Fission Detectors Research Articles

Fission Cross-Section Ratio Measurements of 239Pu and 233U to 235U from 0.24 to 24 keV

The fission cross-section ratios of 239Pu to 235U and 233U to 235U have been measured from 0.24 to 24 keV using parallel-plate, back-to-back fission detectors and a slowing down lead spectrometer a...

A Resonance Self-Indication Technique for Isotopic Assay of Fissile Materials

A resonance self-indication technique, that measures nondestructively the thickness of fissile materials, has been investigated. This method utilizes the resonance structure in the neutron fission cross section by passing an epithermal beam of neutrons through the sample and then to thin fission detectors that are sensitive to the resonance absorption lines in the transmitted flux corresponding to the resonance reaction peaks. The measurements included samples of 283U, 235U, and 239Pu with thicknesses ranging from 5 to 270 mils. The present measurements indicate that this technique could be used to measure the thickness of 239Pu with an accuracy of 1 to 3%, and an accuracy of 2 to 10% for 233U and 235U for thicknesses <150 mils. The influence of extraneous material in the samples was greatly reduced by using ratios of different fission detectors in the measurement. Computer calculations of the fission rates were made, and the theoretical results are in good agreement with the measurements.

Preparation of thin layers of 237Np for absolute counting and fission detectors

A review of electrodeposition procedures for the preparation of 237Np is given and two new procedures reported making use of “molecular plating”. High voltage - low current yield determinations were carried-out in the region 500–1700 V dc with a deposition current constant at 1.0 mA. The suitability of aluminium, titanium, copper and stainless steel as cathode material was also investigated. It was found that using stainless steel as a cathode and a voltage of 1700 V for a 10 min period resulted in a yield of 100% 237Np, providing that the total solids in solution were within the limits of 20 μg/cm 2 to 500 μg/cm 2. The quality of the layers produced by the described method make them very suitable for alpha-spectrometry and for the fabrication of fission detectors.

Variation with Neutron Energy of the Yield of High-Kinetic-Energy Fragments fromU233,U235, andPu239

A fission chamber has been devised which detects with high efficiency those fission events in which the kinetic-energy release of the heavier binary fragment is maximal. This detector has been used, in conjunction with a conventional fission detector (whose efficiency is independent of fragment energy), in a study of the relative yield of high-kinetic-energy fission of $^{233}\mathrm{U}$, $^{235}\mathrm{U}$, and $^{239}\mathrm{Pu}$, induced by low-energy neutrons. The variations in the yield of high-kinetic-energy fragments shows a qualitative correlation with results of measurements of the relative symmetric fission yield as a function of neutron energy. The results indicate that an increase in symmetric fission is accompanied by an increased yield of certain high-kinetic-energy fragments. Such behavior shows that the fission process is more complex than has generally been assumed.

A Neutron Counting System with a Fission Detector and a Flexible, Balanced Line

A neutron counting system containing a fission detector and a pulse transformer driving a balanced transmission cable was investigated to determine its usefulness as an instrument for reactor control. Because of the noise rejection capability of the balanced system input, all active elements can be located away from the hostile environment of the reactor, resulting in longer periods of trouble-free operation. The design criteria of Fowler were used as a guide. The primary objective was to measure the response of the system to electrical interference characteristic of the Oak Ridge Research Reactor (ORR). The results showed no noise pulses in excess of the alpha threshold after six days of operation at the ORR. A detector without 235U was used to make this measurement. The input cable, with a characteristic impedance of 250 ohms, was a 40-ft long double-shielded twisted pair. Its outer jacket and inner dielectric were made of irradiated polyolefin compounds. The cable was enclosed in a flexible, stainless steel conduit for underwater operation. The quality of the shield of this cable and four other types of balanced cables was determined by a measurement of their transfer impedance. To determine the effect of the flexible conduit, a separate measurement was made with the cable-conduit combination. The techniques required to make these measurements are critical and are described. The signal from the detector-cable combination requires a preamplifier with a lownoise input stage. A series-shunt feedback stage using two 2N2857 transistors gave an input noise of approximately 2.

Measurement of the Average Neutron Yield per Fission forAm242m

The average number of neutrons per thermal neutron fission, $\overline{\ensuremath{\nu}}$, was measured for ${\mathrm{Am}}^{242m}$ with the use of a spark-chamber fission detector and a paraffin-moderated neutron detector. The value obtained is 3.24\ifmmode\pm\else\textpm\fi{}0.12 neutrons per fission. To check the method, measurements also were made on ${\mathrm{U}}^{235}$ and ${\mathrm{U}}^{233}$, which gave 2.43\ifmmode\pm\else\textpm\fi{}0.08 and 2.54\ifmmode\pm\else\textpm\fi{}0.04 neutrons per fission, respectively.

Response of Silicon Fission Detectors to Channeled 127I and 40Ar Ions

When very heavy ions were directed along the crystalline channels of Si fission detectors, a three-fold improvement in energy resolution was obtained for particles remaining in the channels. For such channeled particles the usual pulse height defect was reduced by more than a factor of 10. The results seem to fit certain theoretical explanations of detector nonlinearities found in the literature, in which the nonlinearities are ascribed to the predominance of atomic collision processes near the end of the range.

The room scattered background produced by a neutron source in a laboratory with concrete walls

The neutron backgrounds present in a laboratory containing a neutron source, due to the scattering in the walls, have been measured for fission detectors and for a BF 3 neutron detector and agree to within a factor 2 with the backgrounds calculated on the continuous slowing down model. For a fission detector containing thermally fissile material the relative background at a distance of 20 cm from a neutron source in a cubical laboratory of 10 m side is typically 20%.

Neutron and Gamma-Ray Rate Sensitivities of Several Dynamic Detectors Used in Radiation Effects

The absolute rate-sensitivities of several burst sensors to neutrons and to gamma-rays in the mixed field of the Sandia pulsed reactor environment have been measured, using the shielding techniques of Coppage and Snyder(1) for variation of the neutron-to-gamma ray ratio over a factor of five to ten. The spectral changes effected by the shields are discussed. The sensors which were tested are: two reverse-biased preirradiated silicon diodes, MgO-RAD, phosphorphotodiode, and a fission (U238) detector. The ratio of the sensitivities of the detectors are 4. 8, 7.8, 9.1, 22, and 46 (x10-11 R-cm2/neutron (Pu)), respectively.

Measurements of Deviation from 1/EDistribution in Epi-Cadmium Neutron Spectrum

Although the neutron spectra in the epi-thermal region in thermal reactors are usually assumed to be of 1/E distribution, it is important in some cases to consider the deviation of the distrtribution from the 1/E shape. In this paper an approach to the representation of the reaction rate of the epi- cadmium energy neutron is presented, and a characteristic parameter defining the deviation from the 1/E shape is introduced. This parameter is useful in studying the spatial dependence of epi-cadmium neutron spectra in thermal reactors. It is shown that a 235U fission detector is one of the best detectors for this purpose. The measurement involved was performed by using a 235U fission counter in the SHE critical assembly. The experimental results obtained are compared with theoretical results, and the agreement between the two is good.

Measurements of Deviation from 1/E Distribution in Epi-Cadmium Neutron Spectrum

Although the neutron spectra in the epi-thermal region in thermal reactors are usually assumed to be of 1/E distribution, it is important in some cases to consider the deviation of the distrtribution from the 1/E shape. In this paper an approach to the representation of the reaction rate of the epi-cadmium energy neutron is presented, and a characteristic parameter defining the deviation from the 1/E shape is introduced. This parameter is useful in studying the spatial dependence of epi-cadmium neutron spectra in thermal reactors. It is shown that a 235U fission detector is one of the best detectors for this purpose. The measurement involved was performed by using a 235U fission counter in the SHE critical assembly. The experimental results obtained are compared with theoreticalresults, and the agreement between the two is good.

(n, γ) Cross Sections ofNa23,I127, andAu197

The cross sections for the ${\mathrm{Na}}^{23}(n, \ensuremath{\gamma}){\mathrm{Na}}^{24}$ and the ${\mathrm{I}}^{127}(n, \ensuremath{\gamma}){\mathrm{I}}^{128}$ reactions have been measured by an activation method over the neutron energy range 20 kev to 1 Mev. The neutron flux was measured with a fission detector which contained a thin layer of known weight of ${\mathrm{U}}^{235}$. The ${\mathrm{Na}}^{23}$ cross-section variation with energy shows incompletely resolved resonance structure; an average curve through the data decreases as $\frac{1}{{v}_{n}}$ from 2.4 mb at 20 kev to 0.6 mb at 300 kev, and as $\frac{1}{{E}_{n}}$ from 0.6 mb at 300 kev to 0.2 mb at 1 Mev. The ${\mathrm{I}}^{127}$ cross section can be represented approximately by a straight line on a log-log plot, decreasing from 1.0 barn at 30 kev to 80 mb at 1 Mev. The ${\mathrm{Au}}^{197}(n, \ensuremath{\gamma}){\mathrm{Au}}^{198}$ cross section, measured at a neutron energy of 180 kev, is 310\ifmmode\pm\else\textpm\fi{}20 mb.

Gas Scintillation Counter

A systematic investigation of the optimum conditions for the design and operation of a gas scintillation counter has been made. The gases studied were xenon, krypton, argon, helium, and various gas mixtures. Both a 6292 and a quartz window K1306 phototube were used with and without quaterphenyl as a wavelength shifter. Under optimum conditions an energy spread of less than 4% is obtainable for the Po210 α particles. The application of the gas scintillation counter as a fast ``slow neutron detector,'' using the reaction B10(n,α)Li7, and its usefulness as a fast fission detector have been examined and discussed.