Boron-Polyethylene Irradiation Containers for High Energy Neutron Activation Analysis W. A. Jester and K. N. Prasad Department of Nuclear Engineering, The Pennsylvania State University, University Park, Pa. 16802
A NUCLEAR REACTOR is an excellent intense source of high energy neutrons, i.e., uncollided fission neutrons. These neutrons can be used to induce (n,p), @,a), (n,2n), and (n,n') reactions, especially in the lower weight elements. This capability is not usually employed in thermal reactor activation analysis because of interference from (n,r> reactions. The traditional way of treating this problem is to surround the sample with cadmium, which is effectively black to thermal neutrons. However, this approach does not significantly remove resonance neutrons above the cadmium cut-off. Thus, elements such as manganese and cobalt can still become highly activated because of their high resonance cross sections. Also, if a pneumatic transfer system is employed for short half-life work, the sample carrier is usually a hydrocarbon material such as polyethylene which will moderate some of the high energy neutrons passing through the cadmium, adding to the (n,?) interference. In this work, two types of boronpolyethylene irradiation containers were constructed and tested to evaluate their effects in reducing (n,r) reactions. The boron-10 in the containers has a large l / V neutron capture cross section which extends well into the resonance region and, therefore, is effective both in reducing the resonance flux reaching a sample and in absorbing any neutrons thermalized in the polyethylene. The boron also does not become significantly activated so that pneumatic sample carriers made of this composite material can be safely handled. Fast Neutron Activation with Nuclear Reactors. The fissioning of the most common reactor fuel, uranium-235, results in a spectrum of neutrons with energies from below 0.1 MeV to about 20 MeV. The average energy of this fission spectrum is about 1.9 MeV and the most probable energy is about 0.65 MeV, with 6 6 z of the neutrons having energies between 0.5 and 3 MeV. The number of neutrons falls off exponentially with energy above 3 MeV. The intensity of the fast flux component of most research reactors utilized for activation analysis is in the range of 10" to 1014 neutrons/sec-cm2. The intensity and energy of these neutrons are enough to induce usable (n,p), (np), (n,2n), and (n,n') reactions in many elements. Some of the more interesting threshold reactions are listed in Table I for isotopes that are analyzed more easily this way than with thermal neutron reactions. The neutron flux spectrum contains two other components beside the fast neutron component (Figure 1) ( I ) . These are the resonance or epithermal component which consists of neutrons that have been partially slowed down, and the thermal component which consists of neutrons that have slowed down completely and are in thermal equilibrium with the moderator. The relative intensity of these three components varies with position in and around the reactor core, and with reactor types. It is primarily a function of the moderator. For the common light water research reactor, the thermal flux will be (1) F. F. Dyer, in "Guide to Activation Analysis," W. S. Lyon, Jr., Ed., D. Van Nostrand Company, Princeton, N. J., 1964, p 15.
Table I. Threshold Reactions of Interest Threshold, Cross-section, Reaction Half-life MeV barns" 14N(n,2n) 13N 10 min 10.6 10.2 1 . 4 x'10-6 16O(n,p)'5N 7 . 1 sec 4.2 5 x lo-' lgF(n,p)leO 29 sec 2 . 3 min 2.0 1.4 x lo-* 1P(n,a)2*A1 14 days 0.95 6 . 0 X 10-* azS(n,p) aP 46S~(n,a) 42K 12 hr 0.61 5 x 10-8 2.9 8.7 x 10-4 s6Fe(n,p)a5Mn 2 . 6 hr 47 days -0.30 2 x 10-6 zo3Tl(n,p)20 3Hg 204Pb(n,n1)2o4mPb 68 rnin 2.2 ... a Approximate fission flux averaged cross sections.
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NEUTRON ENERGY (ev) Figure 1. Reactor neutron energy distribution about a factor of three to five more intense than the fast flux. This difference in flux intensity combined with the fact that the (n,r) thermal reactions have cross sections which are usually larger by two or more orders of magnitude, usually results in the masking of any threshold activation products. In addition, the threshold reaction often produces the same activation product as an (n,r) reaction with a neighboring isotope. For example, phosphorus-32 is produced from both sulfur and phosphorus. Suppression of (n,?) Reactions Using Neutron Shields. The common method of reducing the amount of (n,r) reactions occurring in an irradiated sample is to surround the sample with a cadmium liner. The existence of a large resonance at 0.173 eV makes cadmium an effective neutron shield for thermal neutrons of energy below 0.5 eV (Figure 2). However, it is not effective in reducing the (n,-y) activation due to the epithermal neutrons. The (n,r) cross sections of all the elements have 1jV tails which extend into the epithermal region. In addition, many of them have large resonances in this region, Examples of two such cross-sections are shown in Figure 2.
ANALYTICAL CHEMISTRY, VOL. 43, NO. 3, MARCH 1971
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Figure 2. Neutron cross sections Some researchers have used boron in combination with cadmium to obtain suppression in both the thermal and epithermal ranges (2). As shown in Figure 3, boron-10 in natural boron has a large 1jV cross section that extends from the thermal well into the resonance range (3). Difficulties in Employing Neutron Shields. Cadmium metal in the form of thin sheets has found wide application as a thermal neutron shield. Most research reactors have cadmium-lined vertical tubes and pneumatic rabbit terminals for thermal neutron suppression. Boron unfortunately is not available in a form usable in a similar manner at such facilities. Another problem that is encountered when conducting activation analysis involving short-lived radioisotopes is related to the hydrogenous containers used in pneumatic transfer systems. Inserting container materials such as polyethylene into a shielded irradiation terminal will cause ther-
(2) D. C. Borg, R. E. Segel, P. Kienle, and L. Cambel, Znt. J. Appf. Rudiat. Zsotop. 11, 10 (1961). (3) D. DeSoete, R. Gijbels, and J. Hoste, “Proceedings, International Conference, Modern Trends in Activation Analysis,” Vol. 11, National Bureau of Standards, Washington, D. C., 1968, pp 699-750.
Figure 3. Neutron crow sections
482
ANALYTICAL CHEMISTRY, VOL. 43, NO. 3, MARCH 1971
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Table 11. Some Common ( o n ) Interference Reactions Cross-section, Reaction Half-life barns” aaNa(n,y)*4Na 15 hr 5.3 x 10-1 2.3 min 2.3 x 10-l a’Al(n,y)a*Al 1.9 x 10-1 * 1P(n,y)aaP 14 days ~ l ( n , yW ) l 37 min 4.3 x 10-1 4OA1(n,y)~lAr 1.8 hr 6.3 x 10-l 2.6hr 1.3 X 101 36Mn(n,ypMn 4.5 x 100 Wu(n,y)64Cu 13 hr Approximate cross section at 0.025 eV.
malization of fast neutrons within the shielded region, considerably increasing the (n,?) reactions. The authors encountered these problems in two projects attempting to utilize threshold reactions for activation analysis using the Penn State TRIGA Reactor. One case involved the pulsed neutron activation analysis for oxygen (4) using the lBO(n,p)lBN threshold reaction (Table I) to measure the oxygen content of minute amounts of rare earth oxides. In this case, the 6.13 and 7.12 MeV gamma rays of nitrogen-16 are well above those of almost all other activation products. The problem encountered was that even using the rabbit terminal with a 20-mil cadmium liner, enough (n,?) activation products were produced to cause such an intense sample activity, that high analyzer dead time and considerable random summing of detector pulses up into the range of the nitrogen-16 peaks caused unacceptable errors in this analysis. A second case involved the analysis for the protein in meat and leaves by determining the nitrogen content using the 14N(n,2n)13Nthreshold reaction (Table I). The 0.511 MeV annihilation photon was to be used for this analysis, but the amount of interference from the 6JMn(n,y)66Mnreaction (Table 11) made this technique unusable. 65Mnhas a high (n,?) cross section in the thermal as well as resonance regions. Even using the cadmium lined rabbit terminal, a satisfactory analysis could not be obtained because of this interference. Polyethylene-Boron Rabbits. In searching for a solution for the problems of resonance activation and thermalization of fast neutrons by the hydrogen in the rabbits, it was noted that polyethylene bricks containing boron were available on the market at a relatively low price. These bricks are sold as neutron shielding materials. It was decided to purchase (4) W. F. Naughton and W. A. Jester, ibid., Vol. I, pp 490-494.
Table 111. Induced Activities (Relative) under Various Neutron Shielding Conditions Cd 5%'B 30xB liner, Cd Cd Reaction Unshielded Basis liner liner No Strong Resonances 27Al(n,y)28Al .,. I 0.44 0.12 ~ ~ P ( I I , ~ ) ~ ~ P22 1 ... 0.23 63C~(n,y)~~Cu 27 1 0.49 0.28 Strong Resonances 1 0.79 0.37 sbM~~(n,y)~~Mn . . . 33 1 0.74 0.34 6QCo(n,y)60Co
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Table IV. Relative Induced Phosphorus-32 Activities for Equal Weights of Phosphorus and Sulfur under Three Different Shielding Conditions
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