How to Calculate 
Gamma Radiation 
Induced in 
Materials 
Reactor 
Gamma activity induced in engineering materials, 
under conditions similar to those in the ORNL 
graphite reactor, can be estimated with this data. 
Thin specimens were irradiated in the reactor and 
component activities determined 
By C. D. BOPP and O. SISMAN 
Oak Ridge National Laboratory 
Oak Ridge, Tennessee 
MEASUREMENTS WERE MADE Of the 
gamma activity induced in a number of 
structural and reactor materials that 
were irradiated in the ORNL Graphite 
Reactor (1). From these data, a rea- 
sonably good estimate can be made of 
the activity induced in materials that 
are exposed under similar conditions. 
Table 1 has been constructed for this 
purpose. * 
Table 1 is directly applicable to re- 
actors of the same general class as the 
ORNL Graphite Reactor. For other 
types of reactors, one should recognize 
two limitations: 
1. The neutron flux in the Graphite 
Reactor is low enough so that only for 
very-high-cross-section materials is de- 
pletion of the parent element signifi- 
* The alternative method of estimating 
the activity is to calculate it from the chemi- 
cal composition of the material and the neu- 
tron cross section values from the litera- 
ture (3, 4). Except in the cases noted, 
activities calculated by the two methods are 
in good agreement. 
74 
cant, except for extremely long irradi- 
ation periods. 
2. Activation in the ORNL Graphite 
Reactor is primarily by thermal neu- 
trons. If materials are exposed to a 
higher ratio of fast-to-thermal neu- 
trons (2), (n,p) and (n,a) reactions— 
often more predominant with fast 
neutrons—are likely to give relatively 
greater contribution. 
Basis for Table 
The rate of change in the number of 
radioactive atoms per gram of ma- 
terial when burn-up of the parent ele- 
ment may be neglected, is given ap- 
proximately by 
dn/dt = ¢oN — nd (1) 
and the number of radioactive atoms 
present is 
n = (daN/d) (1 — e*) 
where n = number of radioactive 
atoms (atoms/gm), @ = neutron flux, 
(neutrons/cm?2/sec), N = number of 
TABLE 1—Induced Gamma 
Component 1 
Dee Et 
Material try (cm2/gm) (Mev) 
Iron alloys 
Armco Iron 2.5h 8 XK 1075 1.15 
Duriron 2.5h 1.2 X 107% 1.15 
1015 2.5h 1.4 % 105s) 505 
1030 2.5h 1.4% 1032 105 
1045 2.5h 1.7 X 105? ated 
6150 25h 1.456105) 15 
Nickel alloys 
Nickel 2.5h 2X 1075 1 
Hastelloy A 2.5h 1.0 X 10-3 15 
Hastelloy B 2.5h 1.4% 107? 1.15 
Hastelloy C 2.5h 324 $1052 115 
Hastelloy D 2.5h 124 XaOss eS 
Inconel 2.5h 373° 1034) M115 
Inconel X 2.5h 6X 1054 1.15 
K Monel 2-5h 2 10m 1.15 
Stainless steels 
Carpenter 20 2.5h 1.61073 1.15 
302 2.5h 1.3 X 107% 1.15 
303 2.5h 8 X 1074 1.15 
304 2.5h 8 X 1074 1.15 
309 2.5h 3X 10-2 1.15 
310 2.5h 3 X 1073 1.15 
316 2.5h 3.3 X 10% 1:15 
317 2.5h 4.2 X 107% 2.15 
347 2.5h 3 X 1073 1.15 
405 2.5h 1.0X 107% 1.15 
410 2.5h 9.3 X10 1.15 
414 2.5h 9.31074 1.15 
430 2.5h 7X10 = 1.15 
431 2.5h 1.0X10-* 1.15 
446 2,5 2.1 KalOnee eee 5 
502 2.5h 4.3% 10-4 1.15 
Aluminum alloys 
28 2.5h 2 X 10% 1.15 
38 2.00 1.2'>< 40m Sela: 
528 2.5h 6X10-6 1.15 
728 2.5h <10-5 1.15 
Identity of the Component Activities 
IRON ALLOYS 
The 2.5 hour component may be due 
both to the Fe®® (n,p)Mn** and the Mn‘5 
(n,y)Mn®* reactions. An upper limit on 
the (n,p) reaction is fixed by the lowest 
value of this component measured for any 
of the iron alloys—that for Armco iron. 
The high values for the other iron alloys 
are due to their manganese content. 
The 46-day component checks closely for 
all the iron alloys and is 50% higher than 
the value calculated for the Fe®§ (n,y) reac- 
tion. This difference may be due to error 
introduced by the use of the thermal neu- 
tron cross section as an approximation for 
neutrons with the energy distribution that 
is present in the reactor. 
