f
-
.
#
I OF I
ORNL
P. 2914
.
:
:
:
.
EEEFSEFE
14:25 1.1.4 LE
!
MICROCOPY RESOLUTION TEST CHART
NATIONAL BUREAU OF STANDARDS -1963
.
ORNL-P. 2914
MASTER
MAR 2 1 1957
NEUTRON CROSS SECTIONS AND REACTION PRODUCTS FOR H, C, N, ANDO
FOR THE ENERGY RANGE FROM THERMAL TO 15 MEV*
CISTI 22:0SS
J. A. Auxier and in. D. Brown
Health Physics Division
Oak Ridge National Laboratory
Oak Ridge, Tennessee
H.C.,3.00; MN 65
Abstract--The accurwte calculation of neutron dose must be based on
definitive cross sections and a precise knowledge of the reaction products
in tissue. Although there are still several uncertainties in these parame-
ters, a compilation has been made of the most detailed cross-section data
available and reaction products for the four major elements in tissue
(i.e., H, C, N, and 0). The compilation is for neutron energies below
15 MeV, but the energy interval requiring the most study and analysis was
that from 2.5 to 15 MeV. Particular attention was directed to the non-
elastic reactions (e.g., the C(n,n') 3a reaction). Average values for
the energies of the various charged particles as a function of the energy
of the incident neutron have been computed.
These values were compiled to
provide a basis for revision of the dose-distribution functions for
neutron exposures of man and of animals used in radiobiological studies.
An analysis of the results of various measurements are compared with
calculated values based on these cross sections and with the values listed
in NBS Handbook 63.
*Research sponsored by the U. S. Atomic Energy Commission under contract
with the Union Carbide Corporation.
LEGAL NOTICE
RILLASED FOR ANNOUNCEINT
I WILEAR SCIDCS ABSTRACIS
This report mo prepared u an account of Goverament sponsored work. Neither the United
Suator, por the Commission, por Lay person acting on behalf of the Commission:
A. Makes any warrunty or representation, expressed or implied, with rospect to the accu-
racy, completeness, or ustulness of the information contained in this report, or that the use
of any information, apparatus, method, or procons dincloned in this roport may not infringe
privately owned righta; or
B. Asnmos any liabilities with respect to the use of, or for damage resulting from the
use of any information, apparatus, method, or procon disclosed in this report.
An und in the above, "person acting on behall of the Commirstoo" includes was a.
ployee or coatructor of the Commission, or employne of much contractor, to the extent that
such employee or contractor of the Commisolon, or employee of such contractor preparou,
dienominatos, or provides acco.. to, way Information pursuant to dia employment or coatract
with the Commission, cr his employment with such contractor.

TEXT
Any absolute measurement of neutron fluence or any calculation or
"dose" from fluence requires & knowledge of neutron cross sections for
the materials of interest. For some applications, only an activation
cross sectior is required, and in some instances only elastic cross
sections are needed. However, the health physicist must have available
the best cross sections for many types of reactior.s because his interests
and activities ar; so broad. Of particular interest are the cross sections
of the principal elements in tissue: hydrogen, carbon, nitrogen, and
oxygen, for neutrons. The values for the thermal neutron cross sections
for these materials have roc changed greatly during the past ten years,
and those for the region below 2.5 MeV have changed in a few instances
only. For higher energies, the changes have been greater, for the neutron
energy span from abcut 6 to 14 Mev, many uncertairties have beeri encountered
and many changes made in recent years and until the present. This has
resulted in frequent changes in accepted values at most laboratories.
Similarly, there have been considerable uncertainties in the reaction
products, both in terms of particle or quantum types) and in values
for the energy released by the specific reaction products; Q values (2)
have been known generally for about a decade. The formation of nuclear
data information centers such as those at Oak Ridge National Laboratory
and at Brookhaven National Laboratory has made the compilations easier;
in the future, they may provide most of the information in the form needed.
However, due to the special nature of the dosimetry requirements, the
information presented here was taken from many sources and averages and
SRT
interpolations made as reqrired for a computation of absorbed dose in the
energy range below 15 Mel'. For neut con energies of 2.5 to 15 MeV, a
smooth curve based on extrapolation between the available data points is
gonorally a good approximation of the cross-section curves for dosj.metry
applications. Figure 1 shows the elastic scattering cross section
of
the four principal elements in tissue for the energy range below 14 Mov.
The relative macroscopic cross section has the advantage here that the
relative number of interactions with each element can be seen readily.
The relative contribution of the C, N, and O recoils to dose equivalent
is greater than för absorved dose because of the higher average values
of LET (i.e., greater QF's). The elastic cross sections are generally
decreasing functions of neutron energy to 15 MeV.
Two of the cross sections for neutrons of thermal and near-thermal
energies are shown in Fig. 2. These cross sections, decreasing with
neutron energy as 1/V, have been accepted generally for some time. The
only important reactions are the H(n,r) {H and the 14Ncn,p) 'c reactions
[e.g., all other (n,r) reactions total about 0.5% of the H(n,r)reaction).
All significant nonelastic cross sections for neutrons incident on
tissue are shown in Fig. 3." These are relative macroscopic values with
all proton-producing reactions, all alpha-producing reactionis, etc., summed.
These threshold reactions are generally increasing functions of energy to
15 MeV. The sum of the macroscopic (n,r) reaction cross sections are
particle reactions which yield deexcitation gamma rays (e.8., 12C(n,n')?2C*).
This curve is not the sum of the other curves because some charged particle
reactions do not yield gamma rays (e.g., the 12C(n,n') 30). Sone of the
reactions (e.8., 14N(n,p) 4C] yield greater energy for local depusition
than the incident neutron (i.e., they have positive values of ).
Cross sections for reactions which canse alpha-particle emission from
oxygen ruclei are shown in Fig. 4.3013) Although these general values are
convenient for estimating dose or dose equivalent, an extensive set of
cross sections are required for detailed calculations; the detailed non-
elastic and inelastic cross sections are given in Figs. 5 thru 9. (3-13)
Figure 10 shows a list of the most important charged particle reactions by
neutrons in tissue. The maxiam ranges for the charged particles for 7
and 14 MeV neutrons are shown as are the accepted Q values. The equation
used to compute the average cnergy deposited by · charged particle is
given below. This equation is based on an assumed particle emission with
equally probable energies between zero and maximum.
M3 Q + Ej (M3 - M)
2 M1 M2 E,
F2 = - -
(M2 + M3)2
M2
M2 + M3
where
Mi = mass of incident particle,
M2 = mass of reaction product with energy E2,
M3 - mass of reaction product with energy Ez - Ej + Q - E2,
Q: Q value for reaction,
Ej = energy of incident particle, and
E2 - average energy of reaction product with mass My.
A more comprehensive presentation and analysis of the calculational
results will be presented in a paper by Jones et al. (14)
-
.-
-.-
REIERENCES
1. F. Aj zenberg-Se love and T. Lauritsen, Nucl. Phys. 11, 1 (1959).
2. F. Evorling et al., 1960 Nuclear Data Tables, Parts 1 and 2, Nuclear
Data Project, NAS-NRC (Edited by K. Way) (1951).
3. D. J. Hughes and R. B. Schwartz, Noutron Cross Sections, Brookhaven
National Laboratory Report BNL-325, 2nd Edition (1958).
4. D. J. Hughes, B. A. Maguro, and M. K. Brussel, Neutron Cross Sections,
Brookhaven National Lahoratory Report BNL-325, 2nd Edition, Supplement i
(1960).
5. J. R. Stehr. et al., Neutron Cross Sections, Brookhaven National
Laboratory Report BNL-325, 2nd Edition, Supplement 2, Volume 1 (1969).
6. J. R. Smith, Phys. Rev. 95, 730 (1954).
7. A. B. Lillis, Phys. Rev. 87, 716 (1952).
8. W. J. McDonald et al., Nucl. Phys. 75, 353 (1966).
9. V. V. Nefedov et al., Soviet Progress in Neutron Physics (Edited by
P. A. Krupchitskii), p. 241-247 (1961).
10. R. A. AL-Kital and R. A. Peck, Jr., Phys. Rev. 130, 1500 (1963).
11. V. E. Scherrer, R. B. Theus, and W. R. Faust, Phys. Rev. 21, 1476
(1953).
12. J. P. Conner, Phys. Rev. 89, 712 (1953).
13. J. B. Singletary and D. E. Wood, Phys. Rev. 114, 1595 (1959).
14. T. D. Jones et al., "Dose Distribution Functions for Neutrons and
w
Gama Rays in Anthropomorphous and Radiobiological Phantoms," (To be
published).
PIGURE CAPTIONS
Fig. 1. Macroscopic elastic scattering cross sections for the H, O, C,
and N in tissue for the nout ron energy range from thermal to 14 Mev.
Pig. 2. Macroscopic capture cross sections for the H and N in tissuo for
the thermal and near-thornal nout ron energy range.
Fig. 3. Total macroscopic nonelastic cross sections for tissue for the
energy range from thermal to 14 Mev.
Fig. 4. Total oxyger. cross sections for alpha-producing reactions in tissue
for the neutron energy range from 3 MeV (below the lowest threshold) to
14 MeV.
Fig. 5. Microscopic cross sections for inelastic and nonelastic reactions
in carbon-12 for the neutroit energy range from 5 MeV to 14 MeV.
Fig. 6. Microscopic cross sections for the three reactions, 14N(n,p)?"c,
14N(n,20)1B, and 14N(n.ws)*18*, for the energy range from thermal to
14 MeV.
Fig. 7. Microscopic nonelastic cross sections for the three reactions,
14N(n,az) 11B*, "N(n,a3) '18*, and 14N(n,t)12C, for the neutron energy range
from 5 Mori to 14 MeV.
Fig. 8. Microscopic inelastic cross sections for five (n,n') and one (n, 2n)
reactions in nitrogen-14 for the energy range from 4.75 MeV to 14 Mev.
Fig. 9. Microscopic inelastic and nonslastic cross sections for oxygen-16
for the neutron energy range from 3 MeV to 14 MeV.
Fig. 10. Charged particle reactions in tissue with maximum ranges for 14 MeV
· and 7 MeV noutrons.

ORNL-CWG. 66-8261
NEUTRON ELASTIC SCATTERING
CROSS SECTIONS IN TISSUE
SCATTERING CROSS SECTION X RELATIVE ABUNDANCE
BARNS /qm
H
o
į
ķ à
.
8
9
10
11
12
E, (MeV)

e-
-
-
-
ORML-DWG. 66-8260
NEUTRON CROSS SECTION IN TISSUE IN LOW ENERGY RANGE
TOTAL FOR TISSUE
C'H in,y) H IN TISSUE
CAPTURE CROSS SECTION X RELATIVE ABUNDANCE
BARNS /gm
_AN (n,p)' 4C IN TISSUE
0.01
1.0
0.1
En le!)
ORNL-OWG. 66-3337

NON-ELASTIC CROSS SECTIONS FOR TISSUE
NOI-ELASTIC CROSS SECTION X RELATIVE ABUNDANCE
BARNS /gm.
-.
gorri
wie weet
DI
10%
6h 56 50 2T
:
Fg (MeV)

-
-
-
-
-
-
os
ORNL-OWO. 66-8338
OXYGEN ALPHA PRODUCING CROSS SECTIONS
FOR TISSUE
nouo
OXYGEN On, a CROSS SECTION X RELATIVE ABUNDANCE
BARNS / gm
ENERGY (MeV)

ORNCOWG, 66-12651
SAE
5000
o ctn,nic Es = 4.43 MeV
a rcno) Be
③ Poln,or) Be* E = 1.75 Mey
④ Pc(na Be* Be Heta
6 ctn,wee* tect Beta
© Pen, E - 6.8 Mey
He c+n
Betata
4ool
(mb)
300 아
​200아
​.
6
7
0
9
12
50
E(MeV)
ORNEWG, 66-12650
아
​50
0 NCnp 2
② "Nn,ay"B
③ Nn,ay"B*
Q = 0.63 Mey
Q :-0.16 Mey
E- 2.14 Mey
4000
(mb) |
3000
2000
a
.
100
2.0
30
40
50
60
80
90
100
10
120
13.0
14.0

70
E(MeV)

ORN-OWG. 66-12647
1200-
11 아
​0 wn, ce"B* Ey - 4.46 Mev
② *N(n,a) E : 5.0 Mey
3 'Non, t)'?C 0 =- 4.0 MeV
100
90
8 아
​(mb)
4우
​30
2우
​60
6.0
7.0
8.0
E, (Mev)
-
-
-

ORNL-DWG. 66-12648
L
110
O N (n,n' MN* Ey = 1.63 MeV
@ "Nin,n's Mw* Ey = 2.31 MeV
B inin,n') in* E, -5.1 MeV
MN(n,n'lN* Ey S 10.0 MeV
"Nin, n'OMN* Ey 11.0 MeV
AN(n,an)'w* 0 :- 10.6 MeV
.
.
ܕܢܚܫ
120
E(MeV)

ORNL-DWG. 66-12646
300
275
250
0 D(n,ngayo Ey = 6.1 MeV
O Moinning - 7.0 MeV
Broin,ngly * Ey = 3.8 Mev
'D(n,r) * : 4.8 Mev
visin, et
601n,ają : 3.1 MeV
Odin, a 14 • 3.8 MeV
8 Oin, a 14 Eye 7.0 MeV
ordin, plan
25
:00
OM
E(MeV)
-
-
--
--
-
-
-
-
-
--
-
-
ORNL DWG. 66-8310
Charged Particle Reactions in Tissue with Maximum Ranges
for 14 MeV and 7 MeV Neutrons
E
= 14 MeV
E = 7 MeV
Q Value
(MeV)
max
Reaction
"max
(MeV)
max
(cm)
(Mev)
(cm)
-5.70
7.896
.5 x 10-3
1.28
3.8 x 10-4
0.63
7.62
-4.01
2.98
12C(n,a)9Be
14N(n,p) 140
14N(n,t)120
14N(n,a)11B
160(n,p) 160
160.(n,d)15N
160(n,a)13C
6.90 x 10-2
6.3 x 10-3
4.5 x 10-3
6.30
-0.16
-9.63
-9.90
14.6
9.67
12.8
4.19
4.03
11.04
2.23 x 101
4.5 x 10-2
1.5 x 10-2
2.38 x 10-2
1.34 x 10-2
1.17 x 10-2
0
0
-2.21
4.57
2.6 x 10-3
.
END


DATE FILMED
4/ 27 / 67