Unmoderated neutron

The figures in Table III refer to unmoderated neutrons. Sources con-... 

From a measured value of R(, O Eq. (57.9) allows one to calculate the mean reaction cross section for the respective nuclide. Correspondingly, one may determine the integral of the unmoderated neutron flux cj)fi if the mean reaction cross section is known. Values of fl varies as a function of the location in the reactor core (irradiation position) but is constant for a well-defined location at a given reactor power. 

As pointed out above, the energy distribution of the neutrons in a reactor may be described by the sum of three components belonging to thermal, epithermal, and unmoderated neutrons. The reactor is characterized by the different contributions of these three components. 

In addition, flux hardening (preferential absorption of low-energy neutrons) and seif-moderation (unmoderated neutrons are further moderated inside the sample) may occur. Since the correction is difficult to apply for samples with mixed composition and irregular shape, the effect is often avoided by using as small a sample and a. standard as possible or by diluting the sample with a material having a low absorption cross section (graphite. cellulose). An internal standard can also be applied. 

A simplified illustration of neutron gauging is shown in Fig 2. The essential components are an isotopic neutron source, the sample to be measured, a thermal neutron detector and appropriate nuclear counting instrumentation to record changes in thermal neutron count rates. Neutron gauging can be performed in two distinct modes. If a bare or unmoderated isotopic... 

Fast breeder reactors are not operated, as e.g. light-water reactors, with slow neutrons, but with unmoderated fast neutrons as they occur immediately upon nuclear fission. These fast neutrons are necessary to sustain the chain reaction. The neutron yield per fission is here larger, since more neutrons are left over for the breeding process, once the neutrons lost by absorption and leakage have been subtracted. They are absorbed by or which are... 

As with reactor sources targets are strongly contained and heavily shielded. Their biological shielding is similar to, but thicker than, that found on reactors ( 3.1.1.1). The extra shielding is required since most neutrons from spallation targets remain unmoderated and very penetrating. 

It is obvious that the neutron energy spectrum of a reactor plays an essential role. Figure 19.4 shows the prompt (unmoderated) fission neutron spectrum with 2 MeV. In a nuclear explosive device almost all fission is caused by fast neutrons. Nuclear reactors can be designed so that fission mainly occurs with fast neutrons or with slow neutrons (by moderating the neutrons to thermal energies before they encounter fuel). This leads to two different reactor concepts - the fast reactor and the thermal reactor. The approximate neutron spectra for both reactor types are shown in Figure 19.4. Because thermal reactors are more important at present, we discuss this type of reactors first. 

The function is shown in O Eig. 57.3 ( unmoderated fission spectrum ). These fast neutrons, generally, play no role for the usual (n,y)-reactions because the (n,y)-cross sections become very small at neutron energies above 100 keV. Instead, in this energy range (n,p)-, (n,a)-, (n,2n)-, and (n,n )-reactions become important. The latter are threshold reactions, i.e., their cross sections are zero up to a certain neutron energy (threshold energy) and are then a function of the neutron energy. 

The cross-section curves excitation functions and the curves of the differential products response functions. In measuring R( and assuming relation (4.6), one can measure effective cross sections ([Pg.2625]

A series of critical experiments using as many as 100 subcritical vessels of U-235-enriched uranyl nitrate solution has been performed with the units assembled in unreflected and unmoderated arrays. The problem of neutron interaction between units in storage arrays Is a complex one for which an adequate theory does not exist. It is intended that the data presented here will support analyses of such Interaction problems in general and will provide bases for safe storage and transportation of these particular vessels which are widety used in industry. 

A si rles of experiments with unreflected and unmoderated cylinders of enriched-uranium metal (93.15% U-23S) has been performed at the ORNL Critical Experiments Facility to determine the dependence of the prompt-neutron lifcitime on the cylinder dimensions. Five cylinders ranging in diameter from 17.77 to 38.09 cm and in height-to-diameter ratio from about 0.2 to 0.7 were assembled and their prompt-neutron decay crnistants measured at delayed critical by the Rossi-a technique. Prompt-neu-tnm lifetimes were obtained from the measured decay constants and effective delayed-neutron fractions calculated by transport theory. The average uranium density for each assembly was greater tium 18.7 g/cm. . 

The direct comparison of results from parasitic to. dedicated Irradiations for the unmoderated CH3CF3/C3F6 system has been shown in Figure 5 and Table III, The quantitative agreement between these data sets shows that no detectable kinetic differences resulted from the use of widely different fast neutron distributions. A less extensive comparison for the 275°K CgFs moderated H2/C3F6 mixture system also yielded identical results for the two modes of Irradiation. 

With bismuth, the external volume indicated in Ref. 10 was used. The molten-salt systems are calculated for 3-39 ft external volumes. tSame as unmoderated blanket fluid. tNeutrons absorbed per neutron absorbed in... 

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