Cylindrical Moderator Research Articles

Design of a cold-neutron source for the Bariloche LINAC with solid mesitylene as moderator material

We present the results of calculations performed with the code MCNP-4C relative to the neutron-field behaviour within the moderator for the Bariloche-LINAC cold-neutron source, using mesitylene at 89 K as moderating material. Throughout the design calculations we used preliminary nuclear-data libraries for that material that were previously generated and partially validated. The optimum dimensions for a slab and a cylindrical moderator were obtained, with and without a premoderator, from the point of view of neutron production and time-width of the neutron pulse.

Neutronic performance of rectangular and cylindrical coupled hydrogen moderators in wide-angle beam extraction of low-energy neutrons

The JSNS design criteria for the coupled hydrogen moderator requires a large number of beam lines. We therefore performed neutronic calculations to maximize time-integrated and pulse-peak intensities of low-energy neutrons for 100% para-hydrogen coupled moderators. It was found that a larger number of beam lines increased the total intensity due to the increase of the angular coverage although the average intensity of each beam line decreased a bit due to a finite reflector missing caused by the beam extraction holes in the reflector. In the case of a rectangular moderator, the spatial distribution of the vector flux of low-energy neutrons becomes undesirable at a large beam extraction angle ( 25 ∘ ). As an alternative, we propose a cylindrical moderator to substantially improve the spatial distribution. Time-integrated and pulse-peak intensities, pulse widths, and pulse decay characteristics were calculated as a function of the moderator diameter. The results indicated that the optimal diameter is about 140 mm. Compared to the rectangular moderator, the cylindrical design gives higher pulse-peak intensities with narrower pulse widths, without penalizing the time-integrated intensities, thus providing a more uniform angle dependence. Some explanations are given for the differences in pulse characteristics between the two moderator shapes.

Design of a single moderator-type neutron spectrometer with enhanced energy resolution in the energy range from a few to 100 keV

The moderator structure for a neutron spectrometer was optimized with the Monte Carlo code MCNP-4B. The spectrometer consists of a cylindrical moderator and a position-sensitive thermal neutron detector and obtains energy spectra from thermal neutron distribution along its cylindrical axis. The structure of the moderator was improved by using a low hydrogen density material on one end and a high hydrogen density on the other, and inserting a neutron absorber that eliminates thermal neutron diffusion. This design improves the energy resolution of the spectrometer, especially for low-energy neutrons from a few to 100 keV. The designed spectrometer can be applied to the measurement of energy spectra over a neutron energy range from a few keV to 20 MeV.

Thermal hydraulic tests for developing two-phase thermo-siphon loop of CARR-CNS

The China Institute of Atomic Energy (CIAE) is now constructing the China Advanced Research Reactor (CARR: 60 MW), and designing the cold neutron source (CNS) with a two-phase hydrogen thermo-siphon loop consisting of a condenser, a single-moderator transfer tube and a cylindrical annulus moderator cell. The mockup tests reported here were carried out on a full-scale loop using Freon-113 as the working fluid in order to validate the self-regulating characteristics of the loop, with a void fraction less than 20% in the liquid of the moderator cell and requirements for establishing the condition under which the inner shell of the cell has only vapor and the outer shell liquid. During these mockup tests, the density ratio of liquid to vapor and the vapor volumetric evaporation rate due to heat load were kept the same as those in normal operation of the CARR-CNS. The results show that the loop has the required self-regulating characteristics and the inner shell of the moderator cell contains only vapor, the outer shell liquid. The local void fractions in the liquid increase with an increase in loop pressure under the condition of a constant vapor volumetric evaporation rate.

Performance comparison of an241Am-Be neutron source-based PGNAA setup with the KFUPM PGNAA setup

Monte Carlo calculations have been carried out to compare the performance of an 241Am-Be neutron source-based prompt gamma neutron activation analysis (PGNAA) setup with that of the 2.8 MeV neutron-based PGNAA setup at King Fahd University of Petroleum and Minerals (KFUPM) to analyze Portland cement samples. This work is a part of a wide Monte Carlo studies being conducted at KFUPM in search of a more efficient neutron source for its 2.8 MeV neutrons, from the D(d,n) reaction, based PGNAA facility. In this study an 241Am-Be neutron source-based PGNAA setup was simulated. For comparison, the diameter of a cylindrical external moderator of the 241Am-Be neutron source, based PGNAA setup, was assumed to be similar to that used in the KFUPM PGNAA setup. The results of this study revealed that although the optimum geometry of the 241Am-Be neutron source-based setup is similar to that of the KFUPM PGNAA facility, the performance of the 241Am-Be neutron source-based setup is slightly poorer than that of the 2.8 MeV neutron-based setup.

Performance tests of external moderators of a PGNAA setup

Performance tests of external cylindrical moderators of an accelerator-based prompt gamma ray neutron activation analysis (PGNAA) setup have been carried out through thermal neutrons and prompt γ-ray yield measurements. The PGNAA setup is to be used for analysis of cement samples. This study was conducted to investigate the effects of geometry of cylindrical moderator on yield of thermal neutrons and prompt γ-rays for two different types of moderator assemblies. One of the moderators was to be used with a small sample and the other to be used with a large sample. Fast and thermal neutron yield was measured inside the sample volume as a function of the front moderator thickness as well as sample length. Neutron yield measurement was carried out at the King Fahd University of Petroleum and Minerals 350 keV pulsed beam accelerator using nuclear track detectors. The pulsed 200 keV deuteron beam with 5 ns pulse width and 31.25 kHz frequency was used to produce 2.8 MeV neutrons via D(d,n) reaction. Neutron yield measurements showed that the large sample moderator has a smaller yield of thermal neutrons as compared to the small sample moderator, which is in complete agreement with the results of Monte Carlo yield calculations of the thermal and fast neutrons from both the moderators. Finally, the prompt γ-ray yield from a Portland cement sample was measured using the two moderators and was compared with each other. As predicted by Monte Carlo simulations, in spite of a smaller yield of thermal neutrons, the large sample moderator has a higher yield of prompt γ-rays.

円筒型冷減速材容器設計のための衝突噴流実験及び解析

A cold moderator using liquid hydrogen is one of the key components in a MW-scale spallation target system. which directly affects the neutronic performance both in intensity and resolution. Flow visualization experiments were carried out in order to clarified a flow field in a cold moderator vessel, so that jet induced flows such as recirculation flows and stagnant regions were obserbed. As for a cylindrical moderator vessel, analytical results of velocity distributions obtained with a standard k-ε turbulent model agreed well with experimental results obtained with a PIV system. However, analytical heat transfer coefficients of a bottom of the vessel could not predict experimental values well.

Simulation and mockup tests for developing TRR-II CNS

The Taiwan Research Reactor improvement and the utilization promotion project (TRR-II) with Cold Neutron Source (CNS) was carried out at Institute of Nuclear Energy Research. The CNS with a two-phase thermosiphon loop consists of an annular cylindrical moderator cell, a single moderator transfer tube, and a condenser. The self-regulating characteristics of a two-phase thermosiphon loop are investigated against variations of heat load. The experiments on the thermal-hydraulic characteristics have been performed using a full-scale mockup loop and a Freon-11 was used as a working fluid. Two cases were evaluated by the simulation and experiments. One case is an ORPHEE-type moderator cell in which an inner shell is open at the bottom, the other case is one with an inner cavity with no hole at the bottom but a vapor inlet opening at the uppermost part of the cavity. The flooding limitations, liquid level and void fraction in the moderator cell as a function of the initial Freon-11 inventory and the heat load are also reported.

冷減速材容器設計のための衝突噴流実験及び解析

A cold moderator using liquid hydrogen is one of the key components in a MW-scale spallation target system, which directly affects the neutronic performance both in intensity and resolution. Flow visualization experiments were carried out in order to clarified a flow field in a cold moderator vessel, the impinging jet induced flows such as recirculation flows and stagnant regions were verified in a moderator vessel. As for a cylindrical moderator vessel, analytical results, such as turbulent energy distribution and heat transfer coefficients, obtained with a non-linear turbulent model predicted experimental results obtained with a PIV system better than those with other turbulent models.

Computational Study of a Small Size Neutron Moderator for Use on Neutron Dosemeters Based on Silicon Diodes

A small size polyethylene moderator for use on neutron dosemeters based on silicon diodes has been studied using the MCNP-4A code to determine the size and geometry that allow an improvement to the dosemeter's response for neutrons in the energy range 10 keV - 1 MeV. Various dimensions of cylindrical and spherical moderators with or without a central cylindrical opening have been modelled for 1 MeV starting neutrons at normal incidence on the detector area. Since silicon diodes provided with a borate polyethylene convertor allow thermal and fast neutron detection, the number of fast and thermalised neutrons reaching the detection area is observed as a function of the detector positioning inside the moderator for each geometry considered and compared with results obtained for Bonner Spheres. The evolution in the number of thermalised neutrons with starting neutron energy is shown for some of the geometries and a moderator with a natural boron absorber layer is proposed.

Energy Response of Neutron Area Monitor with Silicon Semiconductor Detector

Abstract A prototype neutron area monitor with a silicon semiconductor detector has been developed which has the energy response of 1 cm dose equivalent recommended by the 1CRP-26. Boron and proton radiators are coated on the surface of the silicon semiconductor detector. The detector is set at the center of a cylindrical polyethylene moderator. This moderator is covered by a porous cadmium board which serves as the thermal neutron absorber. Neutrons are detected as α-particles generated by the nuclear reaction 10B(n, α)7Li and as recoil protons generated by the interaction of fast neutrons with hydrogen. The neutron energy response of the monitor was measured using thermal neutrons and monoenergetic fast neutrons generated by an accelerator. The response was consistent with the 1cm dose equivalent response required for the monitor within ±34% in the range of 0.025–15 MeV.

Improvement on the THERMOS code (thermalization transport theory code) for calculating spherical geometry

The THERMOS-Q code is specifically designed to evaluate the thermal group cross section of spherical moderating-type neutron detectors, which have been widely used for neutron dosimetry in or around nuclear facilities. This code is a modified version of the THERMOS code (BNL version). The modifications are made on the following points: 1. 1)|A subprogram for treating a spherical geometry is added to the original code. 2. 2)|Any neutron energy spectrum of a slowing-down source may be fed into this code by users. 3. 3)| The maximum number of spatial meshes has been extended from 30 to 100. The thermal group cross sections have been calculated for spherical and cylindrical moderating-type neutron detectors with polyethylene moderators having different thicknesses. This result has proven that neutrons are more thermalized in a spherical moderator than in a cylindrical moderator with the same thickness. The experimental data are in good agreement with the values calculated by use of the thermal group cross section obtained by the THERMOS-Q code. The ANISN-W code is employed to calculate the responses.

Environmental neutron measurements around nuclear facilities with moderated-type neutron detector.

A cylindrical BF3 counter with cylindrical polyethylene moderators of 1-, 3-, 6- and 10-cm wall thickness were developed for neutron spectrometry, having response functions calculated by the time-dependent multi-group Monte Carlo code, TMMCR, and evaluated by a time-of-flight experiment in the energy region from epithermal to about 50 keV. This system of neutron detectors was used to measure the environmental neutron spectrum and dose equivalent around a 252Cf fission neutron source, a fast neutron source reactor, a cyclotron and an electron synchrotron. The neutron spectrum and dose obtained by this detector system showed very good agreement with those measured by a commercially available neutron rem counter and calculated by our MMCR Monte Carlo code. As a result of these comparisons, this detector system has been established as a standard spectrometer and dosimeter for environmental measurements around nuclear facilities.

A Measurement of the Neutronic Interaction Between Two Cylindrical Moderators

The stationary interaction parameter between the plane surfaces of two identical, cylindrical, polyethylene disks has been measured with the pulsed neutron method. It was shown that it is possible to overcome the delay effect caused by the flight of the neutrons in the gap between the moderators through the insertion was varied between 0.06 and 0.77; the resulting values of the interaction was varied between 0.06 and 0.77; the resulting values of the interaction parameter varied from 0.93 to 0.24. Two different thicknesses of the disks were employed; one assembly gave decay constants from 10 000 to 17 000 sec−1 and the other assembly yielded values in the range 17 000 to 40 000 sec−1. Although full equilibrium of the neutron distribution was not reached in all cases for the last series, the two series show a general agreement. A calculation based on a zero’th-order Bessel function as radial distribution and (cos θ)/π as angular distribution of the neutrons leaving the interacting surfaces gives a too-large interaction parameter for all separation distances. However, the overestimate is small when the surfaces are close together.