Fission-neutron Cross Section Research Articles

Area analysis of underdetermined neutron resonance data

A weighted maximum likelihood method is developed for estimating the parameters of the statistical distribution of capture widths (average and number of degrees of freedom) taking into account experimental errors. Results are obtained for 224Cm, 238U and 235U. A formalism is developed for the area analysis of capture and fission neutron cross sections using maximum likelihood methods for deducing the most likely values for the capture widths. This formalism is applied to the analysis of 244Cm resonance data.

Measurement of the Ratio of Fission to Capture Neutron Cross Sections in Uranium-235 in the 200-eV to 15-keV Energy Range

Values of the ratio of the capture to fission neutron cross section, α, of 235U have been determined in the energy range from 200 eV to 15 keV using a lead slowing down time spectrometer. A 235U fission chamber was used to determine the fission rate, and a gamma-ray proportional counter determined both the fission and the capture rates. Both detectors were calibrated in a thermal-neutron flux using the well-known thermal energy α values. Neutron and gamma-ray self-shielding in the samples and background counting rates have been carefully corrected. The resulting α values agree well with recent time-of-flight measurements in the energy range above 0.5 keV. Below 0.5-keV neutron energy, however, large discrepancies were observed. No obvious errors in the experimental method have been found to explain thesev discrepancies.

The preparation of carrier-free 129Cs, 132Cs and 136Cs

Carrier-free 129Cs was produced from enriched 130Ba by a combination of the ( p, 2 n), ( p, pn), and ( p, 2 p) reactions. An upper limit of about 8 mCi of 129Cs per hr could be produced from 100-mg 130Ba targets. A half-life study of the 129Cs product revealed a value of 33·3 ± 0·06 hr. Carrier-free 132Cs was produced in a nuclear reactor by the 132Ba( n, p) 132Cs reaction. The fission-neutron cross section for this reaction was determined to be 2·6 ± 0·4 mb. Cesium-136 was prepared in a nuclear reactor by the 136Ba( n, p) 136Cs and 139Cs and 139La( n, α) 136Cs reactions. The fission-neutron cross sections were determined to be 4·5 ± 0·9 and 0·4 ± 0·2 μb, respectively,, based on a value of 61 mb for the 54Fe( n, p) 54Mn reaction. Based on the results obtained in this study it is concluded that usable quantities of carrier-free 136Cs cannot be prepared in a nuclear reactor.

Measurement of the Ratios of Capture and Fission Neutron Cross Sections of 235U, 238U, and 239Pu at 130 to 1400 keV

The ratios σγ(238U)/σf/(235U), σγ(238U)/σf(239Pu), and σf (239Pu)/σf (235U) have been measured in the energy range 130 to 1400 keV. The activation technique was used to determine the capture rate i...

Preparation of carrier-free 55Fe and 59Fe by reactor irradiation

Millicurie quantities of carrier-free 55Fe and 59Fe were obtained from irradiated cobalt and nickel targets. The fission-neutron cross sections were determined to be 3.30 ± 0.39 mb for the 58Ni( n, α) 55Fe reaction, 1.06 ± 0.12 mb for the 59Co( n, p) 59Fe reaction, and 0.111 ± 0.013 mb for the 62Ni( n, α) 59Fe reaction based on a value of 61 mb for the 54Fe( n, p) 54Mn reaction. The uncertainties quoted are the standard errors at the 95% confidence level.

The preparation of 67Cu from 67Zn in a nuclear reactor

The preparation of 67Cu by the 67Zn( n, p) 67Cu reaction in a nuclear reactor was studied. Millicurie amounts were obtained from isotopically enriched 67Zn targets, and the fission-neutron cross section for the reaction was found to be 0·82 ± 0·04 mb. The 67Cu was separated from zinc and cobalt by a combination of anion exchange and chelate resin techniques.

Production of 69 mZn in the oak ridge research reactor

Production of 69 m Zn in the Oak Ridge Research Reactor by two nuclear reactions was studied using natural abundance Ga 2O 3 and isotopically enriched Ga 2O 3 and ZnO. The fission-neutron cross section for the 69Ga( n, p) 69 m Zn reaction was determined to be 0·496 ± 0.073 mb, and separation of the zinc products from gallium target was achieved by solvent extraction. Production of 69 Pm Zn reaction produced an average isotopic abundance of 2·22 mCi of 69 m Zn per mg of zinc target.

The 127I(n, p) 127mTe; 127I(n, p) 127Te and 127I(n, 2n) 126I Mean Fission Neutron Cross Sections

Summary The mean fission neutron cross sections of the 127I(n, p)127mTe; 127I(n, p)127Te and 127I(n, 2n)126I reactions were determined experimentally from the absolute 127mTe, 127Te and 128I activities induced in irradiated potassium iodide. For this purpose a quantitative separation method for tellurium from iodine was developed using liquid-liquid extraction of iodine and precipitation of tellurium by means of hydrazine hydrate. The absolute tellurium activity was measured with a calibrated Geiger-Müller tube whereas the iodine activity was measured by gamma spectrometry. For the cross sections of the reactions 127I(n, p)127mTe and 127I(n, p)127Te respective values of 11.1 ± 0.2 μbarn and 7.62 ± 0.11 μbarn were obtained. For the cross section of the reaction 127I(n, 2n)126I a value of 0.647 + 0.010 μbarn was found.