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In the present paper we describe a target for the "Fakel" linear electron accelerator that serves as a source of neutrons for neutron-spectrometric investigations, conducted by the time-of-flight method in a wide energy interval from thermal energies up to several MeV. The neutrons are produced in a photonuclear reaction, produced in the target by bremsstrahlung formed in the same target by the accelerated electrons. The target is designed to work with electrons with energy up to 100 MeV and an average electron-beam current of up to 100/~A. A moderator (water) is used to increase the intensity of the resonance neutrons. To maximize the ratio of the flux of moderated neutrons to the dispersion of the time of flight of the neutrons from the moderator, characterizing the quality of the source in time-of-flight experiments, the neutron yield from the target must be maximized and the volume of the target must be minimized.The maximum neutron yield is achieved by using a converter material with a large nuclear charge Z. This is due to the fact that, first of all, the bremsstrahlung energy in a thick target (the thickness is greater than the radiation length) for high electron energies is proportional to ZE e, where E e is the energy of an electron [1]. Second, the critical electron energy (in MeV) above which the radiation losses predominate over ionization losses is -800/Z [2]. In addition, the position of the giant-resonance peak, characterizing the photoabsorption of 3,-rays, is displaced into the region of lower energy for heavier nuclei [3]; the photoabsorption cross section, integrated over the energy in the region of the giant resonance, is proportional to NZ/A, where N is the number of neutrons and A is the mass number [4]. A uranium target is preferable. It should be noted that the use of fissioning material increases the neutron yield from the target due to photofission. As one can see from Fig. 1 [5], the photofission cross section for 238U is comparable in order of magnitude to the cross section of (-g, n) and (3', 2n) reactions. The yield of neutrons per 50.5 MeV incident electron, which are produced in a 238U target with thicknesses g/cm 2 (cm) of 24.3 (1.25), 48.6 (2.5), and 72.9 (3.8) is equal to (%) 2.25 5: 0.18, 2.47 5: 0.22, and 2

At the present time 238pu is finding application as a source of energy in medicine and space technology. To determine the rate of accumulation of 238pu in the reactor fuel and to take account of burnup, it is necessary to know, besides the cross sections of the reactions (n, 3') and (n, 2n), the fission cross section of 238pu in the thermal and resonance neutron-energy ranges. The fission cross section of 238pu on a thermal spectrum (trf) and the resonance integral (If) were measured in Refs. 1-5. The data from ~ese works are presented in Table i. As one can see from the table, af and If exhibit a large variance, and in addition If exhibits large measurement errors.In the present work 238pu was obtained from the c~-decay of 242Cm, from which plutonium was removed beforehand. The target, prepared by spraying in vacuum, contained 2.10 + 0.01 /zg of 238pu and 10.6 + 0.5 ng of 239pu. The impurity 239pu was determined from the area of the resonance at 0.3 eV in measurements, performed with this target, of the energy dependence of the fission cross section using a spectrometer according to the moderation time of the neutrons in lead [7]. The 238pu target was irradiated in the thermal column of a F-1 reactor at the Russian Science Center "Kurchatov Institute." The neutron flux density was equal to 1.6.107 sec-l.cm -2 with a cadmium ratio of 600 with respect to 235U (n, f). The fission resonance integral was measured in the horizontal channel of the F-1 reactor with flux density (I/E) of the resonance neutrons equal to 1.6.107 sec-l.cm -2 with a cadmium ratio of 14 with respect to 55Mn and 63Cu and 30 with respect to 235U (n, 0.The cadmium screen was I mm thick. We employed as reference targets a 238pu target with a mass of 38.3 5:03 ng and a 235U target with a mass of 260 + 13 ng. The mass of the 235U target was determined by the method of backscattering of c~-particles. The fission fragments were detected with mica detectors [8]. The reference values of af and If for 239pu and 235U were taken from Ref. 9: the fission cross section for thermal neutrons with a Maxwellian spectrum is trf(239pu) = af~ = 780.8 + 1.8, where trf is the fission cross section for neutrons with E n = 0.0253 eV; go is the Westcott factor at 290 K; If(235U) = 280b barns. Finally, we obtained for 238pu af = 16.7 + 0.8 barns and If = 26.3 + 1.5 barns (the cutoff energy is 0.5 eV). As one can see by comparing the measured values with the tabulated data, in the present work the error in the determination of lf(238pu) has been substantially reduced.