Fast Neutron Cross Sections

In the neutron accounting procedure it is useful to treat reactions above 100 kev neutron energy separately from the rest. Usually the cross sections in this energy range are so small compared to the thermal and resonance energy cross sections that the fast reactions can be ignored insofar as any measurable effect on the zteutroh balance. An exception exists and this deals vith the fact that fast fission (predominantly in occurs at a sirP- 

Fast neutron cross sections are usually important forcnly materials inside the fuel element or in close proximity to the fuel element. The fission neutron energy spectrum is given by  

A compilation of fast neutron cross sections from Reference 3 which are useful in calculating the fast fission factor is given in Table 2,3 2.2.1. 

The eroes eeetione are defined ae follows ie the fission cross section et M the elastic transport cross section (the scattered neutron remains in the group) is the inelastie scattering cross section (the scattered neutron goes into group 0) 0 is the capture cross see tion and + t  

The only dominant resonance absorber in a reactor such as N is Resonance absorption also occturs in the other heavy isotopes but 

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As mentioned earlier, fission or fast neutrons five rise to threshold reactions of the type (n,p), (n,a), and (n,2n). With Z = atomic number, the first and the second lead to the production of respectively, Z-1 and Z-2 radionuclides, e.g., Fe(n,p) Mn, Cu(n,p) Ni, P(n,a) AI, or Fe(n,a) Cr. Although the cross sections (average fast neutron cross sections or of) for these reactions are low compared to those for thermal neutron activation, serious interferences may ensue. 

Since there is a lack of experimental data on neutron cross sections for Cm-241. fast-neutron cross sections were generated by means of nuclear systematics and nuclear models. The NK.VRREX and 2-PLUS" codes were used to generate the neutron cross sections that were then used to calculate the critical mass of Cm-244 as a function of density. Fig. 1. 

G.D. JOANOU and H. FENECH, Fast Neutron Cross Sections and Legendre Expansion Coefficients tor Oxygen-16, GA-3564, General Atomic (1962). 

It is understood that the sequence of n fast-fission generations was initiated by a single fast neutron produced by a thermal fission. With the function (P , and a suitable set of fast-neutron cross sections we can define several pertinent processes which describe alternate histories for each neutron appearing in the nth generation. These are ... 

Note that if the relation (10.217) is used to obtain 2i. then Eq. (10.218), along with a computed 2t can be used to determine 2,. The approach outlined here for estimating the fast-neutron cross sections is found to give reasonably good agreement with experiment in the case of U. Table 10.6 lists the various cross sections which have been used in the past to relate the theory to actual measurements on the fast effect. 

Sodium is used as a heat-transfer medium in primary and secondary cooling loops of Hquid-metal fast-breeder power reactors (5,155—157). Low neutron cross section, short half-life of the radioisotopes produced, low corrosiveness, low density, low viscosity, low melting point, high boiling point, high thermal conductivity, and low pressure make sodium systems attractive for this appHcation (40). 

A rough estimate of the critical radius of a homogeneous unreflected reactor may be obtained simply by estimating the neutron mean free path according to (14.6). Assuming metal with a density of 19 g cm and a fast fission cross section of 2 X crn, one obtains = 10 cm. A sphere with this radius weighs 80 ks. For an unreflected metal sphere containing 93.5% the correct value is 52 kg. Pu has the smallest unreflected critical size for Pu (5-phase, density 15.8 g cm ) it is 15.7 kg ( 6 kg reflected), and for 16.2 kg ( 6 kg reflected). 

One of its attractive features of rhenium is that it is a spectral shift absorber (SSA), which means that it has a low relative absorption cross section for fast neutrons while in the thermal spectrum its absorption cross section increases dramatically. This has safety applications for the reactor design in accident scenarios. Rhenium has an absorption cross section of in the fast spectrum, however the magnitude of the difference between the absorption cross section and the fast fission cross section of is low compared to the difference at a thermal spectrum. It also provides a barrier that protects Niobium 1% Zirconium from nitrogen attack and damage caused by other fission products that outgas from the fuel. Most of the other SSA materials have a relatively low melting point, making them less attractive. 

C. B. Mills, "Neutron Cross Sections for Fast and Intermediate Nuclear Reactors, LAMS-225S (Jan. 16, 1959). 

R. J. ARCHIBALD and D. R. MATHEWS, "The GAF/ GAR/GAND Fast Reactor Cross Section Preparation System Volume I, GAF/GAR - A Program for the Calculation of Neutron Spectra and Group-Averaged Cross Sections, GA-7542, Vol. I, Gulf General Atomic (1968). ... 

The fast fission factor e, the material buckling B, the quantity Ap/AB were measured in critical assemblies of each fuel mixture. The thermal-neutron diffusion area L wa.8 calculated from thermal-neutron cross sections. The delayed-neutron fraction was determined from the delayed-neutron fractions for U and measured by Keepin and the "U-to- U fission ratio measured in the critical assemblies. 

Calculations of InitegnH parameters in ten fast critical assemblies have bemi performed with several neutron cross-section sets to test the agreement with experimental integral values. These critical assemblies are listed in Table I in approximately decreasing order of spectrum hardness. They and their specifications for one-dimensioiml homogeneous calculations have been selected by Dav and Hess for the next phase of ENDF/B data testing. 

J. J. SCHMIDT, "Neutron Cross Sections for Fast Reactor Materials, Part 1 Evaluation, KFK-120, Kernforschungszentrum Karlsruhe (1966). 

T. A. PITTERLE et aL, "Calculations of Fast Critical Experiments Using ENDF/B Data and a Modified ENDF/B Data File," Second Conf, Neutron Cross Sections Technology, Washington, DC (March 4-7, 1968). 

The parameter 2 corresponds to 2 for the core. Both of these quantities are called the fast removal cross sections. Note that the form (c) explicitly assumes a nonmultiplying reflector material, since a zero source of fast neutrons has been taken in the reflector. As in the core, and 25, are thermal properties of the reflector calculated in the usual way ... 

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