Pulsed Source Method Research Articles

Remarks on the use of well-known reactivity measuring techniques by the pulsed-source method

A mistake in the usual derivation of the Gozani-method is pointed out and its effect is demonstrated. In an example the spatial bracketing feature of the Sjöstrand method is used to obtain a better estimate of the reactivity than by other methods.

Experimental study of neutron thermalization by the pulsed-source method

Experimental study of neutron thermalization by the pulsed-source method

Test of neutron diffusion theory in a medium with channels by the pulsed source method (Single channel in a moderating block)

Test of neutron diffusion theory in a medium with channels by the pulsed source method (Single channel in a moderating block)

Studies of Electron Impact Excitation, Negative Ion Formation, and Negative Ion-Molecule Reactions in Boron Trifluoride and Boron Trichloride

Electron attachment processes in BF3 and BCl3 have been studied with both electron swarm and electron beam techniques. Thermal electron attachment rates were determined by the drift-dwell-drift technique to be 〈5 × 105 sec−1 · torr−1 for BF3 and 9 × 107 sec−1 · torr−1 for BCl3. Beam studies showed that F−, F2−, and BF2− were produced from BF3 by electrons of energy near 11.5 eV while Cl− and Cl2− were produced in BCl3 near 1 eV. The SF6− threshold electron impact excitation spectrum of BF3 exhibited no structure, however, a number of peaks were seen in BCl3, the chief one being near 7.6 eV. A low energy peak was observed in BCl3 at ∼ 2.5 eV. The thermal energy SF6−* ion was found to react readily with both BF3 and BCl3, yielding BF4− and BCl3F−, respectively. BF4− was also produced through the reaction F2−/BF3+BF3→ BF4−+F. Thermal energy rate constants for these reactions determined by a pulsed source method were 1.8 × 10−9, 1.6 × 10−10, and 6.1 × 10−11 cm3 molecules−1 · sec−1 in the order above.

Pulsed-Source investigations-II. The effect of cylindrical void channels in a neutron diffusing medium

A series of consistent experimental measurements of the neutron leakage from cavity-containing media were made and interpreted in terms of effective diffusion parameters. The technique employed was the pulsed-source method, in which the diffusion coefficient is inferred from the asymptotic decay constant curve. The ranges of pertinent variables (e.g., size and shape of the medium, cavity size, cavity spacing) were made large enough to afford a sound basis for checking the adequacy of various theoretical treatments of the problem. The graphite assembly used enabled construction of solid graphite systems or lattices of cylindrical void channels with seven different combinations of channel spacing and diameter (void fractions ranging from 0·0371 to 0·3801). The experimental results are compared with the analytical results of Benoist (1964), Carter (1961), and Behrens (1949) and discussed in the light of the recent measurements of Deniz et al. (1968),

A study of polycrystalline systems with an internal, sinusoidally varying, thermal neutron source

The frequency dependence of the amplitude and phase angle of the detector response, for the case of an internal, periodic source of thermal neutrons in a moderating system, is studied in the diffusion approximation. It is shown that the information obtained from this experimental technique-called the method of characteristic frequency—is closely related to the decay parameters measured by the more familiar pulsed source method. A new parameter, α 0( B 2), is defined for the experiment, which is equivalent to the fundamental pulse parameter, λ 0, for small values of B 2 but which has the advantage of being uniquely determinable (that is independently of waiting time) when B 2 > B ORIT 2. The variation of α 0( B 2) over a large range of B 2 is computed for graphite, beryllium and beryllium oxide, and compared with theoretical and experimental pulsed source values of, λ 0.

Pulsed-Neutron Diffusion Parameters in Spherical Water Systems

The diffusion parameters of light water have been measured by the pulsed-source method. The neutron bursts were produced from a Texas Neutron Generator by pulsing the ion source and using the 3H (d, n) 4He reaction. Neutrons were injected into spherical volumes of H2O and the decay constants of the neutrons were determined by means of an enriched BF3 miniature proportional counter. Neutron lifetime measurements were performed on small and large water samples with values of the geometric buckling from 0.035 to 0.655 cm−2. A harmonic analysis was conducted for the large geometries, while the waiting time method was used for the smaller ones. In the harmonic analysis, it appeared that a detector in a sphere is more sensitive to neutron fluctuation with time than it would be in a rectangular or cylindrical system. The diffusion parameters, D0 and C, were determined by fitting the decay constants to the equations and , where and are the geometric and the corresponding transport buckling, respectively. The second fit gave a lower standard deviation of C than did the first fit.

A study of neutron diffusion in sintered beryllium oxide using the pulsed-source technique

The pulsed-source method has been used to study neutron diffusion in sintered beryllium oxide, the neutron-physics properties of which are not very well known. From this work the diffusion and nuclear constants for thermal neutrons have been obtained (the diffusion coefficient, diffusion length, transport length, effective transport cross section and absorption cross section, the diffusion time and a number of characteristics pertaining to the final stages of the neutron slowing down process).

Thermal neutron diffusion measurements in concrete by the pulsed source method

The diffusion lengths, diffusion coefficients and macroscopic absorption cross sections for thermal neutrons in various concretes have been determined using the pulsed source method. The transport cross section for the hydrogen in the bound water of the concrete has also been determined and is found to decrease with increasing combined water content. It is therefore not possible to select any single value of this cross section for use in neutron shielding calculations.

A study of neutron diffusion in sintered beryllium oxide by a pulse method

BeO, which is an excellent moderator material, possesses high heat resistance, has low absorption cross section, and provides for supplementary neutron multiplication by the Be/sup 9/(n,2n) reaction, is considered a very promising material for high-temperature power reactors. In view of the contradictory data reported in the literature concerning the neutron-diffusion parameters in the sintered material, further thermal neutron-diffusion studies were undertaken using the pulsed-source method. A specimen previously used for determining the slowing-down length (Atomnaya Energ., 13: 258(1962)) was employed in this study. The values of the diffusion coefficient, diffusion length, transport length, effective transport cross section, absorption cross section, and the diffusion time were calculated and plotted. The calculated time of thermalization, 120 mu sec, was considerably lower than the value reported in the literature; this might be due to the approximating assumptions used. (TTT)

Thermalization and Diffusion Parameters of Neutrons in Zirconium Hydride

The pulsed-source method is used to measure the thermal neutron diffusion parameters of Zircaloy-2 hydride, whose approximate composition is ZrH/ sub 1.7/. The asymptotic decay constant (zeroth energy eigenvalue) is measured for the buckling range 0.03 < B/sup 2/ < 0.39 cm/sup -2/ and is fitted by a polynomral containing terms to Bsup 8/. The first energy eigenvalue is measured over the range 0.05 < B/sup 2/ < 0.39 cm/sup 2/ and is used to estimate the thermalization time constant. In additron, the thermalization time constant is also determined from time-dependent transmission measurements wrth l/v absorbers. (auth)

A measurement of the neutron age in graphite by the pulsed source method

The ages of thermal neutrons from the D-D and D-T reactions have been measured in graphite using a pulsed source placed inside a prism. From the time dependence of the thermal neutron density the effective age of D-D neutrons, converted to graphite with a density of 1·6 g cm −3, was calculated to be {ce:inline-formula}τ eff = 355 ± 9cm 2 {/ce:inline-formula}.The slowing down of D-T neutrons can be expressed approximately in terms of two groups of neutrons: neutrons which have suffered a single inelastic collision only ( {ce:inline-formula}τ eff {/ce:inline-formula} = 600 cmz) and neutrons which have undergone several inelastic collisions ( {ce:inline-formula}τ eff {/ce:inline-formula}= 240 cmz). In determining the neutron age 'r, the relative contributions of the groups were taken as 0·65 and 0·25 respectively. A third group, representing neutrons which are slowed down by elastic collisions only, were ignored since their contribution is small (around 0·1) and their age is large.

Thermal Neutron Absorption Cross Sections by the Pulsed Source Method

S>The thermal neutron absorption cross sections of 21 raturally occurring elements and B/sup 10/ were determined by companing the time deperdence of the neutron flux in water with the time dependence of the neutron flux in a water solution of the sample with the same geometric buckling. After making some small corrections anising largely from the change in the number of hydrogen atoms per cm/sup 3/ in the solution, the decay constant of the ratio curve gave the macroscopic absorption cross section averaged over the neutron flux spectrum. For a 1/ nu cross section, the 2200 m/sec cross section could be directly computed. For nonl/ mu cross sections, the effective 2200 m/sec cross section was obtained. (auth)