Type Of Neutron Detector Research Articles

Neutron Detection Time Distributions of Multisphere LiI Detectors and AB REM Meter at A 20-MeV Electron Linac

Neutron detection time distribution is an important factor for the dead-time correction for moderator type neutron detectors used in pulsed radiation fields. Measurements of the neutron detection time distributions of multisphere LiI detectors (2, 3, 5, 8, 10 and 12 in diameter) and an AB rem meter were made inside an ANL 20 MeV electron linac room. Calculations of the neutron detection time distributions were also made using Monte Carlo codes. The first step was to calculate the neutron energy spectra at the target and detector positions, using a coupled EGS4-MORSE code with a giant-resonant photoneutron generation scheme. The calculated detector spectrum was found in agreement with the multisphere measurements. Then, neutrons hitting the detector surface were scored as a function of energy and the travel time in the room using MCNP. Finally, the above neutron fluence as a function of energy and travel time was used as the source term, and the neutrons detected by 6Li or 10B in the sensor were scored as a function of detection time for each detector using MCNP. The calculations of the detection time distributions agree with the measurements. The results also show that the detection time distributions of detectors, the neutron travel moderators depend mainly on the moderator thickness and neutron spectrum. However, for small detectors, the neutron travel time in the field is also crucial. Therefore, all four factors (neutron spectrum, neutron travel time in the field, detector moderator thickness and detector response function) may play inter-related roles in the detection time distributions of moderator type detectors.

Thin GSO scintillator for neutron detection

The new scintillator cerium-doped gadolinium orthosilicate (GSO-Gd 2SiO 5:Ce) has a light output that is about 20% that of NaI(Tl). The enormous cross section of Gd for capture of thermal neutrons makes GSO a candidate for novel types of neutron detectors. The characteristic radiations from neutron capture in Gd can be stopped in about 75 μm of GSO. Data obtained from a GSO detector that was about 0.6-mm thick demonstrated that thermal neutrons could easily be detected and that higher energy gamma rays caused minimal interference.

Neutron emission time measurements for ICF targets

The neutron emission time for laser-driven inertial confinement fusion targets is determined from data recorded with fast neutron and optical detector systems. Two types of neutron detectors are used. Radiation-induced conductivity devices have a 130-ps FWHM response and are sensitive to targets with yields greater than 5×1010 DT neutrons. They measure the average neutron emission time with a precision of ±50 ps and are fast enough to measure the neutron production rate within a target core as a function of time. Plastic scintillators coupled to microchannel-plate photomultiplier tubes have a 1.2-ns FWHM response and measure the average neutron emission time with a precision of ±75 ps for targets with yields greater than 106 neutrons. Streak cameras record the incident laser power. Optical fiducial signals that are injected into each detector are used to cross time between the detector systems. Measurements made on 1-mm-diam spherical targets irradiated with 23 kJ of 0.35-μm light delivered in a temporally square 1.1-ns pulse and yielding 1013 neutrons show a burn duration greater than 300 ps and an average neutron emission time of about 850 ps.

Testing of compton-emission type detectors of neutrons and of triaxial fission chambers in a pulsed reactor

Compton-emission type neutron detectors with an emitter containing hafnium or gadolinium and small triaxial fission chambers are promising detectors for monitoring the energy liberation in nuclear reactors. By contrast to the B-emission type neutron detectors which are widely employed in power reactors and have an emitter of rhodium, silver, or vanadium [1-3], the Compton-emission type neutron detectors with an emitter containing hafnium or gadolinium are almost without the inertia 14, 5] which is characteristic of fission detectors. Therefore the Compton-emission type neutron detectors and the fission chambers are preferably employed in automatic monitoring and controlling systems and in safety systems of reactors where the detectors must meet increased requirements in regard to fast response. A technology which is used in analogy in the manufacture of coaxial and triaxial cables with magnesia insulation is used in the manufacture of Compton emission type neutron detectors and fission chambers. We tested three Compton-emiss~on type neutron detectors in cable form with a hafniumcontaining emitter, three Compton-emission type neutron detectors of the same form with a gadolinium-containing emitter, and four triaxial small-scale KtV fission chambers [6]. In regard to their design, the Compton-emission type neutron detectors resembled the previously tested G-emission type neutron detectors with a silver emitter [7]. Measurements which were made during the manufacture of the Compton-emission type neutron detectors and during their operation in a research reactor have shown that the mean-square spread of the emitter diameters of these detectors and of their initial sensitivity amount to about 1.5%. The sensie tive element of the Compton-emission type neutron detector with the hafnium-containing emitter has an outer diameter of 3.0 m~n and a length of 0.83 m; the detector with the gadolinium-containing emitter has an outer diameter of 3.4 mm and a length of 0.85 m. Thedesign of the fission chambers hardly differs from that described in [8]. The characteristics of the small axial fission chambers with an outer diameter of 6.0 mm and one to two 50-mm-long sensitive elements were studied. In the case of two sensitive elements, the collecting electrode of one of them was without a fissile cover, i.e., it was used to monitor the dose rate of the y radiation. The gap between the collecting electrode and the housing of the fission chamber was about 0.7 mmwide. The operative volume of the chamber was filled with argon ofanalysis-type purity under a pressure of 0.101 MPa (760 mm Hg) at a temperature of 20~ The operating voltage applied to the electrodes of the triaxial chamber was changed from 200 to 300 V; this is much higher than in the operation and testing of the chambers in power reactors and research reactors [6]. The increased operating voltage results from the irradiation intensity which in the pulsed graphite reactor is higher than in the reactors indicated above. Magnesium oxide powder of analysis-type purity corresponding to GOST 452675 was the insulation of the Compton-emission type neutron detectors; magnesium oxide powder

Photoneutron Production and Dosimetry in Tissue During High Energy Bremsstrahlung Radiotherapy

Abstract The total yield of photoneutrons from soft-tissue-equivalent material irradiated with bremsstrahlung of end point energy from 6 MeV to 28 MeV has been measured using the 35 MeV University of Melbourne Betatron and a 4p Halpern type neutron detector. The results are compared with calculations based on the cross sections for neutron production of the constituent elements of tissue (H, C, N and O). It is shown that the less common isotopes of the constituent elements (2H, 13C, 15N, 17O, 18O) must be included to obtain agreement with the measured yields. The dose imparted to tissue by the photoneutrons is estimated for the energy range from 6 MeV to 30 MeV and compared with other available measurements/estimations. Further, the dose equivalent due to all photonuclear reactions in tissue is estimated for energies of 24 MeV and 30 MeV.

Estimation of response functions of moderating type neutron detectors by the time-of-flight method combined with a large lead pile

The energy dependent response functions of moderating type neutron detectors having cylindrical polyethylene moderators of 1, 3, 6 and 10 cm thicknesses were determined using the time-of-flight (TOF) technique. By combining the 35 MeV electron linear accelerator with a large lead pile of about 150 t, a pulsed neutron field which is rich in intermediate neutrons was formed and its spectrum at the standard measuring point for the TOF experiment was measured by a Li glass scintillator. In this kind of TOF, the neutron time delay in the polyethylene moderator due to multiple collisions with hydrogen was evaluated by the time dependent neutron transport Monte Carlo code, and compiled to the time response matrix. By using the neutron spectrum at the standard measuring point and the time response matrix, the energy dependent response functions of the moderating type neutron detectors were determined from the measured time spectra for the first time in a considerably wide energy range from a few eV to about 50 keV. The experimental response functions showed good agreement with the calculated results.

Experiments and Analyses of Neutron and Gamma-Ray Streaming in the Cavity-Duct System of a Fast Neutron Source Reactor

Reaction rate and dose rate distributions due to streaming of neutrons and gamma rays were measured in the nearly cubical cavity and in the ducts leading to the cavity at the fast reactor YAYOI. The experimental configuration was arranged in a way that streaming through the ducts occurred only after scattering in the cavity. This was done by making the duct axes perpendicular to the source radiation from the reactor core. The spectrum of the source was modified from fast to thermal by putting moderators in the beam hole of the core assembly. The spatial distribution of the neutron reaction rates and gamma-ray dose rates in the cavity is almost flat except for a small decrease at the wall. The cadmium ratios at the duct inlets and the attenuation profiles in the ducts are almost identical irrespective of the spectrum of neutrons entering the cavity. The two-dimensional discrete ordinates transport calculation by TWOTRAN-II has shown that although the ray effect is observed in the fast neutron flux distribution in the cavity, the calculated results agree rather well with the experimental distributions in the cavity and with the streaming in the ducts with regard to the results measured with the l/v type neutron detectors. Also, for the gamma rays the agreement is rather good between the calculation and the experiment.

A New Reactor Safety Circuit for Low-Power-Level Operation

In the operation of nuclear reactors at low-power levels, one of the primary instrumentation problems is that the statistical fluctuations of reactor neutron population are accentuated by conventional log-count-rate and differentiating circuits and can cause frequent spurious scrams unless long time constants are incorporated in the circuit. Excessive time constants may introduce undesirable delay in the circuit response to legitimate scram signals. This paper develops the concept of a count doubling-time monitor which generates a scram signal if the number of counts from a pulse type neutron detector doubles in a given period of time. The paper demonstrates the theoretical relation between count doubling time and asymptomatic periods. A practical circuit to implement the function is described.

An On-Line Reactivity Meter Based on Reactor Noise Using Two-Detector Crosscorrelation

A prototype on-line reactivity meter using reactor noise analysis techniques is based on a two-detector cross correlation method, which offers as a special feature the absence of uncorrected noise contribution to the spectral density function of the reactor system. Reactivity shutdown measurements were performed on three different zero power reactors down to seven dollars. Special attention is given to the predicted and measured error margins of reactivity. The described on-line meter, when used with optimized current type neutron detectors for minimum gamma contamination, promises to be an encouraging way of making shutdown reactivity measurements, perhaps also in “dirty” power reactors where the high gamma-ray intensities can considerably destroy the “clean neutron signal-to-noise” ratio.

New Developments in Photoemissive Tubes

A discussion on a variety of special types of photomultipliers shall be presented. Among these will be the CBS type CL-1090, 14-stage, high speed, low dark current type and the 5-inch ruggedized photomultiplier. Also to be presented are new developments in special type dynode multipliers and large area, high current photomultipliers. Recent developments in vacuum type neutron detectors shall be discussed briefly.

High-Energy Neutron Detectors Employing Liquid Scintillators

Two liquid scintillation type neutron detectors with efficiencies of the order of 5% for counting 130-Mev neutrons have been studied. A fairly monoenergetic source of neutrons was obtained from the photodisintegration of deuterons. Estimates were made of the efficiencies of the counters as a function of bias. Experiments indicate: (1) an appreciable effect of the products of neutron reactions with carbon; (2) the efficiency can be more accurately predicted for a small counter although the behavior of large counters can be qualitatively predicted; (3) the attenuation due to a lead shield in the beam depends upon the neutron energy, the bias, and the geometry; (4) lead shielding on the sides is relatively unimportant if kept thin.

Nonelastic Scattering Cross Sections for Fast Neutrons

Measurements have been made of the nonelastic cross sections of neutrons in Be, C, Al, Ti, Cr, Fe, Ni, Cu, Ag, Sn, Pb, and Bi. These cross sections include all processes except elastic scattering. The experiments have been carried out with monoenergetic neutrons of 3.5, 4.7, 7.1, 12.7, and 14.1 Mev. The neutrons were produced with a Van de Graaff accelerator using the nuclear reactions: ${\mathrm{T}}^{3}(p, n){\mathrm{He}}^{3}$, $\mathrm{D}(d, n){\mathrm{He}}^{3}$, and ${\mathrm{T}}^{3}(d, n){\mathrm{He}}^{4}$. Cross sections were determined from transmissions through spherical shells which were 3.8 cm in radius and 2 cm thick. A new type of neutron detector which gives energy information about neutrons and is insensitive to $\ensuremath{\gamma}$ radiation was placed at the center of the spherical shell and was used as a biased detector of neutrons. The results of these experiments show that the nonelastic neutron cross section for 12.7- and 14.1-Mev neutrons is nearly geometrical and is closely approximated by $\ensuremath{\sigma}=\ensuremath{\pi}{(R+\ensuremath{\lambda})}^{2}$, where $R=1.4\ifmmode\times\else\texttimes\fi{}{10}^{\ensuremath{-}13}{A}^{\frac{1}{3}}$. The nonelastic cross sections are roughly constant in the energy range 5 to 14 Mev; below about 5 Mev the cross sections begin to decrease.The elastic cross sections can be obtained by subtracting the nonelastic cross sections from the total cross sections. The variations with energy of the nonelastic and elastic cross sections are quite different.Data with 12.7-Mev neutrons and low counter bias, give the relative number of $\ensuremath{\gamma}$ rays given off per inelastic collision of these neutrons. The number of $\ensuremath{\gamma}$ rays decreases rapidly with increasing atomic number and there are less than 5% as many $\ensuremath{\gamma}$ rays from Bi as from Al, Ti, Cr, and Fe.