Pulsed Neutron Experiments Research Articles

Neutron calorimeter designfor pulsed neutron experiment diagnostics.

Neutron calorimeter for diagnostic support of both fission- and fusion-based pulsed neutron experiments is described. The calorimeter generates a voltage signal output via a pair of thermocouples contacting a sample material, which experiences a temperature increase due to interactions with an incident neutron fluence. Advantages of the calorimeter include the versatility derived from its small size, modularity, and the variety of employable neutron absorber materials and sizes. These characteristics permit use with neutron energies between the thermal and fast (up to 14MeV) spectra. Designoverview and requirements, Monte Carlo particle transport simulation results, test data, and possibilities for calorimeter development are provided and discussed.

Development of a biaxial tensile testing machine for pulsed neutron experiments

A biaxial tensile testing method has been used to get macroscopic information on elastoplastic deformation of a thin steel specimen and improve the accuracy of plastic processing of steel materials. We newly developed a biaxial tensile testing machine for pulsed neutron experiments (BTM-NEU) to provide the microscopic crystallographic information of steel materials under biaxial load and correlate it with the macroscopic mechanical properties of the materials. The performance of the BTM-NEU was experimentally evaluated with cold-rolled mild steel and hot-rolled high-tensile-strength steel materials and compared with that of a standard biaxial tensile testing machine (BTM-std) as follows.•The BTM-NEU can test an ISO-standardized cruciform specimen as the BTM-std and its performance is equivalent to that of the BTM-std.•The BTM-NEU has excellent long-time reliability and stability necessary for pulsed neutron experiments, especially Bragg-edge neutron imaging experiments.•The BTM-NEU can be applied to pulsed neutron experiments using a Bragg-edge imaging method.

Calculating Alpha Eigenvalues and Eigenfunctions with a Markov Transition Rate Matrix Monte Carlo Method

For a nuclear system in which the entire -eigenvalue spectrum is known, eigenfunction expansion yields the time-dependent flux response to any arbitrary source. Applications in which this response is of interest include pulsed-neutron experiments, accelerator-driven subcritical systems, and fast burst reactors, where a steady-state assumption used in neutron transport is invalid for characterizing the time-dependent flux. To obtain the -eigenvalue spectrum, the transition rate matrix method (TRMM) tallies transition rates describing neutron behavior in a discretized position-direction-energy phase space using Monte Carlo. Interpretation of the resulting Markov process transition rate matrix as the operator in the adjoint -eigenvalue problem provides an avenue for determining a large finite set of eigenvalues and eigenfunctions of a nuclear system. Results from the TRMM are verified using analytic solutions, time-dependent Monte Carlo simulations, and modal expansion from diffusion theory. For simplified infinite-medium and one-dimensional geometries, the TRMM accurately calculates eigenvalues, eigenfunctions, and eigenfunction expansion solutions. Applications and comparisons to measurements are made for the small fast burst reactor CALIBAN and the Fort St. Vrain high-temperature gas-cooled reactor. For large three-dimensional geometries, discretization of the large position-energy-direction phase space limits the accuracy of eigenfunction expansion solutions using the TRMM, but it can still generate a fair estimate of the fundamental eigenvalue and eigenfunction. These results show that the TRMM generates an accurate estimate of a large number of eigenvalues. This is not possible with existing Monte Carlo–based methods.

Method and Means for Monitoring the Spent-Fuel Pool Subcriticality at the Smolensk NPP

Research based on the pulsed neutron experiment method (α-method) of subcriticality monitoring of the spent fuel pool at the Smolensk NPP is presented. The procedure includes stationary and nonstationary neutron-physical calculations and measurements of the main characteristics of a subcritical multiplying system, which the spent-fuel pool is. A description of the STEPAN-KhOYaT software specially developed for the computational support of the pulsed experiment is given. The subcriticality for real loads of spent-fuel pools is calculated and measured using the procedure.

Determination of prompt-neutron decay constant from phase shift between beam current and neutron detection signals for an accelerator-driven system in the Kyoto University Critical Assembly

A unique power spectral analysis for a subcritical reactor system driven by a pulsed 14 MeV neutron source was carried out at the Kyoto University Critical Assembly (KUCA). In this analysis, a complex cross-power spectral density between time-sequence signal data from an accelerator beam ammeter and a neutron detector was measured to determine the prompt-neutron decay constant of an accelerator-driven system (ADS) from the phase data of the spectral density. Assuming the one-point kinetics model, in theory, the decay constant can be arithmetically derived from the phase at the integral multiples of the pulse repetition frequency. However, the actual derivation from the phase at a pulse repetition frequency of 20 Hz considerably underestimated the prompt-neutron decay constant, compared with that obtained by a previous pulsed neutron experiment, and the derived decay constant apparently decreased with an increase in the multiple of the pulsed repetition frequency. Considering a lag time in detector response, the above underestimation and the above apparent decrease were solved to obtain the consistent decay constant. While both previous power spectral analysis and Feynman-α analysis for pulsed neutron source require non-linear least-squares fits of the respective complicated formulae, the present analysis makes the fitting unnecessary except at regular calibration of the lag time. This feature is advantageous for a robust online monitoring of subcritical reactivity of an actual ADS.

Measurement of large negative reactivity of an accelerator-driven system in the Kyoto University Critical Assembly

Large negative reactivity of a subcritical system driven by a pulsed 14 MeV neutron source has been measured in the Kyoto University Critical Assembly. The subcriticality of the accelerator-driven system (ADS) ranged in effective multiplication factor roughly from 0.98 to 0.92, which corresponded to an operational range of an actual ADS proposed by Japan Atomic Energy Agency. As the measurement technique, pulsed neutron method, power spectral analysis for pulsed neutron source, accelerator-beam trip method were employed. From neutron count decay data obtained by the pulsed neutron experiment, not only the prompt-neutron decay constant of fundamental mode but also a higher spatial mode could be derived. The subcriticality was also determined from the fundamental decay constant. The measured cross-power spectral density consisted of a familiar correlated reactor-noise component and many uncorrelated delta-function-like peaks at the integral multiple of pulse repetition frequency. The fundamental prompt-neutron decay constant, i.e., the subcriticality determined from the latter uncorrelated peaks was consistent with that obtained by the above pulsed neutron experiment. However, the magnitude of the former correlated component was reduced with an increase in the subcriticality and eventually this component became almost white at deeply subcritical state ranging in the multiplication factor under 0.95. Consequently, the determination of the decay constant from the correlated component was impossible under such a subcritical state. As data analysis method for the beam trip experiment, both the conventional integral count method and the least-squares inverse kinetics method (LSIKM) were employed. The LSIKM analysis led to the consistent subcriticality with that obtained by the pulsed neutron experiment, while the integral count method significantly underestimated the subcriticality. This underestimation originated from a residual background count, which was maintained after the beam trip. The LSIKM was mostly not influenced by such a slight count rate.

Power spectral analysis for a thermal subcritical reactor system driven by a pulsed 14 MeV neutron source

A series of power spectral analyses for a thermal subcritical reactor system driven by a pulsed 14 MeV neutron source was carried out at Kyoto University Critical Assembly (KUCA), to determine the prompt-neutron decay constant of the accelerator-driven system (ADS). The cross-power spectral density between time-sequence signal data of two neutron detectors was composed of a familiar continuous reactor noise component and many delta-function-like peaks at the integral multiple of pulse repetition frequency. The prompt-neutron decay constant inferred from the reactor noise component of the cross-power spectral density was consistent with that obtained by a pulsed neutron experiment. However, the reactor noise component of the auto-power spectral density of each detector was hidden by a white chamber noise in the higher-frequency range and this feature resulted in a considerable underestimation of the decay constant. For several runs with a low pulse-repetition frequency, furthermore, we attempted to infer the decay constant from point data of the delta-function-like peaks. The analysis for a run under a slightly subcritical state resulted in the consistent decay constant; however, those for other runs under significantly subcritical states underestimated the decay constant. Considering the contribution of a spatially higher mode to the point data, the above underestimation was solved to obtain the consistent decay constant. While the Feynman-α formula for a pulsed neutron source is too complicated to be fitted directly to variance-to-mean ratio data, the present analysis on frequency domain is much simpler and the conventional formula based on the first-order reactor transfer function is available for fitting to power spectral density data.

Feynman-α Analysis for a Thermal Subcritical Reactor System Driven by an Unstable 14MeV Neutron Source

In a series of Feynman-α correlation measurements for a thermal Accelerator-Driven System (ADS) with 14MeV neutrons at the Kyoto University Critical Assembly (KUCA), an unstable accelerator condition such as a drift of beam current has been frequently observed. Neutron source instability caused by such unavoidable beam-current instability resulted in a divergent variance-to-mean ratio and, consequently, the correlation analysis failed. Nevertheless, we attempted to apply a difference-filtering technique to the correlation analysis to reduce the influence of the above instability. The present attempt resulted in consistent prompt-neutron decay constants with those obtained in a previous pulsed neutron experiment. The application of the filtering is expected to enhance the robustness of Feynman-α analysis against various instabilities of accelerator operation in actual ADS.

Determination of Subcritical Reactivity of a Thermal Accelerator-Driven System from Beam Trip and Restart Experiment

An experimental technique based on an accelerator-beam trip or restart operation is proposed to determine the subcritical reactivity of an accelerator-driven system (ADS). Applying the least-squares inverse kinetics method to the data analysis, the subcriticality can be inferred from time-sequence neutron count data after these operations. A series ofbeam trip and restart experiments with 14 MeV neutrons were carried out in a thermal ADS of Kyoto University Critical Assembly (KUCA), to demonstrate the applicability of the proposed technique. The subcriticalities evaluated using neutroncounters far from the DT target were consistent with those obtained in a previous pulsed neutron experiment. However, a counter placed close to the target significantly overestimated the subcriticality. The present technique is expected to be available for subcriritcality measurement at startup and shutdown of various ADSs.

Feynman-α Analysis for a Thermal Subcritical Reactor System Driven by an Unstable 14 MeV Neutron Source

In a series of Feynman-α correlation measurements for a thermal Accelerator-Driven System (ADS) with 14MeV neutrons at the Kyoto University Critical Assembly (KUCA), an unstable accelerator condition such as a drift of beam current has been frequently observed. Neutron source instability caused by such unavoidable beam-current instability resulted in a divergent variance-to-mean ratio and, consequently, the correlation analysis failed. Nevertheless, we attempted to apply a difference-filtering technique to the correlation analysis to reduce the influence of the above instability. The present attempt resulted in consistent prompt-neutron decay constants with those obtained in a previous pulsed neutron experiment. The application of the filtering is expected to enhance the robustness of Feynman-α analysis against various instabilities of accelerator operation in actual ADS.

Spatial Corrections for Pulsed-Neutron Reactivity Measurements

For pulsed-neutron experiments performed in a subcritical reactor, the reactivity obtained from the area-ratio method is sensitive to detector positions. The spatial effects are induced by the presence of both the prompt neutron harmonics and the delayed neutron harmonics in the reactor. The traditional kinetics distortion factor is only limited to correcting the spatial effects caused by the fundamental prompt-α mode. In this paper, we derive spatial correction factors fp and fd to account for spatial effects induced by the prompt neutron harmonics and the delayed neutron harmonics, respectively. Our numerical simulations with the FX2-TH time-dependent multigroup diffusion code indicate that the high-order prompt neutron harmonics lead to significant spatial effects and cannot be neglected in calculating the spatial correction factors. The prompt spatial correction factor fp can be simply determined by the ratio of the normalized detector responses corresponding to the fundamental k-mode and the prompt neutron flux integrated over the pulse period. Thus, it is convenient to calculate and provides physically intuitive explanations on the spatial dependence of reactivity measured in the MUSE-4 experiments: overestimation of the subcriticality in regions close to the external neutron source and underestimation of the subcriticality away from the source but within the fuel region.

Determination of Lambda-Mode Eigenvalue Separation of a Thermal Accelerator-Driven System from Pulsed Neutron Experiment

Basic research on the Accelerator-Driven System (ADS) with thermal neutron spectrum has been promoted by the Kyoto University Research Reactor Institute. At the Kyoto University Critical Assembly (KUCA), various experiments on thermal ADS with a pulsed spallation source are planned. In such an ADS, neutron flux distribution may be sensitive to the injection of neutrons, and the high sensitivity results in various spatial effects. In this study, a pulsed neutron experiment with 14MeV neutrons was carried out in a thermal ADS of KUCA, to determine the λ-mode eigenvalue separation, which is a quantitative indication of spatial effects. An original data-processing technique was applied to infer prompt-neutron decay constants of fundamental and higher modes from neutron count decay data, and then the eigenvalue separation around 13%Δk/k was obtained from these decay constants.

Monte Carlo simulation and benchmarking of pulsed neutron experiments in variable buckling Beo systems

Pulsed neutron decay simulation of BeO moderator as a result of injecting 14 MeV neutron pulses into finite size systems are carried out with the MCNP Monte Carlo code. The simulated decay constants as compared against a previous experimental work showed variations of about 2%. The resulting decay constants and the fitted diffusion coefficients based on the simulation results are found to agree with the experimental results within a margin of about 4%.

Diffusion cooling of thermal neutrons in the outer Plexiglas layer of a two-zone system with non-hydrogenous material inside

Diffusion cooling of thermal neutron flux in a finite volume of a medium results from the leakage of neutrons. The cooling effect is particularly complex when the system consists of zones with different neutron scattering properties. A study is presented on the diffusion cooling in Plexiglas (hydrogenous material), which constitutes an outer cylindrical shell surrounding an inner cylinder made of a non-hydrogenous substance. The pulsed neutron method has been used. Numerous series of the desired pulsed neutron experiments in two-zone cylindrical systems have been substituted by Monte Carlo simulations. The decay constant of the fundamental mode in the time distribution of the pulsed thermal neutron flux has been determined in each experiment. The so-called theoretical decay constants (defined under a particular assumption) have been calculated as a function of the system geometry and neutron dynamic parameters including the diffusion cooling coefficient of the outer Plexiglas shell. The variability of this parameter has been determined from a combination of experimental and theoretical results. For a system with a theoretical decay constant varying between 12,000 and 34,000 s −1, the diffusion cooling coefficient of Plexiglas in the outer layer varies between 9000 and 4800 cm 4 s −1 with the latter being close to the value for homogeneous material. A function has been obtained and successfully applied in interpreting real experiments using Czubek’s method of measuring the absorption cross-section of small samples.

Questions of modeling the non-stationary thermal neutron transport in two-zone systems with Plexiglas by means of the MCNP code

The pulsed thermal neutron experiments have been modeled in two-zone spherical and cylindrical systems. The outer zone of Plexiglas has surrounded either hydrogenous or non-hydrogenous inner zone. The MCNP code has been used in the numerical simulations. The standard poly.01t MCNP library and its modification by the authors have been used for Plexiglas. A considerable improvement of the agreement with experimental results is observed while using the proposed modification.

Monte Carlo simulations of the pulsed thermal neutron flux in two-zone systems with Plexiglas – Using the MCNP code with a modified hydrogen-data library

A comparison of real and Monte Carlo simulated pulsed neutron experiments in two-zone cylindrical systems is presented. Such geometry is met when a neutron moderator surrounds a sample of the investigated material. In this study, a Plexiglas shell (hydrogenous medium) surrounds the inner zone filled with a non-hydrogenous medium: copper oxide or chrome oxide. The time decay constant of the thermal neutron flux is determined as the result of the experiment. The primary simulations have been made using the MCNP code with the attached standard thermal neutron scattering library for hydrogen in polyethylene (poly.01t). A modification of this library is proposed to obtain the data dedicated more precisely for scattering of neutrons on hydrogen in Plexiglas in the thermal energy region. Results of the simulations for two-zone cylindrical systems, using the MCNP code with the modified hydrogen-data library, show a considerably better agreement with the experimental results. The average relative deviations have decreased from about 2% (always positive) to less than 0.5% fluctuating around zero. Adequacy of the applied modification is also confirmed in simulations of the pulsed neutron experiments on homogeneous cylinders of Plexiglas.

Modelling of the non-stationary thermal neutron transport in hydrogenous media using the MCNP code

The work is aimed to compare results of the Monte Carlo simulations of pulsed neutron experiments with results of real experiments. The simulations have been performed for homogenous and two-zone system. In the first case the cylinder of Plexiglas has been used. In two-zone systems, Plexiglas has been used as an outer moderator and aqueous solutions of H 3BO 3 or KCl of known concentrations have been used as the inner samples. The simulations have been performed (1) using for neutron scattering in Plexiglas a standard library for H in polyethylene (a commonly suggested way) and (2) using a modification of this library. The modification gives very good agreement between simulations and experimental results.

Decay of the pulsed thermal neutron flux in two-zone hydrogenous systems – Monte Carlo simulations using MCNP standard data libraries

Pulsed neutron experiments in two-zone spherical and cylindrical geometry has been simulated using the MCNP code. The systems are built of hydrogenous materials. The inner zone is filled with aqueous solutions of absorbers (H3BO3 or KCl). It is surrounded by the outer zone built of Plexiglas. The system is irradiated with the pulsed thermal neutron flux and the thermal neutron decay in time is observed. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances have been used to simulate the neutron transport. The time decay constant of the fundamental mode of the thermal neutron flux determined in each simulation has been compared with the corresponding result of the real pulsed neutron experiment.

Some aspects of random geometry and absorption in multiplying systems and the pulsed neutron problem

The effect of random displacements of absorbers in a homogeneous multiplying medium is studied via the source sink method. Simulation is used to build up a statistical history sufficiently accurate to obtain the probability density function P(k) of the multiplication factor k. In addition, we examine the effect of randomly placed absorbers in the classic pulsed neutron problem in a pure moderator. In this case the probability density function P(λ) is constructed for the decay constant λ. These two problems are repeated using fixed absorber position but random absorber strength: the associated forms of P(k) and P(λ) are obtained. In order to assess the usefulness of perturbation theory to study such stochastic problems we have obtained, for varying degrees of plate absorption, the mean value and standard deviation of λ in the pulsed neutron experiment. Tables of results are given which give some guidance on the accuracy of perturbation theory and also illustrate some pitfalls that can arise if the method is used indiscriminately.

Diffusion cooling of thermal neutrons in basic rock minerals by Monte Carlo simulation of the pulsed neutron experiments

The pulsed neutron experiment (the variable geometric buckling experiment) in spherical geometry has been simulated using the MCNP code. The time decay of the thermal neutron flux has been observed as a function of the sample size. The thermal neutron diffusion cooling coefficient C with the correction F has been determined for three basic rock minerals (quartz, calcite, dolomite) at the given specific densities. The corresponding density-removed parameters have also been obtained.