for the (n,2n) method and the PGNAA
technique which were obtained by MCNP calculation.
As mentioned earlier, the background problem was excluded in the MCNP
simulation for both nitrogen measurement methods. However, the signal-to-background
ratio is a critical factor in evaluating accuracy and reproducibility of a measurement
system because the standard deviation of the counting results is calculated as
G +B
where G is the gross counts and B is the background.
Although the nitrogen counting for the (n,2n) method is performed in the whole
body counting room well-shielded with steel and lead, the nitrogen peak at the low energy
of 511 keV contains significant background contributions, which mainly come from
natural radiation sources such as cosmic ray, terrestrial radiation, and radioactivity of the
constituent materials of the measurement facility. Such high background in the low
energy region is shown in Figure 3-5. On the other hand, the contribution from natural
sources to the 10.83 MeV photopeak region for the PGNAA technique is negligible,
while the dominant background contribution from the 238Pu-Be radiation source itself
(due to the random summing of 4.4 MeV photons emitted from the 238Pu-Be source).
102
5.2 Irradiation and counting time
The relationships of the net counts of nitrogen and the composite sensitivity with
the measurement time can be obtained as shown in Figure 5-1, by use of following
equations derived in chapter 4.
Cn2n
6511
n2n=511 x
Sn2n
CPG
=10.83
SPG
=10.83
t )(e-t - e-
t2
x R 13(1 - e 2 to )(e-2t - e
mNn2n x dn2n x t o xA
t2
X e 13 X
e
13
-
(8)
3
X e1 5 x R15 x t'
(9)
(10)
x 1e 5 x R15(11)
mNPG x dPG
where C: nitrogen net count for the (n,2n) method ('n2n') and the PGNAA method ('PG')
S: composite sensitivity for the (n,2n) method ('n2n') and the PGNAA method
('PG')
s: detector efficiency at 511 keV and 10.83 MeV
e: emission probability of photons of radioactive products; '3 N ('13') and
15N('15')
R: reaction rate of 1'4N(n,2n)13N ('13') and 14N(n,y)' 5N ('15')
m: nitrogen amount
d: incident dose
X: decay constant of 13N
103
(n,2n) method at BCL
PGNAA at BNL
6000
180
160
5000
140
120-
S 4000
0
o
100
o
Ti 3000
80
60
40
22000
z
I
|
--* irradiation (c.t.=20 min)
p-C counting ( i.t.=20 min)
1000
!
u-
0
40
20
30
0
10
Irradiation / Counting time (i.t./ c.t.) [min]
10
20
30
40
Measurement time [min]
PGNAA at BNL
(n,2n) method at BCL
108.5
7,
6
108
P55
*
E
N
-4
,, . O
12
107.5
A
A
AA
A
107
/
0''
/c
co 2
1-
A
106.5 I
- * irradiation c.t.=20 min
-
counting (i.t.=20 min
106
0
10
20
30
40
Irradiation / Counting time (i.t./ c.t.) [min]
0
|
I
I
10
20
30
Measurement time [min]
Figure 5-1 : Relationship between measurement time
and figures of merit: nitrogen counts and sensitivity
Nitrogen net counts are related to the accuracy and
reproducibility of the system.
104
40
to: irradiation time for the (n,2n) method
tl: counting starting time for the (n,2n) method
t 2 : counting ending time for the (n,2n) method
t': measurement time for PGNAA method
The nitrogen counts are related to the accuracy and the reproducibility of the
system. To determine the irradiation time for the (n,2n) method, a compromise between
the net counts and the composite sensitivity needs to be made since they are related with
the irradiation time in the opposite way. On the other hand, the longer counting time
provides a greater gain in both the nitrogen counts and the composite sensitivity. It is
important, however, to remember that the body elements which have long half-lives could
present a major interference for the nitrogen measurement when the counting is
conducted for a long time. It is observed that the effects of the irradiation and counting
time above 20 minutes on these factors are not significant.
For the PGNAA method, the composite sensitivity is unaffected by the
measurement time while the nitrogen counts are proportionally related. One should note,
however, that the nitrogen background increases proportionally with the measurement
time as well and thus the nitrogen signal-to-background ratio is theoretically constant.
The PGNAA system at BNL employs a 400 second-bilateral irradiation for one section of
the body to obtain sufficient nitrogen counts for the estimation of nitrogen level in the
subject.
105
5.3 Dose
5.3.1 Incident dose
The equivalent dose of 0.7 mSv was calculated by the MCNP simulation for the
PGNAA method from the simulation of a 400 second-bilateral irradiation of a 15 cmthick phantom using the weighting factors recommended by ICRP. This result is close to
the 80 mrem reported by BNL researchers [42], although the method of calculating the
equivalent dose is not explained in detail.
As can be seen in Table 4-11, the MCNP simulation results showed that in vivo
nitrogen measurement by the (n,2n) technique delivers about half of the skin equivalent
dose delivered by the PGNAA method with the same irradiation time. The total
equivalent dose is proportionally dependent on the irradiation time. Therefore a roughly 6
minute-unilateral irradiation for the (n,2n) method delivers the total dose of same level as
the current PGNAA technique.
5.3.2 Dose distribution
Figure 4-17 - 22, 24, 25, 27, 28 showed the highest dose in femur among three
kinds of thigh tissues, confirming the fact that radiation is absorbed in denser tissues
more efficiently. The exceptions were total neutron dose for the accelerator source shown
in Figure 4-23, 26 which was most absorbed at the surface of the thighs. It can be
106
observed that the neutron spectrum of the 4 MeV 9Be(p,n) source is much softer than the
238Pu-Be
source or the 14 MeV monoenergetic fast neutron source from Figure 4-15.
These simulation results suggest that the relatively low energy neutrons (100 keV to 1
MeV) from the 4 MeV 9Be(p,n) source are attenuated primarily near the surface leaving
predominantly thermal neutrons to interact in the inner medium. On the other hand, fast
neutron components of the
238Pu-Be
source and the 14 MeV neutron source produce high
neutron flux within the subject which lead to high dose in condensed tissues.
It is observed from Table 4-14 that, excluding the highest dose in femur, the
(n,2n) system delivers higher fast neutron dose at the surface while the inner region is
more affected by thermal neutrons for the PGNAA method. Comparing the maximum and
minimum dose among three tissues in the thighs, the fast neutron dose decreased only by
a factor of 4 from the surface nearest to the neutron beam to the furthest for the (n,2n)
method, whereas the other three PGNAA systems showed 2 orders of magnitude
difference. Now it can be assumed that the variation of activating neutron dose correlates
with the activating neutron flux because the kerma factors is almost constant in the
energy range defined for activating neutrons. Therefore these results suggest that the
sensitivity in the human thighs is more uniform for the (n,2n) method than the PGNAA
technique for unilateral irradiation. The use of the (n,2n) method, therefore, may be more
suitable than the PGNAA method for absolute nitrogen measurement of human thighs.
Table 4-14 also shows that the dose from y-rays for the accelerator source relative
to total neutron dose is high compared to the other two fast neutron sources: the D-T
generator and the 238Pu-Be radionuclide. Since the neutron energy spectrum of the 4 MeV
107
9Be(p,n)
is much 'softer' than the other two sources, the high photon dose can be
considered due to secondary y-rays from neutron capture reactions arising from the
elements existing in the surrounding materials. Such thermal neutron reactions include 'H
(2.23 MeV),
209Bi
(0.32, 0.16, 4.17, 4.05 MeV), 7Li (2.03 MeV), and 'oB (4.43 MeV).
Photons in the energy range of 4 - 7 MeV may interfere with nitrogen measurement
because of random summing (See chapter 2.2).
Comparing the BCL (Figure 4-17 - 19) and BNL systems (Figure 4-20 - 25), a
superior collimating effect is obvious for the BNL PGNAA facility which has been
designed and developed for the irradiation of sectioned parts of a human subject. As
mentioned in chapter 3, the BCL system was originally developed for the in vivo
measurement of carbon, hydrogen and oxygen. In these measurements a patient is
scanned from the shoulders to the knees, where the collimating effect is not a determining
factor for the evaluation of the measurement system. Considering the partial body
irradiation for the thigh measurement, however, improvements need to be made for the
current (n,2n) collimating system in order to reduce dose to the rest of the patient.
108
5.4 Sensitivity
5.4.1 Supine- and prone-position irradiation for (n,2n) method
The supine-positioned irradiation showed about 70 % higher average sensitivity
than the prone-position irradiation (Table 4-18). This is because the nitrogen detection is
most sensitive in the region closest to the D-T generator and Nal(TI) detectors for the
supine measurement. These data result in about three times lower uniformity of detection
in the subject compared to the prone irradiation for unilateral measurement. The choice
between prone or supine irradiation should be made according to the purpose of the
measurement. When sequential measurements are intended rather than the absolute
determination of nitrogen, the magnitude of sensitivity may be more important than the
uniformity in order to acquire a desirable reproducibility of the measurements. However,
when the absolute amount of nitrogen in the subject needs to be measured, the uniformity
of sensitivity is a critical factor which should be fulfilled as well as the sensitivity yield.
5.4.2 Accelerator sources for PGNAA method
Table 4-16 showed that all of the accelerator neutron sources investigated in the
simulation provided higher yields of the number of 14N(n,2n)'SN reactions than the
2 38
pu-
Be source with comparable uniformity using same irradiation geometries. Particularly the
3 MeV 9Be(p,n) source presented about three times greater nitrogen activating effect than
109
the
238Pu-Be
source. Consequent calculations also showed higher sensitivity for the 4
MeV 9Be(p,n) accelerator source by a factor of 2 than the conventional source (Table 418). These results are very promising for the use of the LABA accelerator as a neutron
source for in vivo nitrogen measurement. Nonuniformity of the number of '4 N(n,y)1SN
reactions observed in the accelerator results is due to the relativaly large size of the
phantom compared to the size of the collimator/reflector assembly. Better uniformity of
the thermal neutron fluence in the subject may be achieved by using a moderator of
adequate thickness and/or modifying the size of collimating systems for human study.
5.4.3 (n,2n) method and PGNAA technique
Comparing the simulation data shown in Table 4-15 for the BCL and BNL
irradiation systems, the number of 14N(n,2n)15N reactions are less by 3 orders of
magnitude than
14N(n,y)' 5 N
with the incident dose of same level. Considering that the
number of reactions is proportional to the activating neutron fluence and the cross-section
of the reaction, it can be estimated that thermal neutrons delivered to the subject by the
BNL PGNAA system is by 2 orders of magnitude higher than fast neutrons delivered by
the BCL (n,2n) facility. However the (n,2n) counting system designed for the thigh
measurement provides about 20 times higher absolute detector efficiency than the
PGNAA system (Table 4-17), because of the low energy photons which will be more
efficiently detected. These simulation results provided about 10 - 20 times higher
110
sensitivity for the PGNAA method than the (n,2n) method, depending on the irradiation
position of the patient, as presented in Table 4-18.
Now it may be informative to examine the effects of the experimental parameters,
such as measurement times, which were determined by investigators. For the (n,2n)
method, those sensitivity results presented in chapter 4 for the 20 minute-irradiation and
the 20 minute-counting could be theoretically improved by shortening the irradiation time
and increasing the counting time, according to the relationships shown in Figure 5-1.
However the sensitivity increases only by a factor of 2.5 even with an irradiation time of
40 minutes and a counting time of 1 minute with considerable loss in the accuracy and the
reproducibility, which results in at the best 20 % of the sensitivity for the PGNAA
technique.
The hypothetical bilateral measurement by the BCL (n,2n) method showed
sensitivity uniformity comparable to the PGNAA method. The use of the (n,2n) method
for the absolute measurement of nitrogen may be possible and worth the effort to examine
the feasibility as carried out by some investigators in other (n,2n) facilities [11, 25, 26,
36].
5.5 Conclusion
In this thesis comparisons between two methods for in vivo nitrogen measurement
were made on the basis of MCNP simulation results, some of which were verified by
nitrogen measurements performed at the BCL (n,2n) facility. In nitrogen measurement
using the BCL system, only 4.1 % oxygen interference was found as opposed to the
results obtained by other investigators (- 20 %). Incident dose delivered to a patient per
unit irradiation time was about a factor of 2 higher for the PGNAA method employed at
BNL than the BCL (n,2n) method. The sensitivity results, however, showed that the
PGNAA method provides the nitrogen sensitivity yield which is 10 - 20 times higher than
that of the (n,2n) method, depending on the patient position during the irradiation. Prone
position irradiation for the (n,2n) method showed the excellent uniformity of sensitivity
within the subject for unilateral irradiation (34 %) compared to the PGNAA method (42
%). The simulations with the BNL irradiation system using the proton accelerator source
spectra showed high potential to employ the LABA accelerator as an alternative neutron
source for nitrogen measurement in vivo.
112
5.6 Future work
5.6.1 (n,2n) method
In order to improve sensitivity to nitrogen detection, some modifications can be
made with the current BCL measurement system. Although the counting system which
was constructed for the human thigh measurement provided significantly higher detector
efficiency compared to that of the PGNAA detector arrangement, the counting system
needs to be still more improved since the composite sensitivity is increased with the
detector efficiency. The detector efficiency is increased by using multiple detectors. The
use of two more NaI(TI) crystals identical to the two detectors currently employed is
under consideration, each of those to be located by the side of the patient's thighs. This
configuration is also expected to provide better uniformity of detector efficiency in the
subject.
For absolute determination of nitrogen further interference investigation should be
made. Although the oxygen interference was estimated to be much smaller than expected
by other investigators, the effect of differences in distribution of nitrogen and oxygen in
the human body needs to be still examined since a homogenous phantom would be used
to determine a calibration factor which converts the nitrogen level in the phantom to that
in the patient. Furthermore, the bilateral irradiation system is imperative for absolute
measurement although it seems difficult to install another D-T neutron generator above a
patient with bulky collimator/reflector material in addition to the present system.
113
5.6.2 PGNAA method
After modified and upgraded several times by dedicated researchers, there remains
very little which can be done to improve the performance characteristics of BNL PGNAA
facility. Although the better uniformity of the composite sensitivity is desired for the
thigh measurement, further modifications of the system, such as changing the thickness of
D2 0 moderator, would not necessarily improve the sensitivity uniformity for other parts
of the body. It should be noted, however, that the graphite collimator which defines the
neutron beam window may be a major interference for nitrogen measurement, because
4.43 MeV photons emitted by the inelastic reaction
12
C(n,n')' 2C (threshold = 4.8 MeV, a
= 0.2 b) cause random summing which may contribute to 10.83 MeV nitrogen
background. Despite the advantages of the low cost, relatively light weight, and good
moderating and reflecting properties (a high elastic and inelastic neutron cross section),
the use of graphite as a part of the nitrogen measurement system may need to be
reconsidered. The use of an accelerator as a neutron source for the PGNAA method may
be desirable not only for because sensitivity may be possible, as demonstrated in this
thesis, but also because a number of neutron spectra can be generated with neutron
energies below the threshold energy of the inelastic reactions of carbon.
114
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