Pulsed Neutron Source Technique Research Articles

Effect of higher harmonics in the area-ratio pulsed neutron source technique

The area-ratio pulsed neutron source technique, which can yield a subcriticality in dollar units, is given by Ap/Ad, the area ratio of the prompt neutron component to the delayed neutron component. In this paper, both components are reconstructed by the summation of the spatial harmonics that are obtained by solving the α–mode eigenvalue equations based on the neutron transport and diffusion theories. Monte Carlo simulations for the area-ratio method in a one-dimensional slab are performed. The reconstructed prompt and delayed components agree well with the simulation results. The areas Ap and Ad contain a higher harmonic effect, and the ratio Ap/Ad differs from the true subcriticality. The effect of the higher harmonics in Ap can be reduced by the extrapolated area-ratio method if the earlier period after the pulse generation is successfully excluded from the fitting period. However, there is no way to remove the effect of the higher harmonics from the delayed neutron component, but it alleviates itself because it is given by the subtraction of the prompt component from the total (delayed and prompt) component. The subcriticality in dollar units adjusted by the extrapolated area-ratio method is considerably improved compared with the unadjusted Ap/Ad. The improvement is remarkable for a near-critical system in which the denominator Ad is relatively large, and the effect of the higher harmonics in Ad is minor.

Development and Experimental Validation of a Correlation Monitor Tool Based on the Endogenous Pulsed Neutron Source Technique

AbstractA correlation measuring tool for an endogenous pulsed neutron source experiment is developed in this work. Paroxysmal pulses generated by a bursts of neutron chains are detected by a 10-kbit embedded shift register with a time resolution of 100 ns. The system is implemented on a single reprogrammable device making it a compact, cost-effective instrument, easily adaptable for any case study. The system was verified experimentally in theEsfahan heavy-water zero power reactor(EHWZPR). The results obtained by the measuring tool are validated by the Feynman-α experiment, and a good agreement is seen within the boundaries of statistical uncertainties. The theory of the methods is briefly initiated in the text. Also, the system structure is described, the experimental results and their uncertainties are discussed, and neutron statistics in EHWZPR is examined experimentally.

Differential Time of Flight Technique for the Detection of Special Nuclear Materials

An adaptation of the differential die away (DDA) technique, which enhances the detection of fissile materials, has been developed and demonstrated in the laboratory. The differential time-of-flight (DTOF) technique applies to situations where special nuclear material (SNM) is located some distance (up to a few meters) away, is accessible from only one side, and is either bare or embedded in an environment that is slightly neutron moderating, and where the more conventional DDA pulsed neutron source technique is less effective. The technique is based on the fission reaction of thermal neutrons with fissile material, if present. The thermal neutrons are furnished by the moderation of neutrons from an electronic neutron generator (ENG). Fast neutrons are generated in relatively narrow pulses of tens to hundreds of microseconds. They are slowed down in a suitable moderator surrounding the source. The time of flight (TOF) of these neutrons, before arrival at the SNM, spreads over orders of magnitude due to the wide spread in their velocities, which lead to a time spread of the neutron-induced fission in the SNM. Neutron signatures resulting from the fissions, e.g., prompt fast neutrons (as well as few delayed neutrons) can then be detected by fast neutron detectors well after the original source neutrons have died away. The fast neutron detectors must be insensitive to the numerous source-related thermal neutrons and are a critical component of the technique, since they can differentiate, by time as well as energy, between the fission signatures and the source-related background. The choice of source moderator is also important to the success of the technique. Two moderators were designed and studied: one made of polyethylene and the other made of beryllium (Be). The latter is superior delivering a higher flux and longer neutron die-away times. The feasibility of the DTOF technique was demonstrated by detecting a sample of 350 g 235U (in the form of 19.9% enriched uranium) at a range of distances and under a variety of conditions employing a commercial (d, T) pulsed neutron generator and the two moderators.

Advantage of the area-ratio pulsed neutron source technique for ADS reactivity calibration

In this work, one shows the area-ratio pulsed neutron source (PNS) technique is as good candidate for reactivity calibration of accelerator-driven systems (ADS). Some properties of that method are discussed. One starts with a newly developed point-kinetics model that is able to simulate the PNS histogram experimentally obtained. The question of the convergence rate of the estimator based on that method is addressed. Some simulations that are in good agreements with experimental data reveal the convergence time is clearly longer than the flux stabilization time due to the delayed neutron. The sensitivity of the area-ratio estimator to the pulse width, the time discretization and the source strength variation is then explored. Finally, the initial model is extended to a reflected neutronic system. It is thus shown that the area-ratio reactivity estimator still provides satisfactory results in the core and reflector region.

Reactivity Measurements under Conditions Typical to Fuel Element Dissolution

Pulsed neutron source measurements have been made on a heterogeneous lattice of plutonium-uranium oxide fuel rods in 258 g(Pu + U)/ℓ nitrate solution containing up to 1.34 g Gd/ℓ. The experimental system on which the measurements were made is not unlike that encountered in fuel element dissolvers. The objectives of the measurements were to demonstrate the use of the pulsed neutron source technique for measuring the effectiveness of a neutron poison in reducing the reactivity of such a system and to determine the kinetic parameter β/l for these systems. Reductions in keff from unity down to 0.64 were observed upon the addition of 1.34 g Gd/ℓ to a critical system. Based on the prompt and delayed critical conditions determined for each gadolinium concentration, a continuous reduction, from $4.35/cm of solution depth down to $0.42/cm, was observed in the reactivity worth of the plutonium-uranium nitrate solution as gadolinium was added to the solution. The values of β/l as a function of gadolinium concentration was observed to vary essentially linearly from 197 to 262 sec−1 as the gadolinium concentration was increased to 1.28 g/ℓ. At the maximum gadolinium concentration of 1.34 g/ℓ, the measurements indicated a β/l value lying above this linear correlation, but not far enough above that it could not be explained by the 0.4% difference observed in the approach-to-critical and the pulsed-neutron-determined delayed critical conditions for this system. The effective delayed neutron fraction, βeff, for these mixed plutonium-uranium systems was calculated to be 0.0033 and was essentially constant over the gadolinium concentration covered. The βeff, calculational technique was subjected to an experimental-calculational verification and was found to be adequate.

An investigation of neutron streaming using the pulsed neutron source technique

The radial streaming of neutrons from H 20 moderator assemblies pierced by a uniform array of cylindrical voids is investigated by the pulsed neutron source technique. A method of synthesis is developed by which the fundamental mode in the spatial distribution of the thermal neutron flux within a moderator assembly may be readily extracted from experimental data. The decay constants determined by the synthesis method were used to derive the magnitude of the radial diffusion constant D r v. The value of D r v agrees with that predicted by the theory of Leslie.

Measurement of the thermal neutron diffusion parameters in diphenyl and monoisopropylbiphenyl (MIPBy pulsed neutron methods

The diffusion parameters in diphenyl (C 10H 10) at 77°C and MIPB (C 15H 16) at 30°C were determined using the pulsed neutron source technique. The 1 MeV Van de Graaff accelerator at North Carolina State College, Raleigh, was used as the pulsed source of fast neutrons utilizing the 9Be (d, n) 10B reaction. The decay of the thermal neutron leakage from a finite cylindrical geometry was measured using a 26-channel time analyser. By measuring the thermal neutron decay constant, A, as a function of buckling, B 2, a least squares fit of the data to the equation λ = λ 0 + D 0 B 2 − CB 4 yielded values of D 0, the diffusion constant, and C, the diffusion cooling coefficient. The infinite medium decay constant, λ 0, was calculated from known cross-section data. For diphenyl measurements in the buckling range from 0·0816 cm −2 to 0·2811 cm −2, the diffusion parameter values obtained were D 0 = 5437 ± 0117 × 10 4 cm 2-sec −1 and C = 9·85 ± 5·50 × 10 3 cm 4-sec −1. The results obtained for MIPB in the buckling range from 0·0945 cm −2 to 0·2952 cm −2 were D 0 = 3780 ± 0·033 × 10 4 cm 2-sec −1 and C = 125 ± 1·43 × 10 3 cm 4-sec −1. The values of the transport mean free path, λ tr, and diffusion length, L, derived from these results were found to be less than those values predicted by current theoretical models. A preliminary investigation of λ tr, in MIPB as a function of temperature showed a non-linear behaviour of the transport mean free path with temperature.

The Investigation of Nonelastic Cross Sections for Iron by the Nanosecond Pulsed Neutron Source Technique

The Investigation of Nonelastic Cross Sections for Iron by the Nanosecond Pulsed Neutron Source Technique

The Investigation of Neutron Kinetics and Cross Sections in Fast Nonmoderating Assemblies by the Nanosecond Pulsed Neutron Source Technique

Nanosecond bursts of monoenergetic neutrons in the range 0.8-- 1.6 Mev are injected into non-moderating assemblies of bismuth, lead, and natural uranium. The flux in these assemblies is observed to decay exponentially with characteristic nanosecond time constants in good agreement with one velocity transport theory, and the known inelastic scattering and absorption cross sections. These experiments serve as a check on the validity of the assumptions of transpont theory. The technique also serves as a method for measuring macroscopic inelastic and absorption cross sections directiy, without the necessity of making the corrections required in other methods, e.g., for double scattering and for the angular distribution. (auth)

A Note on the Measurement of Diffusion Parameters by the Pulsed-Neutron Source Technique

(1962). A Note on the Measurement of Diffusion Parameters by the Pulsed-Neutron Source Technique. Nuclear Science and Engineering: Vol. 12, No. 3, pp. 433-435.

Neutron Diffusion in Small Volumes of Water: Pulsed Source Measurements

The thermal neutron diffusion parameters for light water have been measured using the pulsed neutron source technique. It has been demonstrated that the value of the diffusion cooling coefficient is critically dependent on the physical shape of the experimental system. The experimental results have been analysed using two different values of the extrapolation length, since the value of this parameter is not accurately known. The discrepancies between previously published values of the diffusion parameters are largely explained by their dependence on system shape and extrapolation length.

Laboratory Studies of a Pulsed Neutron-Source Technique in Well Logging

Introduction Refinements in radiation logging techniques during recent years have involved increasing usage of scintillation detectors. These detectors produce voltage pulses whose heights are related to the energies of the gamma rays which initiate them. Analysis of the gamma-ray spectrum, as indicated by the pulse heights, yields information about the chemical elements composing the formations surveyed. Refined scintillation counter techniques can furnish chemical information concerning earth formations in situ, from a study of the gamma-ray spectra emitted by the formation either naturally or as a result of neutron bombardment. Accompanying the rising interest in gamma-ray scintillation spectroscopy, there has been increased activity in the development of accelerator-type neutron sources (in contrast to encapsulated chemical-mixture sources). Such neutron generators are attractive for several reasons:they greatly reduce radiation hazards to personnel;there is a great reduction in contamination danger if they are lost in the hole;they can produce larger neutron intensities than can conveniently available encapsulated sources; andthey are capable of being pulsed, thus permitting new techniques in logging. In the past, both accelerator and encapsulated neutron sources have been used by others in conjunction with scintillation-detector pulse-height analysis. The results have not been too encouraging, due to the interference among different gamma-ray spectral lines and to the fact that the gamma-ray peaks were not too clearly distinguishable above the large and ill-defined background "noise". This paper is a status report on laboratory studies of a technique using a borehole accelerator as a neutron source, which gives an improved scintillation spectrum, thus permitting more accurate chemical analyses of the formations penetrated. The Schlumberger-accelerator neutron source is presented; the origins of inelastic and of thermal-neutron, capture gamma rays are discussed, and results are given for some laboratory measurements performed in borehole geometry.